[Federal Register Volume 60, Number 148 (Wednesday, August 2, 1995)]
[Notices]
[Pages 39430-39462]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-10802]



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NUCLEAR REGULATORY COMMISSION

Biweekly Notice


Applications and Amendments to Facility Operating Licenses 
Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 7, 1995, through July 21, 1995. The 
last biweekly notice was published on Wednesday, July 19, 1996 (60 FR 
37084).

Notice Of Consideration Of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, And Opportunity For A Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By September 1, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be 

[[Page 39431]]
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested persons 
should consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units Nos. 1, 2, and 3, Maricopa County, Arizona

    Date of amendments request: July 3, 1995
    Description of amendments request: The proposed Technical 
Specification (TS) amendment temporarily adds new ACTION Statements 
3.8.1.1.f and 3.8.1.1.g to TS 3.8.1.1, ``A.C. Sources - Operating,'' to 
provide a method of responding to sustained degraded switchyard 
voltage. Bases 3/4.8.1, ``A.C. Sources,'' 3/4.8.2, ``D.C. Sources,'' 
and 3/4.8.3, ``Onsite Distribution Systems,'' are also being revised to 
provide guidance on how and why degraded offsite power voltage and the 
number of startup transformers in service affect compliance with GDC 17 
and to give the basis for the additional ACTION statements.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not significantly increase the 
probability of an accident previously evaluated in the Updated Final 
Safety Analysis Report (UFSAR). The safety function of the 
Electrical Distribution System (EDS) is to provide sufficient 
capacity and capability to assure that 1) specified acceptable fuel 
design limits and design conditions of the reactor coolant pressure 

[[Page 39432]]
boundary are not exceeded as a result of anticipated operational 
occurrences and 2) the core is cooled and containment integrity and 
other vital functions are maintained in the event of postulated 
accidents. In addition, it shall have sufficient independence, 
redundancy, and testability to perform its safety function assuming 
a single failure. The proposed ACTIONs will restore the EDS to 
conformance with General Design Criterion (GDC) 17 of Appendix A to 
10 CFR 50. Once in conformance with GDC 17, the system will be 
capable of performing its safety function as analyzed in Chapters 6 
and 15 of the UFSAR. The proposed temporary change has no effect on 
the probability of accident initiation, therefore, the probability 
of an accident previously evaluated has not been significantly 
increased.
    The consequences of an accident previously evaluated in the 
UFSAR will not be significantly increased. Restoring one train to 
OPERABLE, by blocking Fast Bus Transfer (FBT), within one hour is 
consistent with the response time of Technical Specification (TS) 
ACTION 3.0.3. The second train will be restored to OPERABLE by 
having its Emergency Diesel Generator (EDG) started, loaded, and 
separated from offsite power within two hours or FBT will be blocked 
within two hours. Action within two hours is consistent with the 
plants TS since TS ACTION 3.8.2.1.a, ``D. C. Sources - Operating,'' 
would be the most limiting requirement with one train of inoperable 
electric power. In a degraded voltage event, the ability of the 
Class 1E 125VDC battery chargers to perform their function is 
indeterminate, therefore, the Class 1E 125VDC batteries must be 
assumed to provide the 125VDC control power to the Class 1E 
Engineered Safety Features (ESF) circuit breakers for both of their 
sequences. The battery capacity calculations assume only one 
sequence. Once one train is restored to OPERABLE and the other 
trains EDG demonstrated to be OPERABLE by loading and separating 
from the grid, ACTION 3.8.1.1.a, for one INOPERABLE offsite power 
supply, allows operation to continue for up to seventy-two hours. If 
both trains are blocked, then both trains are OPERABLE.
    The proposed change will ensure that the train that blocks FBT 
will be in conformance with GDC 17 should a subsequent accident 
occur. As such, that train of ESF equipment will be supplied Class 
1E preferred and standby power in the manner assumed by Chapters 6 
and 15 analyses. Starting, loading, and separating the other trains 
EDG from offsite power ensures that the second train of ESF 
equipment is prepared to respond to any subsequent accident. This 
configuration presents one OPERABLE offsite circuit and two OPERABLE 
EDGs to any subsequent accident, and would be capable of 
withstanding the single failures in the UFSAR Table 15.0-0, ``Single 
Failures.'' Optionally, with both trains blocked, both are OPERABLE 
and would be capable of withstanding the single failures in the 
UFSAR Table 15.0-0, ``Single Failures.''
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Given the current licensing basis, the proposed temporary TS 
change does not create the possibility of an accident of a new or 
different kind. The plant is currently licensed to have both trains 
of FBT blocked when low switchyard voltages exist in order to 
prevent the loss of power generated by the nuclear power unit from 
causing the loss of the preferred power circuits. The proposed 
temporary TS ACTIONs 3.8.1.1.f and 3.8.1.1.g are being added as 
ACTIONs to prevent a double sequencing event from occurring. The 
train that is blocked is consistent with previous UFSAR Chapter 6 
and Chapter 15 safety analyses since it will conform to GDC 17 prior 
to the onset of the accident. Under this condition it will be able 
to contribute to the mitigation of an accident and withstand the 
effects of any single failure equal to its ability when initially 
analyzed and licensed. The EDG which is loaded and isolated from 
offsite power also contributes to GDC 17 compliance since the entire 
system can withstand a Loss of Offsite Power (LOP) and a single 
failure of an EDG. With both trains blocked, the EDS is in 
compliance with GDC 17 and is analyzed.
    It is understood that an accident of a different kind will exist 
if a degraded voltage condition occurs coincident with an accident 
(e.g., LOCA [versus the analyzed LOP + LOCA]). Should such an 
accident occur, the manual action described in the proposed ACTION 
statements could not be credited to protect the plant. However, the 
purpose of proposed ACTIONs 3.8.1.1.f and 3.8.1.1.g is to provide an 
appropriate response to degraded voltage prior to an accident by 
eliminating the malfunction of a different type (double sequencing) 
and an accident of a different type (e.g., degraded voltage + LOCA) 
for one train within one hour and for the second train within two 
hours. This duration of response is consistent with the required 
responses currently in the TSs 3.0.3, 3.8.2.1.a, and 3.8.1.1.a.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety has not been reduced in that the train 
which has FBT blocked prior to the onset of an accident will be in 
conformance with GDC 17 (which is the basis to TS 3/4.8.1). Since 
the blocked train is in conformance with GDC 17 prior to the onset 
of an accident, it will support the single failure analyses and the 
safety analyses to the extent previously analyzed and licensed. The 
train not blocked will have its EDG started, loaded, and separated 
from offsite power prior to the end of the second hour. Action 
within two hours is consistent with TS 3.8.2.1.a. The proposed 
action recovers one train of A.C. sources in one hour and places the 
plant in a configuration of one less power source than is required 
by LCO 3.8.1.1 within two hours. Currently, TS ACTION 3.8.1.1.a (one 
power source inoperable) has a duration of seventy-two hours. The 
proposed ACTION requires responses within time frames consistent 
with TSs 3.0.3, 3.8.2.1.a, and 3.8.1.1.a, and therefore, does not 
reduce the margin of safety. Optionally, restoration of the second 
train by blocking FBT within two hours is also consistent with 
response times required by TS 3.0.3 and 3.8.2.1.a and therefore, 
also does not reduce the margin of safety. TS 3.8.1.1.a would not be 
required with both trains of FBT blocked as all four AC power 
sources would then be OPERABLE.
    Regulatory Guide 1.93, ``Availability of Electric Power 
Sources,'' Revision 0, December 1974 recognizes that under certain 
conditions it may be safer to continue operation at full or reduced 
power for a limited time than to effect an immediate shutdown based 
on the loss of some of the required electric power sources. In an 
effort to minimize the risk to the health and safety of the public, 
the proposed ACTIONs 3.8.1.1.f and 3.8.1.1.g balance the risk of a 
forced shutdown against the risk of remaining at power with a 
degraded switchyard voltage.
    Probabilistic Risk Analysis (PRA) has compared the probability 
of a core melt event for 1) blocking fast bus transfer in one train 
after one hour for the next seventy-one hours, and in the second 
train after two hours for the next seventy hours; 2) blocking fast 
bus transfer in one train after the first hour for the next seventy-
one hours, and supplying power to the other train from the EDG after 
the second hour for seventy hours; and 3) a normal shutdown assuming 
the plant is in a normal configuration and no other transients or 
accidents except an uncomplicated reactor trip occurs during the 
shutdown process. Seventy-two hours was chosen for comparison 
purposes as the proposed ACTIONs would allow operation for up to 
seventy-two hours with one offsite circuit INOPERABLE.
    The PRA has shown that the probability of a core melt event 
during power operation with FBT blocked in one train after one hour 
for the next seventy-one hours, and in the second train after two 
hours for the next seventy hours is approximately 1.91E-6. The 
probability of a core melt event during power operation with FBT 
blocked in one train after one hour for the next seventy-one hours 
and the EDG powering the opposite train after the second hour for 
the next seventy hours (the proposed configuration) is between 
approximately 1.91E-6 and 1.93E-6. A range is provided because the 
current PRA model can only model blocking both trains or the EDGs 
supplying both trains. The risk lies somewhere between the two 
values. The probability of a core melt event due to a normal 
shutdown assuming the plant is in a normal configuration and no 
other transients or accidents except an uncomplicated reactor trip 
occurs during the shutdown process is 2.4E-6. The risk can not be 
calculated for a forced shutdown with degraded switchyard voltage 
present but it is expected to be higher. Therefore, the analysis 
provided is conservative.
    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

[[Page 39433]]

    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: July 14, 1995
    Description of amendment request: The proposed amendment would 
change the scram insertion times, Section 3.3.C, Minimum Critical Power 
Ratio section, Section 4.11.C and the associated bases in Section 2.1.1 
and 3/4.3.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Section 2.1 Bases - Safety Limits
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because equivalent fuel cladding protection (99.9 percent 
of all fuel rods do not experience transition boiling following a 
design basis transient) is provided.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed change does not affect the function of any 
structure, system or component.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety because the utilization of current General Electric 
fuel designs provides an equivalent margin of safety. As stated
    previously, equivalent fuel cladding protection is provided and 
ensures that 99.9 percent of all fuel rods will not experience 
transition boiling following a design basis transient.
    Section 3.3.C - Scram Insertion Times
    1. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant increase in the 
probability of consequences of an accident previously evaluated. The 
correlation of the scram insertion times with the actual notch 
position will simplify the surveillance procedure while maintaining 
the accuracy of the test.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because no physical modifications are associated with the proposed 
change and it does not affect the function of any structure, system 
or component.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety. The notch positions were chosen to coincide with 
the relative insertion values specified in the Technical 
Specifications. Use of the proposed combination of notch positions 
and scram insertion times will maintain the existing margins of 
safety that 99.9 percent of all fuel rods will not experience 
transition boiling following a design basis transient.
    Section 4.11.C - Minimum Critical Power Ratio (MCPR) Calculation 
Method
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the method used to calculate the measured scram 
speed distribution is consistent with the PNPS [Pilgrim Nuclear 
Power Station] licensing basis.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed change does not affect the function of any 
structure, system or component.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in the 
margin of safety because the proposed changes provide equivalent 
fuel
    cladding protection which ensures that 99.9 percent of all fuel 
rods will not experience transition boiling following a design basis 
transient.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Ledyard B. Marsh

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois

    Date of application for amendment requests: September 17, 1993, as 
supplemented July 20, 1995
    Description of amendment requests: As a result of findings by a 
Diagnostic Evaluation Team inspection performed by the NRC staff at the 
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company 
(ComEd, the licensee) made a decision that both the Dresden Nuclear 
Power Station and sister site Quad Cities Nuclear Power Station needed 
attention focused on the existing custom Technical Specifications (TS) 
used.
    The licensee made the decision to initiate a Technical 
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities. 
The licensee evaluated the current TS for both Dresden and Quad Cities 
against the Standard Technical Specifications (STS) contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' The licensee's evaluation identified numerous potential 
improvements such as clarifying requirements, changing TS to make them 
more understandable and to eliminate interpretation, and deleting 
requirements that are no longer considered current with industry 
practice. As a result of the evaluation, ComEd has elected to upgrade 
both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.
    The TSUP for Dresden and Quad Cities is not a complete adaption of 
the STS. The TSUP focuses on (1) integrating additional information 
such as equipment operability requirements during shutdown conditions, 
(2) clarifying requirements such as limiting conditions for operation 
and action statements utilizing STS terminology, (3) deleting 
superseded requirements and modifications to the TS based on the 
licensee's responses to Generic Letters (GL), and (4) relocating 
specific items to more appropriate TS locations.
    The September 17, 1993, and July 20, 1995, applications proposed to 
upgrade only Section 3/4.7 (Containment Systems) of the Dresden and 
Quad Cities TS.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. 
Implementation of these changes will provide increased reliability 
of equipment assumed to operate in the current safety analysis, or 
provide continued assurance that specified parameters remain within 
their acceptance limits, and as such, will not significantly 
increase the probability or consequences of a previously evaluated 
accident.
    Some of the proposed changes represent minor curtailments of the 
current 

[[Page 39434]]
requirements which are based on generic guidance or previously approved 
provisions for other stations. The proposed amendment for Dresden 
and Quad Cities Station's Technical Specification Section 3/4.7 is 
based on STS guidelines or later operating BWR plants' NRC accepted 
changes. Any deviations from STS requirements do not significantly 
increase the probability or consequences of any previously evaluated 
accidents for Dresden or Quad Cities Stations. The proposed 
amendment is consistent with the current safety analyses and has 
been previously determined to represent sufficient requirements for 
the assurance and reliability of equipment assumed to operate in the 
safety analysis, or provide continued assurance that specified 
parameters remain within their acceptance limits. As such, these 
changes will not significantly increase the probability or 
consequences of a previously evaluated accident.
    The associated systems that make up the Containment Systems are 
not assumed in any safety analysis to initiate any accident sequence 
for Dresden or Quad Cities Stations; therefore, the probability of 
any accident previously evaluated is not increased by the proposed 
amendment. In addition, the proposed surveillance requirements for 
the proposed amendments to these systems are generally more 
prescriptive than the current requirements specified within the 
Technical Specifications. The additional surveillance requirements 
improve the reliability and availability of all affected systems 
and, therefore, reduce the consequences of any accident previously 
evaluated, as the probability of the systems outlined within Section 
3/4.7 of the proposed Technical Specifications performing their 
intended function is increased by the additional surveillances.
    Create the possibility of a new or different kind of accident 
from any previously evaluated because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Other 
changes represent minor curtailments of the current requirements 
which are based on generic guidance or previously approved 
provisions for other stations. These changes do not involve 
revisions to the design of the station. Some of the changes may 
involve revision in the operation of the station; however, these 
provide additional restrictions which are in accordance with the 
current safety analysis, or are to provide for additional testing or 
surveillances which will not introduce new failure mechanisms beyond 
those already considered in the current safety analyses.
    The proposed amendment for Dresden and Quad Cities Station's 
Technical Specification Section 3/4.7 is based on STS guidelines or 
later operating BWR plants' NRC accepted changes. The proposed 
amendment has been reviewed for acceptability at the Dresden or Quad 
Cities Nuclear Power Stations considering similarity of system or 
component design versus the STS or later operating BWRs. Any 
deviations from STS requirements do not create the possibility of a 
new or different kind of accident previously evaluated for Dresden 
or Quad Cities Stations. No new modes of operation are introduced by 
the proposed changes. Surveillance requirements are changed to 
reflect improvements in technique, frequency of performance or 
operating experience at later plants. Proposed changes to action 
statements in many places add requirements that are not in the 
present technical specifications. The proposed changes maintain at 
least the present level of operability. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The associated systems that make up the Containment Systems are 
not assumed in any safety analysis to initiate any accident sequence 
for Dresden or Quad Cities Stations. In addition, the proposed 
surveillance requirements for affected systems associated with the 
Containment Systems are generally more prescriptive than the current 
requirements specified within the Technical Specifications; 
therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    Involve a significant reduction in the margin of safety because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Other 
changes represent minor curtailments of the current requirements 
which are based on generic guidance or previously approved 
provisions for other stations. Some of the later individual items 
may introduce minor reductions in the margin of safety when compared 
to the current requirements. However, other individual changes are 
the adoption of new requirements which will provide significant 
enhancement of the reliability of the equipment assumed to operate 
in the safety analysis, or provide enhanced assurance that specified 
parameters remain with their acceptance limits. These enhancements 
compensate for the individual minor reductions, such that taken 
together, the proposed changes will not significantly reduce the 
margin of safety.
    The proposed amendment to Technical Specification Section 3/4.7 
implements present requirements, or the intent of present 
requirements in accordance with the guidelines set forth in the STS. 
Any deviations from STS requirements do not significantly reduce the 
margin of safety for Dresden or Quad Cities Stations. The proposed 
changes are intended to improve readability, usability, and the 
understanding of technical specification requirements while 
maintaining acceptable levels of safe operation. The proposed 
changes have been evaluated and found to be acceptable for use at 
Dresden or Quad Cities based on system design, safety analysis 
requirements and operational performance. Since the proposed changes 
are based on NRC accepted provisions at other operating plants that 
are applicable at Dresden or Quad Cities and maintain necessary 
levels of system or component reliability, the proposed changes do 
not involve a significant reduction in the margin of safety.
    The proposed amendment for Dresden and Quad Cities Stations will 
not reduce the availability of systems associated with the 
Containment Systems when required to mitigate accident conditions; 
therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power Company, and Northeast Nuclear 
Energy Company, et al., Docket Nos.50-213, 50-245, 50-336, and 50-
423 Haddam Neck Plant, and Millstone Nuclear Power Station, Units 
1,2, and 3, Middlesex County and New London County, Connecticut

    Date of amendment request: June 6, 1995
    Description of amendment request: The proposed amendment will 
modify the size of the Plant Operations Review Committee (PORC) which 
will collectively have the experience and expertise in various areas of 
plant operation, and will clarify the composition of the Site 
Operations Review Committee (SORC).
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration 
(SHC), which is presented below:
    These proposed changes do not involve an SHC because the changes 
do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The PORC is an oversight group and helps to ensure that the 
units are operated in a safe manner. To accomplish this the PORCs 
provide their recommendations on the safety related activities to 
the Vice President - Haddam Neck Plant for Haddam Neck and to the 
respective Nuclear Unit Directors for Millstone. Each Millstone
    Unit has its own PORC. It is proposed that the members of the
    PORC be selected by the respective Nuclear Unit Director based 
on their knowledge and 

[[Page 39435]]
expertise in specific key plant functions. The Millstone Station has 
one SORC. The SORC is also an oversight group whose charter is to 
advise the Senior Vice President - Millstone Station on all matters 
related to nuclear safety at the Millstone site. The Haddam Neck 
Plant, being a single unit site, has one PORC, which advises the 
Vice President - Haddam Neck Plant. The members of the Haddam Neck 
Plant PORC will be selected by the Vice President - Haddam Neck 
Plant based on their knowledge and expertise in specific key plant 
functions. The PORC and SORC add to the defense-in-depth concept 
provided by the design, operation, maintenance, and quality 
oversight by promoting excellence through the conduct of their 
affairs and by maintaining a diligent watch over their 
responsibilities.
    These administrative changes will revise the composition section 
of the technical specifications for the PORC members. Millstone Unit 
individuals will be appointed by the Nuclear Unit Directors if the 
individual meets one or more of the following areas of expertise: 
Plant Operations, Engineering, Reactor Engineering, Maintenance, 
Instrumentation and Controls, Health Physics, Chemistry, Work 
Planning and Control, and Quality Services. The Haddam Neck Plant, 
due to its broader scope of review also include[s] an individual 
experienced in Security and specific experience in Electrical 
Maintenance and Mechanical Maintenance. The individuals who will 
serve on PORC shall continue to meet the criteria of ANSI N18.1-
1971. This approach is consistent with the standard technical 
specifications and NUREG 0800, Section 13.4. For SORC at the 
Millstone Station, the method of identifying who shall serve as Vice 
Chairperson has been modified for clarity. The Site Services 
Director position is proposed to be eliminated since this position 
no longer exists. The functions previously performed by this 
individual have been assumed by those individuals who currently 
serve on SORC. Finally, [the TS relating to] the individual who 
shall represent Quality and Assessment Services shall be modified to 
allow a qualified member of Quality and Assessment Services to serve 
on SORC.
    The remaining portions of the technical specifications related 
to PORC and SORC are not being revised.
    These modifications broaden the unit committee participation and 
reflect current organizational positions and will not increase the 
probability of occurrence or the consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed administrative enhancements to the composition of 
the PORC and Millstone Station SORC will not affect the way in which 
the units are physically operated. These administrative changes to 
PORC and SORC continue to meet the guidelines of ANSI N18.7-1976. 
The modifications to PORC and SORC continue to allow these groups to 
provide a thorough review of activities at the units.
    The proposed modification does not impact any initiating events, 
and, therefore, cannot create the possibility of any new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    These proposed administrative changes will not impact the margin 
of safety provided by PORC and SORC. The PORC and SORC will continue 
to be staffed by qualified individuals experienced in the operation 
of the plants. These administrative changes will modify how the 
composition of the PORC and SORC members are presented in the 
technical specifications, but will not adversely impact their 
ability to review and comment on operations at the units.
    These changes do not impact any protective boundaries nor do 
they impact the safety limits for the protective boundaries. These 
proposed changes are administrative in nature. Therefore, there is 
no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Russell Library, 123 Broad 
Street Middletown, Copnnecticut 06457, for the Haddam Neck Plant, and 
the Learning Resources Center, Three Rivers Community-Technical 
College, 574 New London Turnpike, Norwich, CT 06360, for Millstone 1, 
2, and 3.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan

    Date of amendment request: July 5, 1995
    Description of amendment request: The proposed amendment would 
change the Administrative Controls section of the Palisades Technical 
Specifications. The changes involve deleting training requirements in 
the Administrative Controls section, revising the Plant Review 
Committee composition, and revising the function and composition of the 
plant safety and licensing staff review requirements.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    A. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This change does not affect the probability or consequences of 
an accident. The changes are administrative, deleting an unnecessary 
specification on staff training requirements, eliminating the 
specific references to the Nuclear Engineering and Construction 
Organization (NECO) staff, and requiring that the Plant Review 
Committee (PRC) chairman, alternate chairman, and members be 
designated in administrative procedures by the Plant General 
Manager. Further administrative changes clarify the function of the 
Plant Safety and Licensing organization and eliminate the numerical 
requirement for five staff members to fulfill the organization 
function.
    The removal of an obsolete staff training requirement does not 
diminish the regulatory requirement to have an adequately trained 
staff. The accredited training programs for the plant staff ensure 
an appropriate level of training is conducted to maintain an 
appropriate skill and knowledge base for the staff. The requirements 
of 10CFR55 provide the necessary rules for operator licenses. Since 
a trained staff will be maintained, there will [be] no increase in 
the probability or consequences of an accident as a result of this 
change.
    The composition of the PRC will not be affected by this change 
as it will, at a minimum, be comprised of personnel from the 
operations, engineering, radiological services and maintenance 
departments as required by the Technical Specifications. The 
composition of the Plant Safety and Licensing organization as a 
whole may change. The function of the organization as it relates to 
these Technical Specifications, however, will not be affected. These 
changes have no affect on the plant accident analyses. Qualified 
personnel will continue to conduct the PRC and Plant Safety and 
Licensing reviews. Therefore, the changes do not increase the 
probability or consequences of an accident.
    B. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed changes are administrative and do not create the 
possibility of a new or different kind of accident. Staff training 
will continue to meet the accreditation requirements of the National 
Academy for Nuclear Training Accreditation Board and the 
requirements for the Systematic Approach to Training. Operators' 
license training will continue to meet the regulatory requirements 
of 10CFR55. Activities conducted by the Plant Review Committee and 
the Plant Safety and Licensing staff will continue to be 
accomplished by a staff which meets the qualification requirements 
of the Technical Specifications. These administrative changes will 
not affect the operation of the plant or the safety function of 
plant equipment nor will it affect the quality of the review 
activities. Therefore, there will be no possibility that a new or 
different kind of accident will be created.
    C. Involve a significant reduction in a margin of safety.
    The changes do not affect installed plant equipment nor do they 
affect plant 

[[Page 39436]]
operations. These administrative changes have not affected the 
probability or consequences of a previously analyzed accident or 
created the possibility of a new or different kind [of] accident 
from any previously evaluated. Therefore, they do not involve any 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: John N. Hannon

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of amendment request: October 11, 1994, as supplemented June 
23, 1995.
    Description of amendment request: The proposed amendments would 
revise Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-
2) Technical Specifications (TSs) 1.18, ``Quadrant Power Tilt Ratio,'' 
3/4.2.4, ``Quadrant Power Tilt Ratio,'' the Table Notation of TS Table 
3.3.-1, ``Reactor Trip System Instrumentation,'' and associated Bases 
to incorporate the guidance provided in the NRC's Improved Standard 
Technical Specifications (NUREG-1431) applicable to these TSs. The 
proposed amendments would clarify the requirements of the subject TSs 
with regard to the use of excore power range neutron flux detectors to 
monitor quadrant power tilt ratio when an excore power range neutron 
flux instrument is inoperable. The proposed change would also make 
several minor editorial changes in the subject TSs.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The existing quadrant power tilt ratio (QPTR) definition and 
Surveillance Requirement (SR) 4.2.4.c are inconsistent concerning 
reactor power limitations when performing QPTR surveillance 
requirements. The proposed change modifies these and related 
requirements to improve the understanding and consistency by 
generally incorporating the Improved Standard Technical 
Specification (ISTS) requirements of NUREG-1431.
    Editorial changes have been incorporated throughout the proposed 
specifications to address ISTS or plant specific convention and do 
not affect the accident analyses. The QPTR definition has been 
modified to reflect the ISTS wording and eliminate the inconsistency 
with SR 4.2.4.c. This change does not reduce the QPTR testing 
requirements or affect the accident analyses assumptions. The 
current action statements require power reduction along with a 
reduction in power range high neutron flux trip setpoints when the 
QPTR exceeds the limit. This ensures the core conditions are 
consistent with the accident analyses assumptions. With the modified 
action statements and the QPTR exceeding the limit, power reduction 
is also required along with performing a flux map to verify the 
peaking factors are within the accident analyses assumptions. In 
addition, the safety analyses must be re-evaluated to confirm the 
results remain valid prior to increasing power with an indicated 
tilt condition. The new action statements provide methods different 
from the current requirements. However, they satisfy the same 
objective, to ensure the conditions assumed in the accident analyses 
are maintained. Therefore, these changes will not involve 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The current surveillance requirements define the methods and 
frequencies for verifying the QPTR is within the limit specified in 
the limiting condition for operation. The proposed SRs include 
associated notes that allow separation of a power range channel into 
two portions made-up of the Nuclear Instrumentation System (NIS) and 
the excore detector portion. If an excore detector portion of a 
power range channel is inoperable, then the power range channel is 
inoperable since the detector provides input to the NIS which inputs 
to the solid state protection system. However, if the excore 
detector is operable and the NIS is inoperable, then the power range 
channel is inoperable but the ability to monitor the QPTR is 
unaffected. When the NIS portion of a channel is inoperable, 
appropriate actions are applied in accordance with Specification 
3.3.1. The new SRs continue to require the same testing and 
frequencies as the current SRs along with reducing the need to 
interpret the requirements when special conditions exist. Therefore, 
the proposed SRs will not affect the accident analyses or 
significantly increase the probability or consequences of an 
accident previously evaluated.
    Table 3.3-1 Action 2 applies when a power range channel is 
inoperable. This action has been reformatted to incorporate changes 
similar to those adopted in the QTPR SR which allow separation of a 
power range channel into the NIS portion and the excore detector 
portion. Proposed Action 2.a applies to an inoperable power range 
high neutron flux channel and Action 2.b applies to ``all other 
channels'' which includes the Low Setpoint function along with the 
High Positive and High Negative Rate functions. The new action is 
modified by Note (3) to allow bypassing the inoperable channel for 
surveillance testing and setpoint adjustment and by Note (4) that 
only requires performing SR 4.2.4 when the power range high neutron 
flux channel input to QPTR is inoperable. The new action does not 
require reducing the power range neutron flux setpoint like the 
current action since the proposed action is to perform the QPTR 
surveillance or shutdown which is more conservative than the current 
action requirement, otherwise, the new action requires essentially 
the same steps to be performed as the current action. Therefore, the 
proposed action will not affect the accident analyses or involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    These changes are proposed to allow flexibility in plant 
operations by modifying the QPTR action and surveillance 
requirements to allow separation of a power range channel into the 
NIS portion and the excore detector portion. The modified action and 
surveillance requirements continue to provide monitoring of those 
parameters required to ensure the core is operating safely. Since 
these changes are not significantly different from the current 
requirements and no change is being introduced that would affect the 
accident analyses assumptions, we have concluded that the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes incorporate modifications generally 
consistent with the ISTS QPTR requirements to ensure the core power 
distribution is adequately monitored. The revised action statements 
provide for peaking factor verification as a logical compensatory 
measure to ensure the core is operating within required limits. This 
is more conservative than the current requirements and provides 
additional assurance that Specification 3.2.4 will continue to 
govern the QPTR limitations in a manner consistent with the accident 
analyses assumptions. The revised SR provides clear and 
understandable testing requirements to reduce confusion concerning 
how the QPTR is to be monitored based on plant conditions. The 
proposed change does not introduce any new mode of plant operation 
or require any physical modification to the plant, therefore, this 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The QPTR limit ensures that the gross radial power distribution 
is maintained within the assumptions used in the safety analyses. 
The QPTR is one of the variables that is monitored to ensure the 
core operates within the bounds used in the safety analyses. When 
the QPTR is maintained below 1.02 it provides an indication that the 
peaking factors are within the limiting values by preventing and 
undetected change in the 

[[Page 39437]]
gross radial power distribution. The proposed changes ensure the 
required parameters are verified during the applicable conditions 
and on a consistent basis, therefore, these changes will not reduce 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: May 19, 1995
    Description of amendment request: The proposed amendments revise 
the specifications to permit the reactor building personnel airlock 
doors to remain open during fuel handling.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed change would allow the containment personnel 
airlock doors to remain open during fuel movement and core 
alterations. These doors are normally closed during this time period 
in order to prevent the escape of radioactive material in the event 
of a fuel handling accident. These doors are not initiators of any 
accident. The probability of a fuel handling accident is unaffected 
by the position of the containment personnel airlock doors.
    The proposed change alters assumptions made in evaluating the 
radiological consequences of a fuel handling accident inside the 
reactor containment building. Allowing the containment personnel 
airlock doors to remain open during fuel movement and core 
alterations does increase, however not significantly, the 
consequences of a fuel handling accident inside containment. 
Previously, the fuel handling accident inside containment was 
bounded by the fuel handling accident analysis in the spent fuel 
pool area of the auxiliary building. Part of the dose increase has 
been offset by the increase in the minimum decay time before 
irradiated fuel may be moved inside the reactor containment 
building. Extending the minimum decay time actually decreases the 
consequences of a fuel handling accident by reducing the radioactive 
inventory of the irradiated fuel which could possibly be released 
during a fuel handling accident. The revised fuel handling accident 
analysis results in maximum offsite doses of 43.4 Rem and 41.8 Rem 
to the thyroid and 0.616 Rem and 0.598 Rem to the whole body for 
ANO-1 and ANO-2, respectively. The calculated offsite doses are well 
within the limits of 10CFR Part 100. Also, the calculated doses are 
larger than the actual doses which would be expected during a fuel 
handling accident because the calculation does not incorporate the 
closing of at least one of the personnel airlock doors following 
evacuation of containment. The proposed change would significantly 
reduce the dose to workers in the containment in the event of a fuel 
handling accident by expediting the containment evacuation process.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed change does not involve the addition or 
modification of any plant equipment. Also, the proposed change would 
not alter the design, configuration, or method of operation of the 
plant.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    This proposed change has the potential for an increased dose at 
the site boundary due to a fuel handling accident; however, the dose 
remains within acceptable limits. The margin of safety as defined by 
10CFR Part 100 has not been significantly reduced. There is an 
increase in the calculated offsite dose resulting from a fuel 
handling accident; however, the increase is not significant and is 
well within the limits specified in 10 CFR Part 100. The overall 
significance will be offset by the increased minimum decay time, the 
decreased potential radiation dose to workers, and the increased 
availability of the personnel airlock door in the event of a fuel 
handling accident. Closing at least one of the personnel airlock 
doors following an evacuation of containment, further reduces the 
offsite doses in the event of a fuel handling accident which 
partially compensates for the higher offsite doses calculated as a 
result of this proposed change.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. BecknerEntergy Operations, Inc., 
Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, 
Arkansas
    Date of amendment request: March 17, 1995
    Description of amendment request: The proposed amendment revises 
requirements associated with channel functional tests of the core 
protection calculator following a high temperature alarm.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The core protection calculators (CPCs) are not accident 
initiators, therefore this change does not increase the probability 
of an accident previously evaluated.
    The core protection calculators (CPCs) are dedicated 
minicomputers that receive key parameters necessary to calculate the 
departure from nucleate boiling ratio (DNBR) and local power density 
(LPD) and issue a reactor trip command prior to reaching plant 
conditions that may damage the fuel in the reactor. Subjecting a 
computer to elevated temperatures may affect the reliability of the 
computer calculations. This change in the Arkansas Nuclear One-Unit 
2 (ANO-2) Technical Specifications (TS) will require a verification 
of the CPC operability, by the performance of a channel functional 
test, in the event a cabinet high temperature switch is actuated. 
This is a more accurate indication of the operating environment of 
the CPCs than the current requirement to perform the test based upon 
room temperature. The ability of the CPCs to monitor DNBR and LPD 
and issue a trip command when appropriate will not be affected in 
any way by this change, therefore the consequences of an accident 
previously evaluated are not increased.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    Because the proposed changes do not alter the design, 
configuration, or method of operation of the plant, they do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    These proposed changes do not alter the acceptance criteria of 
any surveillance requirements. The changes do not alter any 
assumptions used in accident analysis, change any actuation 
setpoints, nor allow 

[[Page 39438]]
operations in any configuration not previously analyzed. This change 
will trigger a verification of affected CPC operability based on 
cabinet temperature instead of room temperature, which is a more 
accurate indication of the operating environment of the CPC 
computer. Therefore, this change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of amendment request: April 4, 1995
    Description of amendment request: The proposed amendment revises 
operating criteria and requirements associated with containment 
personnel air locks.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the

     Probability or Consequences of an Accident Previously Evaluated.
    The containment air locks are passive components integral to the 
containment structure and are not evaluated to be accident 
initiators, therefore, the proposed amendment does not involve an 
increase in the probability of an accident previously evaluated.
    Each air lock door is rated for and tested to full design 
pressure of the containment building. If one door were inoperable in 
each air lock, the remaining door, since required to remain closed 
and locked, would provide the necessary fission product barrier to 
prevent an uncontrolled release, therefore the amendment allowance 
for an inoperable air lock door in each air lock does not increase 
the consequences of any previously evaluated accident.
    During a situation where one containment air lock door is 
inoperable and the operable door is opened, a breech in containment 
integrity would essentially exist while the operable door remains 
open. The time required for a containment air lock door to be open 
for ingress or egress does not exceed two to three minutes. The 
amendment provision to allow unlocking and opening an operable air 
lock door for ingress and egress to facilitate air lock maintenance 
necessary to restore operability does not increase the consequences 
of any previously evaluated accident since the time necessary for 
the door to be open is bounded by the existing one hour time 
allowance for an actual breech of containment integrity (TS 
3.6.1.1.)
    The containment air lock interlock functions to prevent 
simultaneous opening of both air lock doors thereby creating a 
breech in containment integrity. A dedicated individual stationed at 
the air lock to administratively control door operations, or locking 
closed an operable door will adequately assure containment 
integrity. The addition of this technical specification action 
statement, therefore, does not increase the consequences of any 
previously evaluated accident.
    Performance of the overall air lock leakage test requires 
opening the outer air lock door for installation of the mechanical 
dogging devices on the inner door. The current technical 
specifications make no provisions for this entry and thus would 
require a plant shutdown if the inner door was inoperable in an air 
lock. The proposed amendment removes the requirement to shut down 
when the barrel leak rate is due. The time required for the 
containment air lock doors to be opened for dog installation would 
be the same as for ingress and egress as discussed above, therefore 
this change does not increase the consequences of any previously 
evaluated accident.
    10 CFR 50, Appendix J contains containment leakage testing 
requirements, including specific requirements for containment 
building air locks. Changing the TS surveillance requirements to 
refer to 10 CFR 50, Appendix J for these test requirements will not 
degrade these tests, therefore this change does not increase the 
consequences of any previously evaluated accident.
    The air lock door seal pressure test is performed any time the 
air lock is used for containment access during modes of operation 
when containment integrity is required. The door seal test is 
intended to be a gross test to verify that the door seals were not 
damaged during the opening and closing cycle(s). This test does not 
replace the required overall barrel leakage test. Based on 
information provided by the air lock vendor, a test pressure of 10 
psig is sufficient to perform this gross seal verification. A change 
in the allowable leakage rate is requested to remove a specific 
numerical value from the TS surveillance requirements section and 
replace it with a fraction of LaG. This new acceptable leakage 
rate remains relatively insignificant and is bounded by the overall 
air lock leakage rate. Based on these facts this change in test 
pressure and associated acceptance criteria does not increase the 
consequences of any previously evaluated accident.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    Because the proposed changes do not change the design, 
configuration, or method of operation of the plant, they do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed changes to ANO-2 TS involve allowing brief breaches 
in containment integrity for the purpose of repairing inoperable air 
lock components or performing surveillances required by 10 CFR 50, 
Appendix J. These cases are adequately bounded by the one hour 
allowable outage time afforded by TS 3.6.1.1.
    The addition of a specific action statement addressing an 
inoperable air lock interlock provides those actions necessary to 
assure the maintenance of containment integrity. This is achieved by 
locking an operable door in the affected air lock when not in use 
and stationing a dedicated individual at the air lock, during 
periods of ingress and egress, whose sole responsibility is to 
insure only one air lock door is opened at a time thereby 
duplicating the function of the mechanical interlock.
    The proposed changes also consist of administrative changes 
removing an outdated exemption to 10 CFR 50, Appendix J and removing 
specific surveillance requirements from the specifications, instead 
referring to the controlling requirements of 10 CFR 50, Appendix J. 
This is consistent with the provisions of NUREG 1432 ``Revised 
Standard Technical Specifications for Combustion Engineering 
Plants,'' Rev. 0.
    None of the proposed changes increase the allowable overall air 
lock leakage rate, nor affect the acceptance criteria of the overall 
integrated containment leakage rate. All of the changes are bounded 
by existing analyses for all evaluated accidents and do not create 
any situations that alter the assumptions used in these analyses. 
Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of amendment request: May 19, 1995

[[Page 39439]]

    Description of amendment request: The proposed amendment adds 
criteria to address optional inspections of steam generator tubes.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    Steam generator tubes are inspected on a periodic basis to 
reduce the probability of a steam generator tube rupture or tube 
leakage. Five special interest groups are being added for optional 
inspections in addition to the general tube inspections currently 
required by the technical specifications. These special interest 
groups define areas of tubes where known or potential degradation 
mechanisms may exist for which additional inspection, above that 
currently required in the technical specifications, may be 
beneficial. Inspection of these special interest groups may utilize 
probes which more readily detect indications which may be found in 
the special interest areas. The increased detection capability will 
reduce the probability that a structurally significant flaw will go 
undetected during an inspection. The minimum sample size and 
expansion criteria (should a flaw be found) for inspections of 
special interest groups are based on percentages of tubes 
potentially affected by the specific degradation mechanisms for 
which the special inspection is being performed. The percentages 
used are the same as used for the current general tube inspections. 
The expansion criteria allow expansion within the area of interest 
without affecting the expansions of any general tube inspection. By 
expanding within the area of interest, a more complete inspection 
for the defects caused by a specific degradation mechanism can be 
performed than if the expansion were conducted in tubes not 
necessarily affected by the degradation mechanism, which is possible 
with the current technical specifications. Therefore, this change 
does not involve a significant increase in the probability of an 
accident previously considered.
    The proposed change does not increase the amount of radioactive 
material available for release or modify any systems used for 
mitigation of such releases during accident conditions. The steam 
generator tubing will continue to be examined on the frequency 
currently specified in the technical specifications. This change 
will allow steam generator examinations to focus on known areas of 
interest without requiring unnecessary expansion. The integrity of 
the steam generators will continue to be assured at an equivalent 
level. Therefore, the change does not involve a significant increase 
in the consequences of any accident previously evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    Special inspections such as the ones being added to the 
technical specifications have been conducted in the past at ANO-2. 
The method of inspection, pushing or pulling a probe through the 
steam generator tubes from the primary side, is the same method 
employed for the current technical specification required 
inspections. Inspection methodology is not being changed by 
incorporation of these special interest groups into the technical 
specifications. No design or operational characteristics of the 
plant are changed by the proposed amendment.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed amendment adds special interest groups for optional 
inspection into the technical specifications. These inspections 
concentrate on areas of interest using inspection methodology that 
is equivalent or better at finding specific types of flaws than the 
methodology used for the currently required general tube 
inspections. If the special interest groups are not inspected, the 
existing technical specification requirements for inspection still 
apply.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of amendment request: May 19, 1995
    Description of amendment request: The proposed amendment increases 
the allowed outage time for an emergency diesel generator from 72 hours 
to seven days. Additionally, the amendment authorizes one, ten-day 
diesel generator maintenance outage every fuel cycle.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The emergency diesel generators (EDGs) are backup alternating 
current power sources designed to power essential safety systems in 
the event of a loss of offsite power. EDGs are not an accident 
initiator in any accident previously evaluated. Therefore, this 
change does not involve an increase in the probability of an 
accident previously evaluated.
    The EDGs provide backup power to components that mitigate the 
consequences of accidents. The proposed changes to allowed outage 
times (AOTs) do not affect any of the assumptions used in 
deterministic safety analysis.
    In order to fully evaluate the EDG AOT extension, probabilistic 
safety analysis methods were utilized. The results of these analyses 
indicate no significant increase in the consequences of an accident 
previously evaluated. These analyses are detailed in CE NPSD-996, 
Combustion Engineering Owners Group ``Joint Applications Report for 
Emergency Diesel Generators AOT Extension.''
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    This proposed change does not alter the design, configuration, 
or method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed changes do not affect the technical specification 
limiting conditions for operation or their bases which support the 
deterministic analyses used to establish the margin of safety. 
Evaluations used to support the requested technical specification 
changes have been demonstrated to be either risk neutral or risk 
beneficial. These evaluations are detailed in CE NPSD-996.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of amendment request: May 19, 1995
    Description of amendment request: The proposed amendment increases 
the allowed outage time for an inoperable 

[[Page 39440]]
Safety Injection Tank (SIT) from one hour to 24 hours. Additionally, 
the amendment limits power operation to 72 hours when certain SIT 
related instrument functions are inoperable.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The Safety Injection Tanks (SITs) are passive components in the 
Emergency Core Cooling System. The SITs are not accident initiators 
in any accident previously evaluated. Therefore, this change does 
not involve an increase in the probability of an accident previously 
evaluated.
    SITs were designed to mitigate the consequences of Loss of 
Coolant Accidents (LOCA). These proposed changes do not affect any 
of the assumptions used in deterministic LOCA analysis. Therefore, 
the consequences of accidents previously evaluated do not change.
    In order to fully evaluate the effect of the SIT Allowable 
Outage Time (AOT) extension, probabilistic safety analysis (PSA) 
methods were utilized. The results of these analyses show no 
significant increase in the core damage frequency. As a result, 
there would be no significant increase in the consequences of an 
accident previously evaluated. These analyses are detailed in CE 
NPSD-994, Combustion Engineering Owners Group ``Joint Applications 
Report for Safety Injection Tank AOT/STI Extension.''
    The change pertaining to SIT inoperability based solely on 
instrumentation malfunction does not involve a significant increase 
in the consequences of an accident as evaluated and endorsed by the 
NRC in NUREG-1366, ``Improvements to Technical Specifications 
Surveillance Requirements.''
    Therefore, this change does not involve an increase in the 
probability or a significant increase in the consequences of any 
accident previously evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    This proposed change does not change the design, configuration, 
or method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed changes do not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses 
to establish the margin of safety. PSA evaluations were used to 
evaluate these changes. These evaluations demonstrated that the 
changes are either risk neutral or risk beneficial. These 
evaluations are detailed in CE NPSD-994.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner
    Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear 
One, Unit No. 2, Pope County, Arkansas
    Date of amendment request: May 19, 1995
    Description of amendment request: The proposed amendment increases 
the allowed outage time for one train of low pressure safety injection 
from 72 hours to seven days.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The low pressure safety injection system (LPSI) is part of the 
Emergency Core Cooling System subsystem. Inoperable LPSI components 
are not considered to be accident initiators. Therefore, this change 
does not involve an increase in the probability of an accident 
previously evaluated.
    The LPSI system was designed to mitigate the consequences of a 
large loss of coolant accident (LOCA). These proposed changes do not 
affect any of the assumptions used in deterministic LOCA analysis.
    In order to fully evaluate the LPSI AOT extension, probabilistic 
safety analysis methods were utilized. The results of these analyses 
indicate no significant increase in the consequences of an accident 
previously evaluated. These analyses are detailed in CE NPSD-995, 
Combustion Engineering Owners Group ``Joint Applications Report for 
Low Pressure Safety Injection System AOT Extension.''
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    This proposed change does not change the design, configuration, 
or method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed changes do not affect the technical specification 
limiting conditions for operation or their bases which support the 
deterministic analyses used to establish the margin of safety. 
Probabilistic evaluations used to support the requested technical 
specification changes have been demonstrated to be either risk 
neutral or risk beneficial. These evaluations are detailed in CE 
NPSD-995.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: June 26, 1995
    Description of amendment request: The amendment revises the snubber 
visual inspection intervals to match the schedule developed by the NRC 
staff for use with a 24 month refueling interval. This schedule was 
documented in Generic Letter 90-09. The licensee has made wording 
changes not contained in Generic Letter 90-09. These changes are as 
follows:
    a) Section 4.5.Q.1 - GL 90-09 wording ''...performance of the 
following augmented inservice inspection program in addition to the 
requirements of Section 4.0.5.''
    Proposed Technical Specification wording ''...performance of the 
following inspection program.''
    b) Section 4.5.Q.1.a - GL 90-09 wording ''...based on the criteria 
of Table 4.7.2 and the first inspection interval determined using the 
criteria shall be based upon the previous inspection interval 
established by the requirements in effect before Amendment (*). 
``Proposed Technical Specification wording ''...based on the criteria 
provided in Table 4.5.1.''
    c) Section 4.5.Q.1.b - GL 90-09 wording ''...All snubbers found 
connected to an inoperable common hydraulic fluid reservoir shall be 

[[Page 39441]]
counted as unacceptable for determining the next inspection interval.''
    Proposed Technical Specification deletes this sentence.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    The proposed amendment would revise the basis for the snubber 
visual inspection to be consistent with the bases described in Generic 
Letter 90-09.
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not affect the probability of 
occurrence nor does it affect the consequences of an accident 
previously evaluated as the requested visual inspection interval has 
been determined generically to be a safe and acceptable alternative 
to the existing visual inspection requirements as documented by the 
NRC in Generic Letter 90-09. With the completion of over 25 years of 
operating experience and only detecting one visual inspection 
failure, GPU Nuclear agrees that the existing intervals are overly 
conservative and can be extended to those described in the generic 
letter.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    As the requested change deals only with the frequency of visual 
inspection and not with the content, scope, or acceptance criteria 
of the inspection, no new or different type of accident has been 
created.
    3. Involve a significant reduction in the margin of safety.
    The margin of safety as defined in the bases of the Technical 
Specifications is not reduced as the requested requirements provide 
the same degree of confidence in snubber operability at the existing 
requirements.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Phillip F. McKee

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
Matagorda County, Texas

    Date of amendment request: May 31, 1995
    Description of amendment request: The proposed amendment would 
modify (by relocation to the Technical Requirements Manual) Technical 
Specification (TS) 3/4.1.2.1, Boration Systems/Flow Paths - Shutdown, 
TS 3/4.1.2.2, Boration Systems/Flow Paths - Operating, TS 3/4.1.2.3, 
Charging Pumps - Shutdown, TS 3/4.1.2.4, Charging Pumps - Operating, TS 
3/4.1.2.5, Borated Water Sources - Shutdown, TS 3/4.1.2.6, Borated 
Water Sources - Operating, TS 3/4.4.2.1, Safety Valves - Shutdown, and 
the associated Bases.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to the subject Technical Specifications is 
of an administrative nature in that the subject Technical 
Specifications and Bases will be relocated in their entirety to the 
Technical Requirements Manual. Future changes to the relocated 
requirements will be in accordance with 10CFR50.59 and approved 
station procedures.
    Whether the listed Technical Specifications and Bases are 
located in Technical Specifications or the Technical Requirements 
Manual has no effect on the probability or consequences of an 
accident previously evaluated.
    The proposed change does not alter the assumptions previously 
made in the listed Technical Specifications. The proposed change 
allows the Commission and the South Texas Project more effective use 
of personnel resources to control requirements that meet the four 
Criteria in the Final Policy Statement. The proposed change will not 
change the dose to workers.
    Since the probability of an accident is unaffected by 
administratively relocating the subject Technical Specification, and 
the doses are not affected and do not exceed acceptance limits, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to the subject Technical Specifications is 
of an administrative nature in that the subject Technical 
Specifications and Bases will be relocated in their entirety to the 
Technical Requirements Manual. Future changes to the relocated 
requirements will be in accordance with 10CFR 50.59 and approved 
station procedures. Whether the listed Technical Specifications and 
Bases are located in Technical Specifications or the Technical 
Requirements Manual has no effect on any previously evaluated 
accident. It does not represent a change in the configuration or 
operation of the plant and, therefore, does not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    The proposed change to the subject Technical Specifications is 
of an administrative nature in that the subject Technical 
Specifications and Bases will be relocated in their entirety to the 
Technical Requirements Manual. Future changes to the relocated 
requirements will be in accordance with 10CFR50.59 and approved 
station procedures. The margin of safety is not reduced when the 
requirements are relocated to a Licensee-controlled document because 
the requirements to change a License Basis Document via the 
10CFR50.59 process ensure the same questions concerning the margin 
of safety required for license amendments are asked. Therefore, this 
proposed change does not significantly reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location:  Wharton County Junior 
College, J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, 
Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: William D. Beckner

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 28, 1995
    Description of amendment request: The proposed amendment would 
revise technical specifications related to the standby liquid control 
(SLC) system. The proposed changes include increasing the required 
reactor pressure vessel boron concentration and modifying the SLC pump 
operability testing surveillance frequency from monthly to quarterly.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the 

[[Page 39442]]
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The current analysis requires the SLC system to be 
capable of bringing the reactor 3% delta - k subcritical assuming a 
cold xenon free condition. The increase in SLC storage tank boron 
concentration limits will ensure this capability is maintained for 
future reload cores using the same 3% delta - k shutdown reactivity 
margin without imposing restrictions in cycle exposure for current 
and future anticipated core configurations. The change in the 
surveillance frequency for SLC pump operability testing to once each 
three months is in agreement with the ASME Code. The relaxation of 
the testing interval for the SLC pumps decreases pump degradation, 
and eliminates an unnecessary burden on personnel resources without 
compromising plant safety. In addition, the administrative changes 
only correct typographical and editorial errors.
    Since these proposed changes do not affect precursors for any 
accident or transient analyzed in Chapter 14 of the USAR, there is 
no increase in the probability of any accident previously evaluated. 
Furthermore, since these changes will ensure the ability of the SLC 
system to mitigate the consequences of an accident for future 
anticipated core designs, they do not involve a significant increase 
in the consequences of any accident previously evaluated.
    2. The proposed changes will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The change in the SLC storage tank boron concentration 
limits will ensure that a cold xenon-free reload core can be brought 
to a subcritical condition as previously analyzed. The change in the 
frequency of the SLC pump operability testing to once each three 
months is in agreement with the ASME Code. The relaxation in the 
testing interval for the SLC pumps decreases pump degradation, and 
eliminates an unnecessary burden on personnel resources without 
compromising plant safety. In addition, the administrative changes 
only correct typographical and editorial errors.
    These proposed changes do not affect the design, function, or 
operation of the SLC or any other system. Also, these changes do not 
introduce any new modes of operation or modify existing equipment 
design. Therefore, they do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes will not create a significant reduction 
in the margin of safety. The proposed increase in the required boron 
concentration in the reactor pressure vessel will ensure the SLC 
system will be capable of bringing a cold xenon-free reload core 
subcritical while maintaining the 3% delta - k shutdown reactivity 
margin as specified in the previous operating cycle. The change in 
the frequency of SLC pump operability testing to once each three 
months is in agreement with the ASME Code. The relaxation in the 
testing interval for the SLC pumps decreases pump degradation, and 
eliminates an unnecessary burden on personnel resources without 
compromising plant safety. In fact, it increases SLC system 
availability. In addition, the administrative changes only correct 
typographical and editorial errors. Therefore, it is concluded that 
the requested changes do not create a significant reduction in the 
existing margin of safety as defined in the Technical 
Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, Nebraska 68602-0499
    NRC Project Director: William D. Beckner North Atlantic Energy 
Service Corporation, Docket No. 50-443, Seabrook Station, Unit No. 1, 
Rockingham County, New Hampshire
    Date of amendment request: June 16, 1995
    Description of amendment request: The proposed amendment would 
change the minimum boron concentration specified for the refueling 
water storage tank (RWST) in Limiting Condition for Operation (LCO) in 
Technical Specification (TS) 3.1.2.5 and would replace the minimum 
specified concentration for boron with an acceptable range of boron 
concentration for the RWST and the accumulators in the LCOs for TS 
3.1.2.6, 3.5.1.1, and 3.5.4.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has reviewed the licensee's analysis against the standards of 
10 CFR 50.92(c). The NRC staff's review is presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)) because the changes are proposed to assure that the post-
event shutdown margin required by the Technical Specifications will 
continue to be met and the consequences of a boron dilution event will 
remain as previously evaluated. The changes do not affect the design or 
manner of operation of any structure, system, or component important to 
safety.
    B. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because they do not affect the manner by which the 
facility is operated and do not involve a change to any structure, 
system, or component important to safety. The proposed changes merely 
assure that station will be operated within original design limits.
    C. The changes do not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)) because the proposed changes merely 
assure that the station will continue to be operated within the 
original design limits. Therefore, the acceptance criteria for 
previously evaluated accidents will continue to be met.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One 
International Place, Boston MA 02110-2624.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, 
Millstone Nuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of amendment request: July 11, 1995
    Description of amendment request: The proposed amendment modifies 
Technical Specification 3.5.F.7 to also allow the use of pull-to-lock 
switches to defeat the automatic initiation of the emergency core 
cooling system (ECCS) while in the refuel condition. The proposed 
amendment also makes administrative changes and makes changes to the 
associated Bases section.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    NNECO has reviewed the proposed change in accordance with 10 CFR 
50.92 and concluded that the change does not involve a significant 
hazards consideration (SHC). The basis for this conclusion is that 
the three criteria of 10 CFR 50.92(c) are not compromised. The 
proposed change does not 

[[Page 39443]]
involve an SHC because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    This change to LCO [Limiting Condition for Operation] 3.5.F.7(e) 
will allow an alternative means of de-energizing power to the 
selected ECCS pump motors during refueling. The current
    technical specification already allows these motors to be de-
energized. Use of the pull-to-lock switches provides a safer method 
of achieving this condition. The pull-to-lock condition of the 
switches is annunciated in the control room. Therefore, the switches 
will not be inadvertently left in the pull-to-lock position.
    Deletion of the statement that the 4160 volt supply breakers are 
racked in does not affect the requirement of LCO 3.5.F.7 to ensure 
the specified ECCS subsystems are OPERABLE.
    Therefore, there is no change in the probability or consequences 
of an accident previously analyzed due to this change.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The use of an alternative means of de-energizing power from the 
selected ECCS pump motors does not create a possibility of a new or 
different kind of accident. Using the control room pull-to-lock 
switch to disable the pump motor circuit breaker has the same effect 
on the ECCS pump as the removal of the circuit breaker from the 
switchgear.
    Deletion of the statement that the 4160 volt supply breakers are 
racked in does not affect the requirement of LCO 3.5.F.7 to ensure 
the specified ECCS subsystems are OPERABLE.
    3. Involve a significant reduction in the margin of safety.
    The proposed change to the Millstone Unit No. 1 Technical 
Specifications does not reduce the margin of safety. By using the 
control room pull-to-lock switches to disable the ECCS pump motors, 
instead of racking out the pump motor circuit breakers, it is 
possible to reenergize the ECCS pumps more quickly in an emergency, 
should one occur. The time savings can be translated into added 
safety margin from a shutdown risk perspective. The ability to 
disable and enable the pumps from the control room, instead of the 
switchgear area, also contributes to this added safety margin.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, 
Millstone Nuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of amendment request: July 18, 1995
    Description of amendment request: The proposed amendment request 
will add operability and surveillance requirements for reactor pressure 
vessel (RPV) overfill protection instrumentation. The proposed 
amendment will also add the associated Bases.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    NNECO has reviewed the proposed change in accordance with 10 CFR 
50.92 and concluded that the change does not involve a significant 
hazards consideration (SHC). The basis for this conclusion is that 
the three criteria of 10 CFR 50.92(c) are not compromised. The 
proposed change does not involve an SHC because the change would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The new LCO [Limiting Condition for Operation] and surveillance 
requirements ensure that the reactor high water level feedwater pump 
trip instrumentation is available. This technical specification 
change does not involve the addition of new equipment or logic. This 
change does not add new surveillance requirements for the 
instrumentation. This change simply establishes requirements for the 
operation and surveillance of
    reactor high water level feedwater pump trip instrumentation in 
the technical specifications. The implementation of this technical 
specification change will decrease the likelihood of an RPV 
overfill. No other postulated event is affected by the addition of 
this instrumentation to the technical specifications.
    Thus, adding the proposed requirements to the technical 
specifications will not increase the probability or consequences of 
any previously evaluated transients or accidents.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    No new failure modes are introduced by the addition of the 
reactor high water level feedwater pump trip instrumentation LCO and 
surveillance requirements. Modifying the technical specifications to 
formally add surveillance requirements already being performed in 
accordance with plant procedures will not modify plant response to 
any operational or transient event. Increasing the surveillance 
interval of the LITS [level indicating transmitter switches] from 
annual to once per operating cycle will not significantly affect 
reliability. Ensuring the operability of installed instrumentation 
does not add new or different kinds of accidents.
    Therefore, the new LCO and surveillance requirements do not 
create the possibility of a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The surveillance requirements being added in this change are 
consistent with current surveillances being performed for this 
instrumentation, with the exception that the LITS are currently 
calibrated on an annual rather than operating cycle basis. These 
surveillance and shutdown requirements ensure that protection from 
RPV overfill is maintained as assumed in the safety analyses.
    Therefore, there is no impact on the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut

    Date of amendment request: July 7, 1995
    Description of amendment request: The proposed change to technical 
specification 3/4.7.6 is being made to: 1) increase the allowable 
control room air conditioning (CRAC) system in-leakage from 100 cubic 
feet per minute (cfm) to 130 cfm; 2) provide a more conservative value 
for the maximum differential pressure across the high efficiency 
particulate air (HEPA) filters and charcoal adsorbers; 3) clarify that 
when the CRAC system is shifted to ``recirculation,'' this will be 
performed from the normal mode; and 4) modify the corresponding basis 
to reflect the above changes and to note that there are certain 
infrequent situations during which the control room emergency 
ventilation system (CREVS) will not automatically operate.
    Basis for proposed no significant haz- ards consideration 
determination: As

[[Page 39444]]

required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration (SHC), which is 
presented below:
    ...The proposed changes do not involve an SHC because the 
changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The CRAC system in the recirculation mode is used to mitigate 
the effects of an accident. Surveillance Requirement 4.7.6.1.e.2 has 
been modified to clarify that the system will automatically switch 
from the normal mode into a recirculation mode. This change and the 
proposed modifications to the acceptance criterion for the 
differential pressure across the HEPA filters and charcoal adsorbers 
and the increase in the control room in-leakage have no [e]ffect on 
the probability of an accident previously evaluated. The 
consequences of the accidents that have been previously evaluated 
have been reviewed to determine the impact of these proposed 
modifications. The increase in the in-leakage will affect the 
results of previously generated accident analysis. The accidents 
evaluated, namely the Millstone Unit No. 1 MSLB [main steam line 
break] and LOCA [loss-of-coolant accident], Millstone Unit No. 2 
LOCA, both high and low wind speed case, and Millstone Unit No. 3 
LOCA have been reviewed. The Millstone Unit No. 1 LOCA doses to the 
Millstone Unit No. 2 control room were qualitatively determined to 
be bounded by the Millstone Unit No. 2 LOCA cases. Therefore the 
Millstone Unit No. 1 LOCA was not performed. The remaining accidents 
were performed. The resultant doses are nearly identical to the 
existing doses found in the Millstone Unit No. 2 Final Safety 
Analysis Report and are all within the regulatory limits. To perform 
these revised control room dose calculations, NNECO used certain new 
assumptions which NNECO believes better model the control room and 
the effects the accident will have on the control room. The most 
significant change with the assumptions is the use of ICRP 30 in 
lieu of Regulatory Guide 1.109, Revision 1 for iodine dose 
conversion factors. The NRC has used ICRP 30 over the past 5 years 
for other applications and its use in this instance is appropriate.
    The change in the acceptance criterion for the differential 
pressure across the HEPA filter and charcoal adsorbers is a 
conservative modification in that the value given is a plant 
specific value and will be more indicative of blocked or clogged 
filters in actual plant conditions. These proposed changes do not 
have any negative impact on the consequences of any accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed modifications to Surveillance Requirement 4.7.6.1 
will clarify a portion of a surveillance requirement and will modify 
the differential pressure across the HEPA filters and the charcoal 
adsorbers. These changes will not create the possibility of a new or 
different kind of accident from any previously evaluated. The 
increase in the allowable control room in-leakage value from it[s] 
current level of 100 cfm to its new value of 130 cfm also does not 
create the possibility of a new or different kind of accident. The 
CRAC system is used to mitigate the consequences of an accident.
    3.Involve a significant reduction in the margin of safety.
    The proposed modifications do not decrease the margin of safety 
provided. Using the new accident assumptions, the limiting accidents 
were re-calculated to determine the impact on the Millstone Unit No. 
2 control room. These values are similar to the values found in the 
Millstone Unit No. 2 Final Safety Analysis Report and the Millstone 
Unit No. 2 Safety Evaluation Report and are within the regulatory 
limits established for the control room operators. Since the re-
calculated doses have been shown to be within limits, it has been 
concluded that there is no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of amendment request: June 8, 1995
    Description of amendment request: The Millstone Unit No. 3 
Technical Specification Section 3/4.8.4.3 requires removal of 
electrical power to the safety injection accumulator isolation valves 
in Modes 1, 2, 3, and 4 in order to protect the containment electrical 
penetrations and penetration conductors. Bases Section 3/4.8.4 states 
that containment electrical penetrations and penetration conductors are 
protected by either deenergizing circuits not required during normal 
plant operation (Modes 1 through 4) or by demonstrating the operability 
of primary and backup overcurrent protection circuit breakers during 
performance of periodic surveillances. It is proposed that Section 3/
4.8.4.3 will be deleted since the containment electrical penetration 
and penetration conductors for these circuits are protected by primary 
and backup penetration circuit breakers which are demonstrated to be 
operable by periodic surveillance testing.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration 
(SHC), which is presented below:
    The proposed changes do not involve an SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The revised Technical Specification Section 3.5.1 requirements 
will provide guidance to ensure that power to the accumulator 
isolation valves is removed when the accumulators are required to be 
operable and will clarify these requirements.
    Removal of the electrical penetration protection requirements of 
Section 3/4.8.4.3 is justified since Section 3/4.8.4.1 (Containment 
Penetration Conductor Overcurrent Protective Devices) will provide 
guidance to ensure that two breakers in series protect the 
electrical penetrations and penetration conductors against an 
overcurrent condition and the single failure of a circuit breaker. 
The two breakers in series also protect the Class 1E buses against a 
variety of overcurrent conditions including electrical faults which 
may be introduced due to the possible submergence of the accumulator 
isolation valves during a LOCA [loss-of-coolant accident].
    Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The amended Technical Specification Section 3.5.1 requirements 
will provide guidance to ensure that the accumulator isolation 
valves are deenergized when the accumulators are required to be 
operable. Deletion of the Technical Specifications Section 3.5.1 
requires that electrical power to the safety injection accumulator 
isolation valves (3SIL*MV8808A, B, C, D) be removed for the 
accumulators to be operable. This requirement prevents the 
inadvertent closure of these isolation valves which would block the 
safety function of the accumulators. Section 4.5.1.c requires 
demonstrating accumulator operability by ``At least once per 31 days 
when the RCS [reactor coolant system] pressure is above 1000 psig by 
verifying that power to the isolation valve operator is disconnected 
by removal of the breaker from the circuit.'' The surveillance 
requirements for verifying removal of power to the accumulator 
isolation valves for Section 4.5.1.c will be changed to ``At least 
once per 31 days when the RCS pressure is above 1000 psig by 
verifying that the associated circuit breakers are locked in a 
deenergized position or removed.''
    The proposed change will clarify requirements for securing these 
breakers in 

[[Page 39445]]
the off (tripped) position in the applicable modes. In addition, index 
page xi has been revised to reflect the deletion of Section 3/
4.8.4.3. Attachments 1 and 2 provide the mark-up and retyped pages 
of the Millstone Unit No. 3 Technical Specifications, respectively 
and reflect the currently issued version of the pages.
    Millstone Unit No. 3 Technical Specifications Section 3/4.8.4.3 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated since two breakers 
in series protect against an overcurrent condition and a single 
failure of a circuit breaker. The proposed amendment will not result 
in physical plant changes and there are no new credible failure 
modes. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The revised Technical Specification Section 3.5.1 will require 
that the accumulator isolation valves have their power deenergized 
when the accumulators are required to be operable. This requirement 
will maintain accumulator operability by assuring the accumulator 
isolation valves remain open.
    The removal of the Millstone Unit No. 3 Technical Specification 
Section 3/4.8.4.3 is safe since redundant circuit breakers in series 
for the accumulator isolation valves will provide assurance that the 
electrical penetration and penetration conductors are protected 
against overcurrent conditions. This will provide assurance that the 
containment boundary is intact.
    The proposed amendment will not adversely impact the physical 
protective boundaries (fuel matrix/cladding, RCS pressure boundary 
and containment) and therefore will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of amendment request: June 9, 1995
    Description of amendment request: The proposed amendment relocates 
Surveillance Requirement 4.6.6.1.d.3 for attaining a negative pressure 
in the secondary containment to Specification 3.6.6.2, Secondary 
Containment. The Action Statement of Section 3.6.6.1 is revised to 
decouple Sections 3.6.6.1 and 3.6.6.2. In addition, Definition 1.12, 
``Secondary Containment Boundary'' is deleted and included in the Bases 
Section 3/4.6.6, Secondary Containment. Bases Section 3/4.6.6.2, 
Secondary Containment is expanded using the guidance of the improved 
standard technical specifications (STS) for Westinghouse plants (NUREG-
1431).
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration 
(SHC), which is presented below:
    The proposed changes do not involve an SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to LCO [limiting condition for operation] 
3.6.1.2, LCO 3.6.6.1 and LCO 3.6.6.2 Action Statements, relocation 
of Surveillance Requirement 4.6.6.1.d.3 to Specification 3.6.6.2, 
changes to Bases Section 3/4.6.6.1, 3/4.6.6.2, and 3/4.6.6.3, and 
deletion of Definition 1.12 will resolve the conflict that currently 
exists between Specifications 3.6.6.1 and 3.6.6.2. Specifically, the 
requirement to establish and maintain a negative pressure in the 
secondary containment boundary included in Specification 3.6.6.1 
belongs to Specification 3.6.6.2. In the event Secondary Containment 
operability is not maintained, the Action Statement for LCO 3.6.6.2 
requires that Secondary Containment operability must be restored 
within 24 hours. Twenty-four hours is a reasonable completion time 
considering the limited leakage design of containment and the low 
probability of a DBA [design basis accident] occurring during this 
time period. Therefore, it is considered that there exists no loss 
of safety function. The proposed changes do not modify the LCO or 
surveillance acceptance criterion, nor do they change the frequency 
of the surveillances. The proposed changes do not involve any 
physical changes to the plant, do not alter the way any structure, 
system, or component functions. Therefore, the structures, systems, 
or components will perform their intended function when called upon. 
The proposed changes do not affect the probability of any previously 
evaluated accident. Additionally, the proposed changes are 
consistent with the new, improved STS for Westinghouse plants 
(NUREG-1431).
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not make any physical or operational 
changes to existing plant structures, systems, or components. The 
proposed changes do not introduce any new failure modes. The 
proposed changes simply resolve a conflict which currently exists 
between Specifications 3.6.6.1 and 3.6.6.2. Thus, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not have any adverse impact on the 
accident analyses. Also, the proposed changes resolve a conflict 
which currently exists between Specifications 3.6.6.1 and 3.6.6.2. 
The structures, systems, or components covered under Specifications 
3.6.6.1 and 3.6.6.2 will performed [sic] their intended safety 
function when called upon.
    Based on the above, there is no significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of amendment request: June 20, 1995
    Description of amendment request: The proposed amendment relocates 
the applicable requirements of Specification 3.6.3 for the main steam 
line isolation valves (MSIVs) to Specification 3.7.1.5, ``Main Steam 
Line Isolation Valves.'' In addition, the Applicability section of 
Specification 3.7.1.5 is revised to indicate that Specification 3.7.1.5 
is applicable in Mode 1 and in Modes 2, 3 and 4, except where all MSIVs 
are closed and deactivated (i.e., in Modes 2, 3, and 4, Specification 
3.7.1.5 is applicable only if the MSIVs are open). Also, the Action 
Statement for the Limiting Condition for Operation (LCO) 3.7.1.5 has 
been revised using the guidance of the improved standard technical 
specifications (STS) for Westinghouse plants (NUREG-1431).

[[Page 39446]]

    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration 
(SHC), which is presented below:
    The proposed changes do not involve an SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to the Applicability section, Action 
Statements, and Surveillance Requirements of Specification 3.7.1.5 
and the proposed changes to Specification 3.6.3 preserve the 
assumptions in the existing safety analysis. The proposed changes to 
the Applicability Section of Specification 3.7.1.5 will require the 
MSIVs to be operable in Mode 1 and in Modes 2, 3, and 4, except when 
closed and deactivated. The closure of the MSIVs in Modes 2, 3, or 4 
is acceptable because when they are closed, they are already 
performing their safety function. Since the MSIV closure time has 
not been changed, there is no adverse impact on the accidents 
previously evaluated.
    The proposed changes do not involve any physical changes to the 
plant, and do not alter the way any structure, system, or component 
functions. Therefore, the proposed changes do not affect the 
probability of any previously evaluated accident. Additionally, the 
proposed changes are consistent with the new, improved STS for 
Westinghouse plants (NUREG-1431).
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not make any physical changes to 
existing plant structures, systems, or components. When the MSIVs 
are closed and deactivated, they are already in the safe position; 
therefore, the proposed changes do not introduce a new failure mode. 
Additionally, the MSIV closure time (i.e., surveillance acceptance 
criterion) is not changed. The purpose of the surveillance is to 
ensure that the MSIVs can perform their safety function, and this 
requirement is preserved.
    Thus, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not revise the closure time of the 
MSIVs. This provides assurance that the MSIVs will perform their 
design safety function to mitigate the consequences of an accident. 
In addition, when they are closed in Modes 2, 3, and 4, they are 
already performing their safety function. Therefore, there is no 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station,Unit No. 1, Washington County, Nebraska

    Date of amendment request: June 26, 1995
    Description of amendment request: This proposed amendment would 
revise Technical Specification 2.3 to extend the allowed outage time 
(AOT) from 24 hours to 7 days for an inoperable low-pressure safety 
injection pump. This amendment request is a collaborative effort of 
participating Combustion Engineering Owners Group members and is based 
on an integrated assessment of plant operations and deterministic and 
probabilistic analyses.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The low pressure safety injection (LPSI) system is part of the 
emergency core cooling system. Inoperable LPSI components are not 
accident initiators in any accident previously evaluated. Therefore, 
these changes do not involve an increase in the probability of an 
accident previously evaluated.
    The LPSI system is primarily designed to mitigate the 
consequences of a large loss of coolant accident (LOCA). These 
proposed changes do not affect any of the assumptions in the 
deterministic LOCA analysis. Hence the consequences of accidents 
previously evaluated do not change.
    In order to fully evaluate the LPSI allowed outage time (AOT) 
extension, probabilistic safety analysis (PSA) methods were 
utilized. The results of these analyses show no significant increase 
in the core damage frequency. As a result, there would be no 
significant increase in the consequences of an accident previously 
evaluated. These analyses are detailed in CE NPSD-995, ``Combustion 
Engineering Owners Group Joint Applications Report for Low Pressure 
Safety Injection System AOT Extension.''
    The CEOG report reviewed the risk factors that are impacted by 
extending the AOT for a single LPSI pump from 24 hours to seven (7) 
days, and demonstrates that the increase in risk is negligible. In 
order to perform a more complete assessment of the overall change in 
risk, an accounting for avoided risks associated with reducing power 
and going to hot or cold shutdown was also considered. This 
``transition risk'' is important in understanding the trade-off 
between the risk of shutting down the plant compared with restoring 
a LPSI pump to operability while at power.
    In assessing overall plant risk, the risk avoided based on LPSI 
system maintenance while in cold shutdown must also be considered. 
Every time the plant is placed in cold shutdown, the LPSI system is 
required for decay heat removal when in the shutdown cooling mode of 
operation. Maintenance performed on the LPSI system during shutdown 
cooling operations may add to the risk of a loss of shutdown cooling 
event. Therefore, performing LPSI system maintenance with the unit 
on-line, when the LPSI system is not normally in demand, represents 
a decrease in shutdown risk.
    The CE study concluded that the change in core damage frequency 
due to increasing the LPSI AOT from 24 hours to seven (7) days is 
insignificant. Additionally, when the reduction in transition and 
shutdown risks are considered, it can be shown that there is an 
overall reduction in plant risk. Thus, it is the conclusion of the 
study that the overall plant impact will either be risk beneficial 
or risk neutral.
    Therefore, the proposed changes would not increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There will be no physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
implementation of setpoints or limits as a result of the proposed 
changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    These proposed changes do not affect the limiting conditions for 
operation or their bases used in the deterministic analyses to 
establish the margin of safety. PSA evaluations were used to 
evaluate this change. These evaluations demonstrate that the changes 
are either risk neutral or risk beneficial. These evaluations are 
detailed in CE NPSD-995. Therefore, the proposed changes do not 
involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 

[[Page 39447]]
    South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station,Unit No. 1, Washington County, Nebraska

    Date of amendment request: June 27, 1995
    Description of amendment request: This proposed amendment would 
revise Technical Specification 2.2 on the chemical and volume control 
system to reformat, clarify the requirements, and be more consistent 
with Combustion Engineering Standard Technical Specifications (STS) as 
presented in NUREG-0212, Revision 2.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes incorporate required actions, restrictions, and 
surveillance requirements for the Chemical and Volume Control System 
(CVCS) similar to Combustion Engineering Standard Technical 
Specifications (NUREG-0212 Revision 2).
    Technical Specification (TS) 2.2(1) specifies the requirements 
for borated water sources and flow paths when the reactor is 
subcritical and fuel is in the reactor. In order for a flow path to 
be operable, a charging or high pressure safety injection pump is 
required to be operable to inject the boric acid solution into the 
Reactor Coolant System. Currently this specification does not state 
any operability requirements for boric acid transfer pumps, charging 
pumps or high pressure safety injection pumps. In addition, this 
specification does not state any required actions to be taken if the 
borated water source or flow path is not operable.
    Therefore, the proposed changes incorporate requirements for the 
CVCS during shutdown into separate Limiting Conditions for 
Operations (LCOs) that will address the requirements for borated 
water sources, boric acid flow paths, charging pumps, and boric acid 
transfer pumps.
    The proposed changes delete operability and surveillance 
requirements for level instrumentation on the boric acid storage 
tanks. Level instrumentation by itself does not fulfill a safety 
function. The proposed changes will still require verification of 
tank level.
    Additionally, level instrumentation on the boric acid storage 
tanks does not meet any of the four criteria for inclusion into 
Technical Specifications as presented in the Final Policy Statement 
on Technical Specifications Improvements. This instrumentation is 
not installed instrumentation used to detect a significant 
degradation of the RCS boundary, a design feature or operating 
restriction that is an initial condition of a Design Basis Accident, 
a component that is part of the primary success path or actuates to 
mitigate a DBA, nor is it a component that has been shown to be 
significant to public health and safety. Therefore, testing and 
maintenance of the level instrumentation will be controlled outside 
of the TS.
    TS 2.2(3) specifies the Modifications of Minimum Requirements 
that are allowed during Power Operation. This specification is 
inconsistent with TS 2.2(2) which states the minimum requirements 
and is incomplete as it does not address components during Modes 3, 
4, and 5. The proposed changes incorporates consistent allowed 
outage times for the various components, and additional required 
actions for component inoperability during Modes 4 and 5 when fuel 
is in the reactor.
    The proposed changes incorporate additional operability 
requirements for the CVCS and required actions to be taken for CVCS 
component inoperability during Modes 4 and 5 when fuel is in the 
reactor. The proposed changes delete inconsistencies and clarify 
operability requirements for the CVCS whenever the reactor coolant 
temperature (Tcold) is greater than or equal to 210 degrees F, 
and ensures that operation of the system is consistent with its 
design bases. The proposed changes also revise the allowed outage 
time for CVCS components from 24 hours to 72 hours based on Standard 
Technical Specifications. This change is insignificant based on the 
FCS plant specific probabilistic risk assessment. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There will be no physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
implementation of setpoints or limits as a result of this proposed 
change. No new modes of operation are proposed. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes incorporate additional operability 
requirements, delete inconsistencies, and clarify operability 
requirements for the CVCS to ensure that operation of the system is 
consistent with its design bases. Therefore, the proposed changes do 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station,Unit No. 1, Washington County, Nebraska

    Date of amendment request: July 11, 1995
    Description of amendment request: The proposed amendment would 
allow up to 24 hours to restore Safety Injection Tank (SIT) operability 
if the SIT is inoperable due to level and/or pressure outside 
prescribed limits or if the associated isolation valve is in other than 
the full open position. The proposed change would also allow up to 72 
hours to restore SIT operability if the SIT is inoperable due to boron 
concentration outside prescribed limits.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The safety injection tanks (SITs) are passive components in the 
emergency core cooling system. The SITs are not an accident 
initiator in any accident previously evaluated. Therefore, this 
change does not involve an increase in the probability of an 
accident previously evaluated.
    SITs were designed to mitigate the consequences of a loss of 
coolant accident (LOCA). These proposed changes do not affect any of 
the assumptions used in deterministic LOCA analysis. Hence the 
consequences of accidents previously evaluated do not change.
    In order to fully evaluate the affect of the SIT allowable 
outage time (AOT) extension, probabilistic safety analysis (PSA) 
methods were utilized. The results of these analyses show no 
significant increase in the core damage frequency. As a result, 
there would be no significant increase in the consequences of an 
accident previously evaluated. These analyses are detailed in CE 
NPSD-994, ``Combustion Engineering Owners Group Joint Applications 
Report for Safety Injection Tank AOT/STI Extension.''
    The AOT extension based upon boron concentration outside the 
prescribed limits 

[[Page 39448]]
does not involve a significant increase in the consequences of an 
accident as evaluated and approved by the NRC in NUREG-1432, 
``Standard Technical Specifications for Combustion Engineering 
Plants.'' This proposed change is applicable to FCS.
    Therefore, the proposed changes would not increase the 
probability or consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There will be no physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
implementation of setpoints or limits as a result of these proposed 
changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses 
to establish the margin of safety. PSA evaluations were used to 
evaluate these changes. These evaluations demonstrated that the 
changes are either risk neutral or risk beneficial. These 
evaluations are detailed in CE NPSD-994. Therefore, the proposed 
changes do not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

Philadelphia Electric Company, Docket No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: June 23, 1995
    Description of amendment request: This Technical Specifications 
(TS) Change Request involves a one-time (i.e., temporary) change 
affecting the Allowed Outage Time (AOT) for the Emergency Service Water 
(ESW) System; Residual Heat Removal Service Water (RHRSW) System; the 
Suppression Pool Cooling, the Suppression Pool Spray, and Low Pressure 
Coolant Injection (LPCI) modes of the Residual Heat Removal (RHR) 
System; and Core Spray System to be extended from 3 and 7 days to 14 
days during the Limerick Generating Station (LGS), Unit 1, sixth 
refueling outage scheduled to begin January, 1996. This proposed 
extended AOT will allow adequate time to install isolation valves and 
cross-ties on the ESW and RHRSW Systems to facilitate future 
inspections or maintenance.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed Technical Specifications changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed one-time TS changes will not increase the 
probability of an accident since it will only extend the time period 
that the 'A' ESW and RHRSW loops and the affected equipment can be 
out-of-service. The extension of the time duration that certain 
equipment is out-of-service has no direct physical impact on the 
plant. The proposed inoperable systems are normally in a standby 
mode while the unit is in OPCON 1 or 2 and are not directly 
supporting plant operation. Therefore, they can have no impact on 
the plant that would make an accident more likely to occur due to 
their inoperability.
    During transients or events which require these systems to be 
operating, there is sufficient capacity in the operable loops to 
support plant operation or shutdown, in-so-much that failures that 
are accident initiators will not occur more frequently than 
previously postulated.
    In addition, the consequences of an accident previously 
evaluated in the SAR [Safety Analysis Report] will not be increased. 
With the 'A' loops of ESW and RHRSW inoperable, a known quantity of 
equipment is either inoperable or the equipment is not fully capable 
of fulfilling its design function under all design conditions due to 
certain support systems not being operable. Based on the support 
functions of the ESW and RHRSW systems, a review of the plant was 
performed to determine the impacts that the inoperable ESW and RHRSW 
'A' loops would have on other systems. The impacts were identified 
for each system, as discussed in the preceding Safety Assessment, 
and it was determined whether there were any adverse affects on the 
systems. It was then determined how the adverse affects would impact 
each system's design basis and overall plant safety. The 
consequences of any postulated accidents occurring on Unit 2 during 
this AOT extension was found to be bounded by the previous analyses 
as described in the SAR.
    The existing AOTs limit the amount of time that the plant can 
operate with certain equipment inoperable, where single failure 
criteria is still met. The minimum equipment required to mitigate 
the consequences of an accident and/or safely shutdown the plant 
will be operable or the plant will be shutdown. Therefore, by 
extending certain AOTs and extending the assumptions concerning the 
combinations of events and single failures for the longer duration 
of each extended AOT, we conclude, based on the evaluations above, 
that at least the minimum equipment required to mitigate the 
consequences of an accident and/or safely shutdown the plant will 
still be operable during the extended AOT. Therefore, the 
consequences of an accident previously evaluated in the SAR will not 
be increased.
    Therefore, these proposed one-time TS changes will not result in 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed one-time TS changes will not create the possibility 
of a different type of accident since it will only extend the time 
period that the 'A' ESW and RHRSW loops and the affected equipment 
can be out-of-service. The extension of the time duration that 
certain equipment is out-of-service has no direct physical impact on 
the plant and does not create any new accident initiators. The 
systems involved are either accident mitigation systems, safe 
shutdown systems or systems that support plant operation. All of the 
possible impacts that the inoperable equipment may have on its 
supported systems were previously analyzed in the SAR and are the 
basis for the present TS ACTION statements and AOTs. The impact of 
inoperable support systems for a given time duration was previously 
evaluated and any accident initiators created by the inoperable 
systems was evaluated. The lengthening of the time duration does not 
create any additional accident initiators for the plant.
    Therefore, the proposed one-time TS changes will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The ESW and RHRSW systems and their supported systems are 
designed with sufficient independence and redundancy such that the 
removal from service of a component/subsystem will not prevent the 
systems from performing their required safety functions. Since 
removal of an ESW and a RHRSW loop from service with one unit in 
operation and the other unit in a refueling outage is allowed by the 
current Technical Specifications, then the concern is the reduced 
margin of safety incurred by extending the affected AOTs.
    The present ESW and RHRSW AOT limits were set to ensure that 
sufficient safety-related equipment is available for response to all 
accident conditions and that sufficient decay heat removal 
capability is available for a LOCA/LOOP [Loss-of-Coolant Accident/
Loss-of-Offsite Power] on one unit and simultaneous safe shutdown of 
the other unit. A slight reduction in the margin of safety is 
incurred during the proposed extended AOT due to the increased risk 
that an event could occur in a fourteen day period versus a three or 
seven day period. This increased risk is judged to be minimal due to 
the low probability of an event occurring 

[[Page 39449]]
during the extended AOT and based on the following discussion of 
minimum ECCS [Emergency Core Cooling System]/decay heat removal 
requirements.
    The reduction in the margin of safety is not significant since 
the remaining operable ECCS equipment is adequate to mitigate the 
consequences of any accident. This conclusion is based on the 
information contained in the UFSAR [Updated Final SAR] reference 
documents NEDO-24708A and NEDC-30936-A. These documents describe the 
minimum requirements to successfully terminate a transient or LOCA 
initiating event (with scram), assuming multiple failures with 
realistic conditions were used to justify certain TS AOTs per UFSAR 
sections 6.3.1.1.2.o and 6.3.3.1. The minimum requirements for short 
term response to an accident would be either one LPCI pump or one 
Core Spray loop in conjunction with ADS [Automatic Depressurization 
System], which would be adequate to re-flood the vessel and maintain 
core cooling sufficient to preclude fuel damage. For long term 
response, the minimum requirements would be one loop of RHR for 
decay heat removal, along with another low pressure ECCS loop. These 
minimum requirements will be met since implementation of the 
proposed TS changes will require the operability of HPCI [High 
Pressure Coolant Injection], ADS, two LPCI subsystems (or one LPCI 
subsystem and one RHR subsystem during decay heat removal) and one 
Core Spray subsystem be maintained during the 14 day period. A 
Special Procedure will be written to ensure the operability of 
specified components and that other appropriate compensatory 
measures are implemented.
    Compensatory measures will be taken prior to or during the 
proposed extended AOT for those fire regions that rely on one or 
more safe shutdown methods which would all be unable to safely 
shutdown the plant with inoperable loops of the ESW and RHRSW 
systems or the inoperable systems that ESW or RHRSW support. These 
compensatory measures will offset the increased risk of a fire event 
occurring in the vulnerable areas, during the fourteen day versus 
three day AOT period. Therefore, the proposed extended AOT does not 
adversely affect the approved level of fire protection as described 
in UFSAR Appendix 9A (Fire Protection Evaluation Report).
    A Special Procedure will be written to administratively control 
the requirement to maintain the operability of specified components 
and implementation of any appropriate compensatory measures which 
are deemed necessary during the proposed AOT. In addition, 
operations personnel are fully qualified by normal periodic training 
to respond to and mitigate a Design Basis Accident, including the 
actions needed to ensure decay heat removal while LGS Unit 1 and 
Unit 2 are in the operational configurations described within this 
submittal. Accordingly, procedures are already in place that cover 
safe plant shutdown and decay heat removal for situations applicable 
to those in the proposed AOTs.
    A Probabilistic Safety Assessment (PSA) Study was performed for 
an ESW and RHRSW loop being out-of-service for 14 days on an 
operating unit. The Core Damage Frequency (CDF) increased by 
3.14x10-6, from 5.11x10-6 /reactor-year to 8.25x10-6/
reactor-year. In absolute terms, this is not a significant increase 
in risk. In addition, the modifications to be installed during this 
proposed extended AOT will allow for future maintenance and 
inspections to be performed on the ESW and RHRSW loops without 
removing an entire loop from service, which will reduce risk in the 
future. For example, if the ESW loop unavailability, due to testing 
or maintenance, is reduced by half, the CDF will decrease by more 
than four percent. It will also minimize the potential need for 
future AOT extensions on these systems.
    Therefore, the implementation of the proposed one-time TS 
changes will not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of amendment request: September 29, 1994
    Description of amendment request: The proposed Technical 
Specification changes represent revisions to Sections 3/4.3.7.2 
``Seismic Monitoring Instrumentation'' and 3/4.3.7.3 ``Meteorological 
Instrumentation.'' The proposed revisions remove the requirements from 
the Technical Specifications and relocates the appropriate descriptive 
information and testing requirements to the Hope Creek Updated Final 
Safety Analysis Report (UFSAR).
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve no hardware changes, no changes to 
the operation of any systems or components, and no changes to 
existing structures. Neither the relocation of the seismic/
meteorological specifications to the UFSAR nor the elimination of 
the Special Report requirements represent changes that affect plant 
safety or alter existing accident analyses.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes are procedural in nature concerning the 
operability and surveillance of instrumentation that are not safety 
related and will not impact the operation of any plant safety 
related component or equipment. Therefore, these changes will not 
create a new or unevaluated accident or operating condition.
    3. Will not involve a significant reduction in a margin of 
safety.
    In accordance with the guidance provided by the NRC regarding 
the improvement of Technical Specifications, SECY-93-067, the 
proposed changes relocate the seismic and meteorological 
instrumentation portions of the Technical Specifications, with the 
exception of the Special Report requirements, to the UFSAR. These 
instruments are not safety related and do not have any associated 
safety margins which could be affected by this change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of amendment request: November 23, 1994
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) would revise TS 4.8.2.1, ``Electrical 
Power Systems D. C. Sources, Surveillance Requirements,'' and 
associated Bases Section B 3/4.8.2. The proposed changes would (1) 
increase the terminal voltage acceptance criteria for the battery 
discharge test from 106 to 108 VDC, (2) delete a ``one time only'' test 
that is no longer applicable, (3) delete the battery load profile from 
the TS, and (4) revise TS Table 4.8.2.1-1, ``Battery Surveillance 
Requirements,'' to agree more closely 

[[Page 39450]]
with the BWR4 Standard Technical Specification format.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    ....will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed changes restore the conservatism to the battery 
voltage requirements by raising the minimum acceptable terminal 
voltage for the 125 VDC system in order to support proper operation 
of the connected loads. This change will cause no change in the 
probability of any accident and will, by providing increased support 
for connected loads, provide assurance [that] the consequences of 
previously evaluated accidents remain within limits. Removal of the 
load profile table does not affect the surveillance test loading 
which is contained in the station procedures. The (*) footnote 
deletion is purely editorial and has no safety bearing. Table 
changes agree with the format and wording of the improved BWR4 
Standard Technical Specifications.
    2....will not create the possibility of a new or different kind 
of accident from any previously evaluated.
    The revision of the battery sizing calculations did not change 
the design base requirement to supply the designed load for a duty 
cycle of 4-hours. The proposed change to the minimum acceptable 
battery terminal voltage for the 125 VDC system ensures proper 
voltages at the battery loads. No other changes to the physical 
plant or to the manner in which it is operated are caused by the 
proposed amendment; therefore, there is no new or different kind of 
accident created by this change.
    3....will not involve a significant reduction in a margin of 
safety.
    The revision of the battery sizing calculations did not change 
the design base requirement to supply the designed load for a duty 
cycle of 4-hours; however, battery capacity sizing parameter of end 
cell voltage was changed to a more conservative value to account for 
minimum load voltage requirements. Load profiles for these batteries 
were slightly modified to incorporate more precise yet conservative 
load current values. These batteries were evaluated using a 25% 
additional capacity margin for aging as required by IEEE-450. In 
addition, the batteries have a design margin of 5 to 10% for load 
growth and/or less than optimum operating condition of the battery; 
thereby, maintaining safety margins. Additionally, changes are 
comparable to the format and ACTIONS of the improved BWR4 STS. 
Permitting 31 days to restore a battery to within CATEGORY A and/or 
B limits per the improved BWR4 STS does not involve a reduction in 
any margin of safety since the battery, in Category C, remains 
operable, as discussed in the BASES.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of amendment request: November 28, 1994
    Description of amendment request: The proposed Technical 
Specification (TS) revisions provide as follows: (1) The setpoints and 
allowable values for the Average Power Range Monitor (APRM) flow-biased 
upscale scram/control rod block would be modified to improve operating 
margin in the Extended Load Line Limit Analysis (ELLLA) region; (2) The 
proposed changes to the Rod Block Monitor (RBM) trip function would 
transfer control of the setpoint and allowable value for the RBM - 
upscale rod block to the Core Operating Limits Report (COLR); (3) For 
the Reactor Coolant System (RCS) recirculation flow upscale trip 
function, the proposed changes would revise the trip setpoint and 
allowable value to reflect 105% of rated core flow.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    A. Changes to APRM Flow-Biased Scram/Control Rod Block
    The proposed changes to the Average Power Range Monitor (APRM) 
flow-biased scram/control rod block setpoints and allowable values 
were evaluated using NRC approved procedures and methods. The 
results of this evaluation are demonstrated in NEDC-31487. 
Application of this change in APRM flow-biased scram/control rod 
block setpoints and allowable values to Reload 5/Cycle 6 is 
confirmed in General Electric Document No. 23A7219.
    Analysis presented in NEDC-31487 demonstrate that performance in 
the ELLLA region is within design limits for overpressure 
protection, stability, loss-of-coolant, containment, reactor 
internals, flow-induced vibration, and reactor internal pressure 
difference. Impact of ELLLA operation on anticipated transients 
without scram is evaluated in Section 7.6.1.7.2 of the UFSAR. 
Application of ELLLA region extension to Reload 5/Cycle 6 has been 
confirmed in GE Document No. 23A7219.
    Because operation with the APRM flow-biased scram/control rod 
block setpoints and allowable values is within the bases reviewed 
and approved by the NRC in the UFSAR [Updated Final Safety Analysis 
Report], this change does not significantly increase the possibility 
or consequences of an accident previously evaluated.
    B. Transfer of RBM Setpoint Control to the COLR
    The proposed changes would transfer control of the setpoint and 
allowable value for the rod block monitor (RBM) - Upscale rod block 
to the Core Operating Limits Report (COLR). Technical Specification 
6.9.1.9, ``Core Operating Limits Report,'' requires that the 
analytical methods used to determine core operating limits be those 
previously reviewed and approved by the NRC and that the core 
operating limits be determined such that all applicable limits of 
the safety analysis are met.
    The setpoint and allowable value incorporate a controlling value 
which will be specified in the COLR and noted as such by reference 
in the Technical Specifications. Therefore, the setpoint and 
allowable value would continue to be controlled in a manner that 
would ensure that safety analysis limits are met and implementation 
of the proposed changes would not reduce the level of assurance 
provided by the existing Technical Specifications. Based upon the 
above information, we conclude that implementation of the proposed 
change would not significantly increase the probability or 
consequences of an accident previously evaluated.
    C. RCS Recirculation Flow Revisions
    The original analysis used to support operation up to 105% of 
rated core flow is contained in NEDC-31487. NEDC-31487 addresses the 
full range of transient and accident events associated with 
operation up to 105% of rated core flow. The affects of operation 
with the revised RCS recirculation flow upscale trip setpoint and 
allowable value are bounded by the analysis presented in NEDC-31487.
    In addition, cycle specific analysis performed for Reload 5/
Cycle 6, have incorporated the assumption of operation up to 105% of 
rated core flow and have confirmed that operation is within 
allowable design limits.
    Based on the above information, we conclude that the proposed 
change would not significantly increase the probability or 
consequences of an accident previously evaluated.
    2. Would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    A. Changes to APRM Flow-Biased Scram/Control Rod Block
    The proposed changes to the APRM flow-biased scram/control rod 
block setpoints and allowable values would not alter the function of 
the APRM system nor involve any type of 

[[Page 39451]]
plant modification. In addition, operation with the revised APRM flow-
biased scram/control rod block setpoints and allowable values would 
not create any new operating modes, accident scenarios, equipment 
failure modes, or fission product release paths. Based upon the 
above information, we conclude that the proposed changes would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.


    B. Transfer of RBM Setpoint Control to the COLR
    The proposed transfer of control of the RBM setpoint and 
allowable value to the COLR would not alter the function of the RBM 
system nor involve any type of plant modification. In addition, 
operation with the revised setpoint and allowable value would not 
create any new operating modes, accident scenarios, equipment 
failure modes, or fission product release paths. Based upon the 
above information, we conclude that the proposed changes would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    C. RCS Recirculation Flow Revisions
    The proposed changes would not alter the function of the RCS 
recirculation flow upscale trip function nor involve any type of 
plant modification. In addition, operation with the revised RCS 
recirculation flow upscale trip setpoint and allowable value would 
not create any new operating modes, accident scenarios, equipment 
failure modes, or fission product release paths. Based upon the 
above information, we conclude that the proposed changes would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    A. Changes to APRM Flow-Biased Scram/Control Rod Block
    The Bases for Hope Creek Technical Specification 2.2.1 state 
that the APRM setpoints were selected to provide adequate margin for 
the safety limits while allowing operating margins that reduce the 
possibility of unnecessary shutdowns.
    The proposed changes would ensure that these objectives are met. 
The Minimum Critical Power Ratio (MCPR) operating limit specified in 
the Hope Creek COLR was determined using the APRM flow-biased scram/
control rod block setpoints and allowable values proposed in this 
amendment application and has been chosen to ensure that the 
cladding safety limit would not be violated during normal plant 
operations and anticipated transients. Since the operating limit 
MCPR is chosen such that the cladding safety limit is maintained, 
adequate margins for the safety limits are ensured. The proposed 
changes would also serve to ensure that the objective of avoiding 
unnecessary shutdowns is met by furnishing greater margin between 
the operating envelope and the setpoint at lower flows.
    Based on the above information, we conclude that the proposed 
changes would not significantly reduce a margin of safety.
    B. Transfer of RBM Setpoint Control to the COLR
    The proposed transfer of control of the RBM setpoint and 
allowable value to the COLR would not affect the methodology for 
establishing the core operating limits. The setpoint and allowable 
value are modified to incorporate a controlling value which will be 
included in the COLR and indicated as such by reference in the 
Technical Specifications. Therefore, the setpoint and allowable 
value would continue to be controlled in a manner that would ensure 
that safety analysis limits are met. We conclude that implementation 
of the proposed changes would not significantly reduce a margin of 
safety.
    C. RCS Recirculation Flow Revisions
    The HCGS was licensed to operate up to 105% of rated core flow 
as part of Amendment 15. The analysis used to justify operation up 
to 105% of rated core flow is contained in NEDC-31487. NEDC-31487 
addresses the full range of transient and accident events associated 
with operation up to 105% of rated core flow. The affects of 
operation with the revised RCS recirculation flow upscale trip 
setpoint and allowable value are bounded by the analysis presented 
in NEDC-31487.
    In addition, cycle specific analysis performed for Reload 5/
Cycle 6, have incorporated the assumptions of operation up to 105% 
of rated core flow and have confirmed that operation is within 
allowable design limits.
    Based on the above information, we conclude that the proposed 
changes would not significantly reduce a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library,190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of amendment request: January 11, 1995
    Description of amendment request: The proposed Technical 
Specification (TS) revision provides changes to TS Section 3/4.3.8 
``Turbine Overspeed Protection System.'' The proposed revision removes 
these requirements from the TS and relocates the Bases to the Hope 
Creek Updated Final Safety Analysis Report (UFSAR) and the Surveillance 
Requirements to the applicable surveillance procedures. The Limiting 
Conditions for Operation (LCOs) would be eliminated.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve no hardware changes, no changes to 
existing structures, and no changes to the operation of any systems 
or components. Specifically, the deletion of the LCO's by this 
submittal will not alter established turbine overspeed protection 
system operation. Procedural guidance will be provided in the event 
of an inoperable control, stop, or intermediate valve to place the 
system in a safe condition. The relocation of this specification to 
the UFSAR and surveillance procedures will continue to ensure that 
the probability of unacceptable damage to safety-related structures, 
systems, and components from turbine missiles remains acceptably 
low. Relocation of this specification's Bases and Surveillance 
Requirements to the UFSAR and surveillance procedures, respectively, 
and the deletion of the LCO's represents changes that do not affect 
plant safety and do not alter existing accident analyses.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes are procedural in nature concerning the 
location of the descriptive information and surveillance 
requirements for the turbine overspeed protection system. Removing 
these specifications from the Technical Specifications and placing 
them in the UFSAR and surveillance procedures will not alter the 
operation of the turbine overspeed protection system or its ability 
to perform its intended function. Procedural guidance will be 
provided to assist in placing the system in a safe condition while 
maintenance and testing of this system will continue in accordance 
with the turbine manufacturers recommendations taking into 
consideration plant operating experience and ASME guidance. 
Therefore, these changes will not create a new or unevaluated 
accident or operating condition.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes relocate the Turbine Overspeed Protection 
System portion of the Technical Specifications to the UFSAR and 
surveillance procedures in accordance with guidance provided by the 
NRC Final Policy Statement regarding the improvement of Technical 
Specifications. The requirements that will reside in the UFSAR for 
the turbine overspeed protection system will ensure that the system 
remains capable of protecting against excessive turbine overspeed. 
Therefore, the proposed changes will not involve a significant 
reduction in any margins of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this 

[[Page 39452]]
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of amendment request: January 20, 1995
    Description of amendment request: The proposed Technical 
Specification (TS) revision represents changes to TS Section 3/4.11.2.6 
``Explosive Gas Mixture,'' TS Table 3.3.7.11-1 ``Radioactive Gaseous 
Effluent Monitoring Instrumentation,'' and TS Table 4.3.7.11-1 
``Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance 
Requirements.'' The proposed revision would remove these TS from the 
Technical Specifications and relocate the Bases to the Hope Creek 
Updated Final Safety Analysis Report (UFSAR) and the Surveillance 
Requirements to the applicable surveillance procedures. The Limiting 
Conditions for Operation (LCOs) would be eliminated.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve no hardware changes, no changes to 
the operation of any systems or components, and no changes to 
existing structures. The relocation of this specification to the 
UFSAR and surveillance procedures will continue to ensure that the 
entrainment of hydrogen in the main condenser is monitored and 
controlled. Relocation of this specification's Bases and 
Surveillance Requirements to the UFSAR and surveillance procedures, 
respectively, and the deletion of the LCO's represent changes that 
do not affect plant safety and do not alter existing accident 
analyses.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes are procedural in nature concerning the 
location of the descriptive information and surveillance 
requirements for the explosive gas mixture monitoring 
instrumentation. Removing these specifications from the Technical 
Specifications and placing them in the UFSAR and surveillance 
procedures will not alter the operation of the explosive gas 
monitors or their ability to perform intended functions. Maintenance 
and testing of these monitors will continue based upon the 
manufacturers' recommendations taking into consideration plant 
operating experience. Therefore, these changes will not create a new 
or unevaluated accident or operating condition.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes relocate the Explosive Gas Mixture 
specifications from the Technical Specifications to the UFSAR and 
surveillance procedures in accordance with guidance provided by the 
NRC Final Policy Statement regarding the improvement of Technical 
Specifications. The requirements that will reside in the UFSAR and 
surveillance procedures for the explosive gas mixture monitoring 
instrumentation will ensure that the ability to determine main 
condenser hydrogen concentrations is properly maintained. Therefore, 
the proposed changes will not involve a significant reduction in any 
margins of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of amendment request: January 20, 1995
    Description of amendment request: The proposed change to the 
Technical Specifications (TS) would revise TS 4.1.3.1.2.b, ``Control 
Rods - Surveillance Requirement'' to change the required action to be 
taken when a control rod becomes immovable due to excessive friction or 
mechanical interference from ``at least once per'' 24 hours to 
``within'' 24 hours. The other control rods would be tested within 24 
hours and every 7 days thereafter, as opposed to the current 
requirement of testing every 24 hours.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves no hardware changes, no changes to 
the operation of any systems or components, and no changes to 
existing structures. The revision of the control rod movement test 
frequency represents a change that does not affect plant safety and 
does not alter existing accident analyses.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change is procedural in nature concerning the 
frequency of control rod movement tests for all withdrawn control 
rods after a control rod has been determined to be immovable due to 
excessive friction or mechanical interference. The methodology for 
determining additional immovable control rods remain unchanged. The 
proposed change while slightly increasing the possibility of an 
undetected immovable control rod will not create a new or 
unevaluated accident or operating condition.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed change is in accordance with recommendations 
provided by the NRC regarding the improvement of Technical 
Specifications. This change will result in the perpetuation of 
current safety margins while reducing regulatory burden and 
decreasing equipment degradation.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: June 1, 1995
    Description of amendment request: The proposed change would revise 
the Technical Specifications to make them more restrictive regarding 
control rod drive (CRD) scram time testing. CRD scram time testing 
would be required following maintenance prior to considering the CRD 
operable, and could be performed at any reactor 

[[Page 39453]]
pressure. Additional testing would be required when reactor coolant 
pressure is greater than or equal to 950 psig and prior to 40 percent 
rated thermal power.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration which 
is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change provides more stringent requirements for 
operation of the facility. These more stringent requirements do not 
result in operation that will increase the probability of initiating 
an analyzed event and do not alter assumptions relative to 
mitigation of an accident or transient event. The more restrictive 
requirements continue to ensure process variables, structures, 
systems and components are maintained consistent with the safety 
analysis and licensing basis. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in the methods governing normal plant operation. The 
proposed change does impose different requirements. However, these 
changes are consistent with assumptions made in the safety analysis 
and licensing basis. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The imposition of more restrictive requirements either has no 
impact on or increase in the margin of plant safety. As provided in 
the discussion of the change, each change in this category is by 
definition providing additional restrictions to enhance plant 
safety. The change maintains requirements within safety analyses and 
licensing bases. Therefore, this change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: April 10, 1995
    Description of amendment request: The proposed amendment would 
revise the following Technical Specifications (TS) and their associated 
Bases: TS 3/4.7.1.2, ``Auxiliary Feedwater System,'' to clarify Action 
``a'' by inserting ``or both'' steam generators'' and to 
remove references to pressure indicators and specific pressure readings 
and adding performance based requirements; TS 3/4.7.1.3, ``Condensate 
Storage Tanks,'' to modify the Limiting Condition for Operation (LCO) 
to more closely conform to standard TS; and TS 3/4.7.1.7, ``Motor 
Driven Feedwater Pump System,'' to consolidate the requirements of 2 
current surveillance requirements and clarify the operability 
requirements when local manual valves are realigned for testing 
purposes.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that asignificant hazards consideration does not exist because 
operation of the Davis-besse Nuclear Power Station, Unit Number 1, 
in accordance with these changes would:
    a. Not involve a significant increase in the probability of an 
accident previously evaluated because no change is being made to any 
accident initiator. No previous analyzed accident scenario is 
changed, and initiating conditions and assumptions remain as 
previously analyzed. The proposed changes are clarifications and the 
incorporations of the guidance provided by NUREG-1430. Therefore, it 
can be concluded that the proposed changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The proposed changes do 
not alter the source term, containment isolation or allowable 
radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not change the way the plant is operated and, no new or 
different failure modes have been defined for any plant system or 
component important to safety, nor has any limiting single failure 
been identified as a result of the proposed changes. No new or 
different types of failures or accident initiators are introduced by 
the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes are clarifications and the 
incorporations of the guidance provided by NUREG-1430, and continue 
to ensure the availability and capability of the Auxiliary Feedwater 
System, Service Water System and the Motor Driven Feedwater Pump 
System when called upon to perform their functions. The proposed 
changes will not adversely impact any safety analysis assumptions.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: June 1, 1995
    Description of amendment request: The proposed amendment would 
change the allowed outage time from 72 hours to 7 days for one 
unavailable emergency diesel generator (EDG) as detailed in Technical 
Specification 3.8.1.1, ``AC Power Sources, Operating,'' and its 
associated Bases 3.0.5.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Toledo Edison has reviewed the proposed change and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit No. 
1, in accordance with this change would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because the proposed change to 
increase the allowed outage time for one emergency diesel generator 
from three (3) days to seven (7) days does not make a change to any 
accident initiator, initiating condition or assumption. The accident 
previously evaluated in the DBNPS Updated Safety Analysis Report 
(USAR) Section 15.2.9, Loss of All AC Power to the Station 

[[Page 39454]]
Auxiliaries (Station Blackout), is not affected by this proposed 
change. The proposed change does not involve a significant change to 
the plant design or operation, only to the allowed outage time, and 
based on a review of the available alternate A.C. power sources, the 
effect on probabilistic risk at power, the effect on shutdown risk, 
and maintenance planning and scheduling, this change has been 
determined to be acceptable.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed change does not 
invalidate assumptions used in evaluating the radiological 
consequences of an accident, does not alter the source term or 
containment isolation and does not provide a new radiation release 
path or alter potential radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any previously evaluated because the proposed change 
does not introduce a new or different accident initiator or 
introduce a new or different equipment failure mode or mechanism.
    3. Not involve a significant reduction in the margin of safety 
because the proposed change does not significantly reduce the margin 
to safety which exists in the present Technical Specification action 
statements. The DBNPS USAR Section 15.2.9 evaluates the 
acceptability of the loss of all A.C. power to the station, 
including the loss of both EDGs, and the margin of safety in this 
analysis is not affected by the proposed change. in addition, since 
the issuance of the original DBNPS Operating License Technical 
Specifications Toledo Edison has installed a Station Blackout Diesel 
Generator (SBODG), comparable in continuous rating to the EDGs and 
capable of providing emergency A.C. power in the event all three 
offsite 345 kV transmission lines and the two EDGs are unavailable. 
This has positive effect on maintaining the margin to safety which 
exists in the Technical Specifications with a three day allowed 
outage time, which was established prior to installation of the 
SBODG.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: June 7, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.9.4, Refueling Operations - 
Containment Penetrations, and associated Bases 3/4.9.4, Containment 
Penetrations. The proposed changes include revising the Limiting 
Condition for Operation (LCO) 3.9.4.b to allow both doors of the 
containment personnel airlock to be open during core alterations or 
movement of irradiated fuel within the containment, provided that 
certain specified conditions are meet. Additional changes are proposed 
to revise or clarify TS LCO 3.9.4.c, TS Action 3.9.4.a, and TS 
Surveillance Requirement 4.9.4, and modify the Bases to reflect the 
requested changes.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Plant (DBNPS), Unit No. 1, in 
accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no Updated Safety Analysis 
Report (USAR) accident initiators are affected by the proposed 
changes.
    The proposed change to TS LCO 3.9.4.b would allow both doors of 
the containment personnel air lock to be open during core 
alterations or movement of irradiated fuel within the containment, 
provided that certain specified conditions are met. The containment 
personnel air lock is not an initiator to any accident. Whether the 
containment personnel air lock doors are open or closed during fuel 
movement and core alterations has no effect on the probability of 
any accident previously evaluated.
    The proposed clarification of TS LCO 3.9.4.c, changing the term 
``outside atmosphere'' to ``atmosphere outside containment,'' and 
the proposed change to TS LCO 3.9.4.c.1, confirming that, in 
addition to a manual or automatic isolation valve, or a blind 
flange, equivalent means may be used to close a containment 
penetration, have no bearing on the probability of an accident 
previously evaluated.
    The proposed changes to TS Action 3.9.4.a, TS Surveillance 
Requirement (SR) 4.9.4, and TS Bases 3/4.9.4 are administrative 
changes and have no bearing on the probability of an accident 
previously evaluated.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
invalidate accident conditions or assumptions used in evaluating the 
radiological consequences of any accident.
    he analysis results for a fuel handling accident inside 
containment, as presented in Section 15.4.7.3 of the DBNPS USAR, are 
well within the 10 CFR 100 guideline values. Since the analysis does 
not take credit for containment isolation, the status of the 
personnel air lock has no impact on the acceptability of the 
results. In the event of a fuel handling accident, release of 
radioactive material will continue to be minimized since the air 
lock door will remain capable of being closed. Further, the proposed 
change could significantly reduce the dose to workers in the 
containment in the event of a fuel handling accident by speeding the 
containment evacuation process.
    Since an engineering evaluation described in proposed Bases 3/
4.9.4 will ensure that a particular containment penetration closure 
technique is capable of restricting the release of radioactive 
material from a fuel handling accident, the proposed change to TS 
LCO 3.9.4.c.1, confirming that an equivalent means may be used to 
close a containment penetration, has no adverse effect on the 
consequences of an accident previously evaluated.
    The proposed clarification of TS LCO 3.9.4.c, and the proposed 
changes to TS Action 3.9.4.a, TS SR 4.9.4, and TS Bases 3/4,9.4 are 
administrative changes and have no effect on the consequences of an 
accident previously evaluated.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because there are no 
new failure modes or mechanisms associated with the proposed 
changes, nor do the proposed changes involve any modification of 
plant equipment or changes in plant operational limits.
    As described above, the analysis results for a fuel handling 
accident inside containment does not take credit for containment 
isolation. Thus the proposed change to TS LCO 3.9.4.b to allow both 
doors of the containment personnel air lock to be open during core 
alterations or movement of irradiated fuel within the containment 
could affect the release path for radioactive material released 
during a fuel handling accident, however no new or different kind of 
accident will result.
    3. Not involve a significant reduction in the margin of safety.
    The analysis results for a fuel handling accident inside 
containment, as presented in [Section 15.4.7.3 of] the DBNPS USAR, 
are well within the 10 CFR 100 guideline values. Since the analysis 
does not take credit for containment isolation, the status of the 
personnel air lock has no impact on the acceptability of the 
results.
    The proposed change to TS LCO 3.9.4.c.1 regarding the use of 
equivalent means of containment penetration closure has no adverse 
impact on the margin of safety since an equivalent containment 
penetration 

[[Page 39455]]
closure technique will provide the same assurance of containment 
closure during core alterations or movement of irradiated fuel 
inside containment.
    The various administrative changes and clarifications proposed 
will not reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: June 23, 1995
    Description of amendment request: The proposed amendment would 
relocate Technical Specifications (TS) 3/4.3.3.3 - Seismic 
Instrumentation, TS 3/4.3.3.4 - Meteorological Instrumentation, and TS 
3/4.4.11 - Reactor Coolant System Vents and associated Bases.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station, Unit Number 1, 
in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no change is being made to any 
accident initiator. No previous analyzed accident scenario is 
changed, and initiating conditions and assumptions remain as 
previously analyzed.
    The proposed changes are deletions and relocations of 
specifications that do not meet the NRC Final Policy Statement [58 
FR 39132, dated July 22, 1993] criteria for inclusion in TS. 
Furthermore, these relocations and deletions are consistent with the 
NRC guidance for TS provided by the ``Improved Standard Technical 
Specifications for Babcock and Wilcox Plants,'' NUREG-1430, Revision 
0. Therefore, it can be concluded that the proposed changes do not 
involve a significant increase in the probability of an accident 
previously evaluated.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The proposed changes do 
not alter the source term, containment isolation or allowable 
radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not change the way the plant is operated, and no new or 
different failure modes have been defined for any plant system or 
component important to safety, nor has any limiting single failure 
been identified as a result of the proposed changes. No new or 
different types of failures or accident initiators are introduced by 
the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because Seismic Instrumentation, Meteorological Instrumentation, and 
Reactor Coolant System Vents are not inputs in the calculation of 
any safety margin with regard to TS Safety Limits, Limiting Safety 
System Settings, other TS Limiting Conditions for Operation, or 
other previously defined margins for any structure, system, or 
component important to safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: July 14, 1995
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would provide a two-hour allowed outage 
time (AOT) for one residual heat removal (RHR) pump to accommodate 
plant safety and emergency power systems surveillance testing and 
permit depressurizing safety injection (SI) accumulators in lieu of 
accumulator isolation.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    Specifically, operation of the Surry Power Station in accordance 
with the proposed change will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Surveillance and testing requirements are necessary to assure 
that RHR and interfacing systems' reliability is maintained. 
Existing analyses demonstrate that adequate shutdown cooling will be 
maintained with one train of RHR Operable and in service. Analyses 
also demonstrate that alternate shutdown cooling modes remain 
available with adequate decay heat removal capability. Furthermore, 
the opposite train of RHR remains available while in the two hour 
surveillance AOT. The response time and operator actions required to 
place the available RHR train in service are consistent with similar 
operator response times and actions
    required to place alternate shutdown cooling modes in service. 
The administrative controls and procedures in place assure adequate 
shutdown cooling capability is maintained as supported by existing 
analyses.
    The existing safety analyses demonstrate that Reactor Coolant 
System [RCS] integrity will be maintained when SI accumulator 
pressure is below the pressurizer PORV [power operated relief valve] 
LTOPS [low temperature overpressure system] setpoint. Therefore, SI 
accumulator isolation is not required to ensure Reactor Coolant 
System integrity. With RCS temperature below the LTOPS enabling 
temperature, automatic actuation of the pressurizer PORVs or other 
TS specified relief paths ensure the assumed design basis reactor 
vessel beltline flaw will not propagate under design basis low 
temperature overpressurization accident conditions. System design 
and configuration adequately mitigate an LTOPS actuation due to an 
SI accumulator discharge with no negative consequences regarding RCS 
structural integrity or SBLOCA [small break loss-of-coolant 
accidents] concerns.
    Therefore, the proposed Allowed Outage Time for an inoperable 
RHR loop and the ability to depressurize the SI accumulator in lieu 
of SI accumulator isolation do not increase the probability or 
consequence of any previously analyzed accidents.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed two hour AOT for one train of the RHR System will 
preclude the possibility of a Technical Specification violation for 
conditions where a train of RHR is out of service for surveillance 
testing. Calculations by Westinghouse with evaluations and 
supporting analyses performed by Virginia Power, confirm the 
adequacy of decay heat removal with one RHR train in service, and 
multiple alternate shutdown cooling modes remain available. There 
are no plant modifications required by this proposed TS change. 
Further, the proposed change does not invalidate any 

[[Page 39456]]
component design criteria or the assumptions of the UFSAR [updated 
final safety analysis report] accident analyses. The RHR System is 
being operated in a manner consistent with the design basis and 
configuration of the system and is supported by existing analyses 
and procedural controls.
    There are no new failure modes or mechanisms associated with the 
proposed change to allow the depressurizing of a SI accumulator to a 
pressure value below the LTOPS setpoint. The LTOPS enabling 
temperature remains unchanged. No operating limits or setpoints are 
added or deleted by the proposed change. Reactor Coolant System 
pressure relief paths are not affected.
    Therefore, the possibility of a new or different kind of 
accident is not being created by the proposed Allowed Outage Time 
for an inoperable RHR loop and the ability to depressurize the SI 
accumulator in lieu of SI accumulator isolation.
    (3) Involve a significant reduction in margin of safety.
    The proposed Technical Specifications change does not involve a 
reduction in a margin of safety. The existing safety analyses 
demonstrate that adequate shutdown cooling will be maintained when a 
train of RHR is out of service for up to two hours for plant system 
surveillance testing, while the operable train of RHR is operating. 
Supporting analyses determined that the RHR System meets the design 
cooldown requirements for a reactor core rating of 2546 MWth 
[megawatt thermal] with either one or both trains of RHR in service. 
Additionally, an evaluation of the technical basis for shutdown 
operations for the proposed Surry core uprating to 2546 MWth 
determined that the administrative controls and Abnormal Procedures 
in place at Surry ensure adequate decay heat removal capability 
during shutdown conditions. The administrative controls and 
procedure revisions are supported by a detailed series of thermal-
hydraulic calculations for various loss of RHR scenarios. There is 
no reduction in shutdown cooling capability due to the proposed TS 
change, and no reduction in the capability to mitigate a loss of 
decay heat removal event since the RHR train affected by the testing 
is available and can be restored in a comparable time period to that 
required to restore RHR to service in the event of loss of station 
power or loss of the operating train of RHR. Consequently, system 
design, plant configuration, and administrative controls remain 
available to adequately mitigate a loss of RHR event with a single 
train of RHR out of service for up to two hours during plant system 
surveillance testing. It may be concluded that there is no reduction 
in the margin of safety due to the proposed Technical Specification 
change.
    Existing safety analyses also demonstrate that Reactor Coolant 
system integrity will be maintained in the event of an inadvertent 
SI accumulator discharge when SI accumulator pressure is below the 
pressurizer PORV LTOPS setpoint. Sufficient administrative controls 
are maintained to ensure LTOPS is ``Enabled'' and SI accumulators 
are isolated at the appropriate RCS conditions to minimize the 
possibility of challenging RCS integrity. Technical Specifications 
administrative controls that prevent inadvertent charging pump 
operation, maintain adequate relief paths, and restrict Steam 
Generator primary to secondary temperature differential remain in 
place. Consequently, the Technical Specifications change ensures 
that an inadvertent SI accumulator discharge cannot challenge RCS 
structural integrity during LTOPS conditions when SI accumulator 
pressure is below the pressurizer PORV LTOPS setpoint.
    Therefore, the proposed Allowed Outage Time for an inoperable 
RHR loop and the ability to depressurize the SI accumulator in lieu 
of SI accumulator isolation does not reduce any margin of safety as 
defined in the Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: David B. Matthews

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: June 14, 1995, as supplemented by letter 
dated July 13, 1995.
    Description of amendment request: This amendment request proposes 
to revise Technical Specification (TS) 3.2.3, ``Nuclear Enthalpy Rise 
Hot Channel Factor,'' TS 6.9.1.9, ``Core Operating Limits Report,'' and 
the associated Bases sections. The revisions are needed to incorporate 
changes associated with the planned implementation of advanced nuclear 
and core thermal-hydraulic design methodologies licensed from 
Westinghouse Electric Corporation for core reload design, starting with 
Cycle 9.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The probability of occurrence and the consequences of an 
accident evaluated previously in the Updated Safety Analysis Report 
(USAR) are not increased due to the proposed technical specification 
changes. The Technical Specification changes being requested are to 
reflect revised calculational methods to be used for core reload 
design, starting with Cycle 9. There are no changes being made to 
any licensed design parameters from previous cycles. Thus, it is 
concluded that the probability and consequences of the accidents 
previously evaluated in the USAR are not increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There is no new type of accident or malfunction being created. 
The proposed changes only provide revised analysis methodologies to 
support core reload design, starting with Cycle 9. The requested 
changes do not change the method and manner of plant operation. The 
safety design bases in the USAR have not been altered. Thus, the 
requested changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not change the plant configuration in a 
way that introduces a new potential hazard to the plant and do not 
involve a significant reduction in the margin of safety. The 
analyses and evaluations discussed in the safety evaluation 
(Attachment I) [Attached to Wolf Creek Nuclear Operating 
Corporation's letter number ET 95-0051, dated June 14, 1995] 
demonstrates that all applicable design criteria continue to be met 
for the changes. Therefore, it is concluded that the margin of 
safety, as described in the bases to any technical specification, is 
not reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the 

[[Page 39457]]
same as above. They were published as individual notices either because 
time did not allow the Commission to wait for this biweekly notice or 
because the action involved exigent circumstances. They are repeated 
here because the biweekly notice lists all amendments issued or 
proposed to be issued involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: February 23, 1995
    Description of amendment request: The amendment relates to 
Commonwealth Edison Company's (ComEd) request to reflect the merger 
between IIGEC, MidAmerican, Midwest Power Systems Inc., and Midwest 
Resources, Inc. By letter dated November 21, 1994, Iowa-Illinois Gas 
and Electric Company (IIGEC) requested approval, pursuant to Section 
50.80 of Title 10 of the Code of Federal Regulations, of the transfer 
of its ownership share of 25 percent of Quad Cities Nuclear Power 
Station, Units 1 and 2, to MidAmerican Energy Company (MidAmerican).
    Date of publication of individual notice in Federal Register: July 
5, 1995 (60 FR 35054)
    Expiration date of individual notice: August 4, 1995
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: March 31, 1995
    Description of amendment request: The proposed amendment will 
delete Technical Specification (TS) Sections 1.38 and 1.39, 
``Definitions, Fuel Assembly Types,'' revise TS Sections 3/4.9.3, 
``Refueling Operations, Decay Time'' and TS 3/4.9.14, ``Refueling 
Operations, Spent Fuel Pool - Reactivity Condition,'' replace TS 
Sections 5.6.1.1, ``Spent Fuel,'' and TS 5.6.3, ``Capacity,'' and add a 
new TS Section 3/4.9.15, ``Refueling Operations, Spent Fuel Pool 
Cooling.'' These changes would support a rerack of the spent fuel pool 
to expand the spent fuel pool's storage capacity from 1168 assemblies 
to 1480 assemblies so as to accommodate a full-core-discharge through 
the current validity date of the Haddam Neck operating license (2007).
    Date of publication of individual notice in Federal Register: May 
12, 1995 (60 FR 25746)
    Expiration date of individual notice: June 12, 1995
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station Units 1 and 2, Lake County, Illinois

    Date of application for amendments: June 14, 1995
    Brief description of amendments: The amendments revise the 
requirement to perform an emergency diesel generator (EDG) automatic 
start and sequence loading test immediately following the 24 hour EDG 
endurance test.
    Date of issuance: July 18, 1995
    Effective date: July 18, 1995
    Amendment Nos.: 166 and 154
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.Public comments requested as to 
proposed no significant hazards consideration determination: Yes (60 FR 
34308). This notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice also provided 
for an opportunity to request a hearing by July 31, 1995, but indicated 
that if the Commission makes a final no significant hazards 
consideration determination any such hearing would take place after 
issuance of the amendment. The Commission's related evaluation of the 
amendments, finding of exigent circumstances and final no significant 
hazards consideration determination is contained in a Safety Evaluation 
dated July 18, 1995.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: March 31, 1995
    Brief description of amendments: The amendments revise Technical 
Specification section 3.9.4 to allow, under certain conditions, both 
containment personnel airlocks to be open during core alterations.
    Date of issuance: July 12, 1995
    Effective date: July 12, 1995
    Amendment Nos.: 197 and 182
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29879)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 12, 1995.No significant 
hazards consideration comments received: No.

[[Page 39458]]

    Local Public Document Room location:  Maud Preston Palenske 
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station,Nemaha County, Nebraska

    Date of amendment request: May 2, 1995
    Brief description of amendment: The amendment revised Surveillance 
Requirement 4.7.A.2.f.1 to allow a one-time extension for the 
performance of Type B local leak rate testing of the drywell head and 
manport from July 17, 1995, until startup from Refueling Outage 16, 
scheduled to commence on October 13, 1995.
    Date of issuance: July 11, 1995
    Effective date: July 11, 1995
    Amendment No.: 170
    Facility Operating License No. DPR-46. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29879)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 11, 1995No significant hazards 
consideration comments received: No.
    Local Public Document Room location:  Auburn Public Library, 118 
15th Street, Auburn, NE 68305.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
MillstoneNuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: January 10, 1995
    Brief description of amendment: The amendment revises the Technical 
Specifications to delete the power range negative flux trip from Tables 
2.2-1, 3.3-1, and 4.3-1, and delete the associated Bases Section 2.0.
    Date of issuance: July 11, 1995
    Effective date: As of the date of issuance to be implemented 
within30 days.
    Amendment No.: 116
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11135)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 11, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
MillstoneNuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment:  March 29, 1995
    Brief description of amendment: The amendment revises Technical 
Specification 3.10.5 to allow more than one control bank to be fully 
withdrawn from the core simultaneously in order to conduct rod drop 
time response testing.
    Date of issuance: July 11, 1995
    Effective date: As of the date of issuance to be implemented 
within60 days.
    Amendment No.: 117
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29880) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 11, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

Northern States Power Company, Docket No. 50-263, Monticello 
NuclearGenerating Plant, Wright County, Minnesota

    Date of application for amendment:  February 12, 1993, as 
supplemented by letters dated March 22, 1993, and August 25, 1994
    Brief description of amendment: The amendment increases the minimum 
core spray pump flow to more conservatively account for emergency core 
cooling systems bypass leakage paths. The amendment also makes various 
typographical, editorial and administrative corrections and changes.
    Date of issuance: July 12, 1995
    Effective date: July 12, 1995
    Amendment No.: 93
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41508). The August 25, 1994 letter provided clarifying information 
within the scope of the original submittal and did not change the 
staff's initial proposed no significant hazards considerations 
determination.The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 12, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station,Unit No. 1, Washington County, Nebraska

    Date of amendment request: November 11, 1994, as supplemented by 
letters dated April 7, 1995, and June 26, 1995
    Brief description of amendment: The amendment implements 
administrative changes to TS 5.2 and 5.5. These changes reflect 
organizational changes in OPPD senior management, delete specific 
titles of personnel on the Plant Review Committee (PRC), revise the 
makeup of the PRC quorum, revise the membership of the Senior Audit and 
Review Committee (SARC), delete SARC audit frequencies and add minor 
clarifications to the descriptions of SARC reviews and audits.
    Date of issuance: July 21, 1995
    Effective date: July 21, 1995
    Amendment No.: 168
    Facility Operating License No. DPR-40. Amendment revised the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65819). The April 7, 1995, and June 26, 1995, letters provided 
clarifying information and did not change the initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 21, 1995.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station,Unit No. 1, Washington County, Nebraska

    Date of amendment request: March 1, 1995
    Brief description of amendment: The amendment revises TS 2.5, 2.8, 
2.11, 3.2, and 3.10 and relocates administrative controls for the 
emergency and security plans from TS 5.5 and 5.8 to the plans. The 
relocation is in accordance with Generic Letter (GL) 93-07, 
``Modification of the Technical Specification Administrative Control 
Requirements for Emergency and Security Plans.''
    Date of issuance: July 21, 1995
    Effective date: July 21, 1995
    Amendment No.: 169
    Facility Operating License No. DPR-40: The amendment revised the 
TechnicalSpecifications.

[[Page 39459]]

    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18627) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 21, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay 
Power Plant, Unit 3, Humboldt County, California

    Date of application for amendment: April 10, 1995
    Brief description of amendment: This amendment revised License No. 
DPR-7, to permit the provisions of 10 CFR 50.59 to be applied with 
respect to changes to the facility or procedures described in the 
Decommissioning Plan or changes to the Decommissioning Plan, and the 
conduct of tests or experiments not described in the Decommissioning 
Plan.
    Date of issuance: July 7, 1995
    Effective date: This license amendment is effective as of the date 
of its issuance and must be fully implemented no later than 30 days 
from the date of issuance.
    Amendment No.: 29Facility License No. DPR-7: This amendment revised 
License No. DPR-7
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29885)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 7, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Humboldt County Library, 636 F 
Street, Eureka, California 95501.

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit 
No. 3, York County, Pennsylvania

    Date of application for amendment: November 21, 1994
    Brief description of amendment: This amendment changes the 
technical specifications (TS) by allowing the third Type A Containment 
Integrated Leakage Rate Test in the second 10-year service period to be 
conducted during refueling outage 11 scheduled for September 1997. This 
TS change is consistent with a one-time exemption from Appendix J to 10 
CFR Part 50 that extends the 10-year service period and allows the 
three type A tests to be performed at intervals that are not 
approximately equal.
    Date of issuance: July 10, 1995Effective date: July 10, 1995
    Amendment No.: 210
    Facility Operating License No. DPR-56: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27340)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 10, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas 
Company,Delmarva Power and Light Company, and Atlantic City 
Electric Company,Docket No. 50-278, Peach Bottom Atomic Power 
Station,Unit No. 3, York County, Pennsylvania

    Date of application for amendment: June 23, 1993, as supplemented 
by letters dated April 5, May 2, June 6, June 8, July 6 (two letters), 
July 7, July 20, July 28 (two letters), September 16, September 30, and 
October 14, 1994 and June 22, 1995.
    Brief description of amendment: The amendment raises the authorized 
maximum power level from 3293 MWt to a new limit of 3458 MWt. The 
amendment also approves changes to the Technical Specifications to 
implement operation at the increased power limit.
    Date of issuance: July 18, 1995
    Effective date: As of date of issuance and is to be implemented 
prior to startup in Cycle 11, currently scheduled for October 1995.
    Amendment No.: 211
    Facility Operating License No. DPR-56: Amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: August 29, 1994 (59 FR 
44432)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 18, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: April 10, 1995
    Brief description of amendments: Remove the response time limit 
Tables 3.3.1-2, 3.3.2-3, and 3.3.3-3 from the Technical Specifications, 
and add the information to the Final Safety Analysis Report in 
accordance with Generic Letter 93-08.
    Date of issuance: July 11, 1995
    Effective date: July 11, 1995
    Amendment Nos.: 148 and 118
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60FR 
29887)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 11, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of application for amendments: November 21, 1994, as 
supplemented April 6, and July 3, 1995
    Brief description of amendments: These amendments make changes 
affecting the Administrative Controls Section of the Technical 
Specifications. The areas changed are Nuclear Effectiveness and 
Efficiency Design Study (NEEDS) Organization Title Changes; Minimum 
Shift Crew Composition; delete Independent Technical Review Section 
from TS; delete Nuclear Review Board (NRB) Review Section from TS; and 
delete NRB Audit Section from TS.
    Date of issuance: July 18, 1995
    Effective date: Units 1 and 2, as of the date of issuance and shall 
be implemented within 30 days.
    Amendment Nos.: 96 and 60
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24914)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 18, 1995.No significant 
hazards consideration comments received: No

[[Page 39460]]

    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of application for amendments: August 31, 1994, as 
supplemented July 3, 1995
    Brief description of amendments: These amendments modify TS 
Sections 3.4.9.1, 3.4.9.2, 3.9.11.1, 3.9.11.2, and the associated Bases 
Sections 3/4.4.9 and 3/4.4.11, to permit the use of either an 
``analytical approach'' (i.e., calculation) or ``demonstrations'' to 
ensure the operability of an alternate decay heat removal method, 
rather than the existing TS requirement which stipulates that 
operability of the alternate decay removal method be demonstrated.
    Date of issuance: July 18, 1995
    Effective date: Units 1 and 2, as of the date of issuance and shall 
be implemented within 30 days.
    Amendment Nos.: 97 and 61
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55884)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 18, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of application for amendments:  August 22, 1994, as 
supplemented July 3, 1995
    Brief description of amendments: These amendments revise Technical 
Specification Surveillance Requirement 4.1.3.1.4a to delete the 
requirement that the Scram Discharge Volume (SDV) be determined 
operable by testing the SDV vent and drain valves from a configuration 
of less than or equal to 50% rod density.
    Date of issuance:  July 18, 1995
    Effective date: Units 1 and 2, effective as of date of issuance and 
shall be implemented within 30 days.
    Amendment Nos.: 98 and 62
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55881)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 18, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of application for amendments:  August 22, 1994, as 
supplemented by letter dated July 3, 1995
    Brief description of amendments: The amendments revise the 
Technical Specifications surveillance requirements for scram insertion 
times and revise the TS surveillance requirements for control rod block 
and source range monitoring instrumentation.
    Date of issuance: July 18, 1995
    Effective date:  Units 1 and 2, effective as of the date of 
issuance and shall be implemented within 30 days.
    Amendment Nos.: 99 and 63
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55881)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 18, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Public Service Company of Colorado, Docket No. 50-267, Fort St. 
Vrain Nuclear Generating Station (FSV), Unit No. 1, Platteville, 
Colorado

    Date of application for amendment:  Amendment No. 88, April 14, 
1995.
    Brief description of amendment: This amendment would revise the FSV 
Decommissioning Technical Specifications (DTS) by: revising the FSV DTS 
to reflect recent organizational changes resulting from corporate 
restructuring to prepare for repowering the site with natural gas-power 
turbines and to incorporate editorial changes. The staff has determined 
that the proposed amendment does not require a significant hazard 
consideration, pursuant to 10 CFR 50.92.Possession-Only License No. 
DPR-34: Amendment revises the DTS.
    Local Public Document Room location:  Weld Library District - 
Downtown Branch, 919 7th Street, Greeley, CO 80631.

Sacramento Municipal Utility District, Docket No. 312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of application for amendment: February 28, 1995
    Brief description of amendment: This amendment relocates the 
quality assurance audit frequencies from the technical specifications 
to the Rancho Seco Quality Manual and changes the reporting frequency 
of the Radioactive Effluent Release Report from semi-annual to annual.
    Date of issuance: July 19, 1995
    Effective date: July 19, 1995
    Amendment No.: 122
    Facility Operating License No. NPF-1: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16200)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 19, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Central Library, Government 
Documents, 828 I Street, Sacramento, California 95814.

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of application for amendments: September 15, 1993, as 
supplemented by letter dated September 6, 1994.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) Table 2.2-1, ``Reactor Protective Instrumentation 
Trip Setpoint Limits,'' Table 3.3-1, ``Reactor Protective 
Instrumentation,'' Table 3.3-3, ``Engineered Safety Feature Actuation 
System Instrumentation,'' and Table 3.3-4, ``Engineered Safety Feature 
Actuation System Instrumentation Trip Values,'' and the associated 
Bases. The revisions to the notes in these tables change the pressure 
at which the low pressurizer pressure trip bypass shall be 
automatically removed to a consistent value of ``before pressurizer 
pressure exceeds 500 psia (the corresponding bistable allowable value 
is less than or equal to 472 psia).'' In addition, the wording of the 
notes is revised to make the notes more consistent with each other.
    Date of issuance: July 14, 1995

[[Page 39461]]

    Effective date: July 14, 1995, to be implemented within 30 days of 
the date of issuance.
    Amendment Nos.: Unit 2 - Amendment No. 120; Unit 3 - Amendment No. 
109
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50975). The September 6, 1994, supplemental letter provided 
additional clarifying information and did not change the initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated July 14, 1995. No significant hazards consideration 
comments received: No
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557,Irvine, California 92713.

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of application for amendments: September 3, 1992
    Brief description of amendments: These amendments revise TS 3/4.4.8 
``Pressure/Temperature Limits - Reactor Coolant System,'' and their 
associated Bases, following NRC guidance provided in Generic Letter 91-
01, ``Removal of the Schedule for Withdrawal of Reactor Vessel Material 
Specimens from Technical Specifications.'' This generic letter allows 
licensees to remove the reactor vessel material surveillance capsule 
withdrawal schedules from the TS because they are a duplication of the 
requirements of 10 CFR Part 50 Appendix H.
    Date of issuance: July 17, 1995
    Effective date: July 17, 1995, to be implemented within 30 days of 
issuance
    Amendment Nos.: Unit 2 - Amendment No. 121; Unit 3 - Amendment No. 
110
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 17, 1993 (58 
FR 8781)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 17, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of application for amendments: April 30, 1993, as supplemented 
by letters dated July 6, 1994 (separate letters for each unit), and 
letter dated January 27, 1995.
    Brief description of amendments: These amendments revise TS 3/
4.4.8.1, ``Pressure-Temperature Limits,'' TS 3.4.8.3.1, ``Overpressure 
Protection Systems-RCS Temperature less than or equal to  deg.F [for 
Unit 2, less than or equal to 246 deg.F for Unit 3],'' TS 3.4.8.3.2, 
``Overpressure Protection Systems-RCS Temperature 256 deg.F 
[for Unit 2, 246 deg.F for Unit 3],'' and the associated TS 
Bases. The proposed change (1) revises the reactor coolant system (RCS) 
pressure-temperature (P-T) limits and the low temperature overpressure 
protection (LTOP) enable temperatures to be effective until 20 
effective full power years (EFPY) of operation and (2) makes minor 
editorial changes.
    Date of issuance: July 18, 1995
    Effective date: July 18, 1995, to be implemented within 30 days of 
issuance
    Amendment Nos.: Unit 2 - Amendment No. 122; Unit 3 - Amendment No. 
111
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: Unit 2 - July 7, 1993 
(58 FR 36445); Unit 3 - June 23, 1993 (58 FR 34094). The two 
supplemental letters dated July 6, 1994, and the January 27, 1995, 
supplemental letter provided clarifying information and did not change 
the initial no significant hazards consideration determination.The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated July 18, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: March 30, 1994
    Brief description of amendment: The amendments implement an analog 
transmitter/trip system on BFN Unit 3, revise the reactor vessel water 
level safety limit and limiting safety system setting for BFN Units 1 
and 3, add instrument identifiers and revise calibration frequencies 
and functional test requirements for BFN Unit 2, revise the calibration 
frequency for instrumentation actuating the suppression chamber-reactor 
building vacuum breakers, and provide editorial changes.
    Date of issuance: July 17, 1995
    Effective date: July 17, 1995
    Amendment Nos.: 222, 237, 196
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994(59 
FR 49435) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 17, 1995.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 356114.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of application for amendment: January 30, 1995
    Brief description of amendment: The proposed amendment revises 
reactor coolant system pressure-temperature curves, changes bases for 
Technical Specification 3/4.4.9, Pressure Temperature Limits, and 
revises License Condition 2.C(3)(d) to reflect a change from 10 
effective full power years (EFPY) to 21 EFPY.
    Date of issuance: July 20, 1995
    Effective date: July 20, 1995
    Amendment No.: 199
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14029)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 20, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: September 8, 1994
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) 4.2.2.2, 4.2.2.4, and 6.9.19. The changes address 

[[Page 39462]]
incorporating a penalty in the Core Operating Limits Report (COLR) to 
account for heat flux (FQ) increases greater than 2 percent 
between measurements.
    Date of issuance: July 20, 1995
    Effective date: July 20, 1995
    Amendment No.: 101
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specification Surveillance Requirements and Administrative 
Controls.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65823). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 20, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of application for amendments:  November 22, 1994
    Brief description of amendments: The amendments revised the 
Technical Specifications to delete unnecessary descriptive phrases 
regarding the number of cells in the station and emergency diesel 
generator batteries.
    Date of issuance: July 11, 1995
    Effective date: July 11, 1995
    Amendment Nos.: 201 and 201
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18630)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 11, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Dated at Rockville, Maryland, this 2nd day of August, 1995.
    For the Nuclear Regulatory Commission
Jack W. Roe, 4Director, Division of Reactor Projects - III/IV, Office 
of Nuclear Reactor Regulation
[Doc. 95-18810 Filed 8-1-95; 8:45 am]
BILLING CODE 7590-01-F