[Federal Register Volume 62, Number 117 (Wednesday, June 18, 1997)]
[Notices]
[Pages 33117-33142]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-15827]


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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 May 23, 
1997, through June 6, 1997. The last biweekly notice was published on 
June 4, 1997 (62 FR 30629).

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

[[Page 33118]]

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 Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, 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 July 18, 1997, 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 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 20555-0001, 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. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, 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.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of amendment request: May 16, 1997

    Description of amendment request: The modification involves 
replacing the service water (SRW) heat exchangers with new plate and 
frame heat exchangers having increased thermal performance capability. 
The saltwater (SW) and SRW piping configuration will be modified as 
necessary to allow proper fit-up to the new components. A flow control 
scheme to throttle saltwater flow to the heat exchangers and the 
associated bypass lines will be added.

[[Page 33119]]

Saltwater strainers with an automatic flushing arrangement will be 
added upstream of each heat exchanger. The majority of the physical 
work associated with this modification is restricted to the SRW pump 
room.
    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 significant increase in the probability or 
consequences of an accident previously evaluated.
    None of the systems associated with the proposed modification 
are accident initiators. The SW and SRW Systems are used to mitigate 
the effects of accidents analyzed in the UFSAR [Updated Final Safety 
Analysis Report]. The SW and SRW Systems provide cooling to safety-
related equipment following an accident. They support accident 
mitigation functions; therefore, the proposed modification does not 
increase the probability of an accident previously evaluated.
    The proposed modification will increase the heat removal 
capacity of the SRW System. The design provided under this activity 
ensures that the safety features provided by the SW and SRW are 
maintained, and in some instances enhanced; i.e., the availability 
of important-to-safety equipment required to mitigate the 
radiological consequences of an accident described in the UFSAR is 
enhanced by the flexibility and increased thermal margin provided 
with this design.
    The redundant cooling capacity of the SW and SRW Systems have 
not been altered. Furthermore, the proposed activity will not 
change, degrade, or prevent actions described or assumed in any 
accident described in the UFSAR. The proposed activity will not 
alter any assumptions previously made in evaluating the radiological 
consequences of any accident described in the UFSAR. Therefore, the 
consequences of an accident previously evaluated in the UFSAR have 
not increased.
    Therefore, the proposed modification does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed activity involves modifying the SW and SRW System 
components necessary to support the installation of new SRW heat 
exchangers. None of the systems associated with this modification 
are identified as accident initiators in the UFSAR. The SW and SRW 
Systems are used to mitigate the effects of accidents analyzed in 
the UFSAR. None of the functions required of the SRW or SW System 
have been changed by this modification. This activity does not 
modify any system, structure, or component such that it could become 
accident initiator, as opposed to its current role as an accident 
mitigator.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The safety design basis for the SW and SRW Systems is the 
availability of sufficient cooling capacity to ensure continued 
operation of equipment during normal and accident conditions. The 
redundant cooling capacity of these systems, assuming a single 
failure, is consistent with assumptions used in the accident 
analysis.
    The design, procurement, installation, and testing of the 
equipment associated with the proposed modification are consistent 
with the applicable codes and standards governing the original 
systems, structures, and components. The design of instruments and 
associated cabling ensures that physical and electrical separation 
of the two subsystems is maintained. Common-mode failure is not 
introduced by this activity. The equipment is qualified for the 
service conditions stipulated for that environment. New cable and 
raceways for this design will be installed in accordance with 
seismic design requirements. The additional electrical load has been 
reviewed to ensure the load limits for the vital 1E buses are not 
exceeded. The circuits and components related to the control valves 
control loops are safety-related, and are similar to those used for 
the other safety-related flow control functions. The proposed 
modification will not have any adverse effects on the safety-related 
functions of the SW and SRW Systems.
    For the above reasons, the existing safety bases have not been 
altered by the proposed modification. This activity will not reduce 
the margin of safety as it exists now. In fact, the margin of safety 
has been increased by this activity due to the increase in the 
thermal capacity of the dual train design and the increased 
availability of safety-related components.
    Therefore, this proposed modification 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 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: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Alexander W. Dromerick, Acting Director.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina
Date of amendment request: April 23, 1997

    Description of amendment request: The proposed changes would revise 
surveillances 4.3.2.1.1.a, 4.3.2.1.4.b, 4.3.2.1.6.g, 4.3.2.1.10a, 
4.3.2.1.10.b, and 4.7.3.b.3 to provide enhanced descriptions of the 
tests being performed and the tested components.
    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:

    This change clarification does not involve a significant hazards 
consideration for the following reasons:
    (1) The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The components affected by the proposed changes are not 
initiators of any accident previously evaluated. The proposed 
changes to specification 4.3.2.1 items affect only the description 
of the testing and make no changes in actual operation or testing. 
The sample heat exchanger valves isolate on receipt of a Safety 
Injection signal and that feature is unaffected by the additional 
testing in the proposed change. Therefore, there is no increase in 
the probability or consequence of a previously analyzed accident.
    (2) The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the surveillance frequencies do not 
involve physical alterations or additions to plant equipment or 
alter the manner in which safety-related systems function or are 
normally operated. The additional testing proposed for the sample 
heat exchanger valves demonstrates the proper operation of a design 
feature but does not operate the valve in any new way. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes to specification 4.3.2.1 clarify existing 
testing. The additional testing for the CCW [component cooling 
water] surge tank level instrumentation adds two components to the 
surveillance documentation. 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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

[[Page 33120]]

    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Mark Reinhart, Acting.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 14, 1997

    Description of amendment request: The proposed amendments would 
revise TS 3/4.3.8, ``Feedwater/Main Turbine Trip System Actuation 
Instrumentation'' by changing the minimum channels required from 3 to 
4. This change reflects a modification that is being installed to 
correct a design deficiency that could have resulted in a failure to 
trip the feedwater pumps and main turbine on high water level due to 
the loss of one of the two instrument lines. The modification adds an 
auxiliary contact to the trip system logic resulting in an additional 
channel. The licensee is also proposing to modify the TS action 
statements for inoperable channels to be similar to TS 3.3.1, ``Reactor 
Protection System.''
    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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposed Technical Specification (TS) change will resolve 
the common instrument line failure (break) from preventing reactor 
high water level trip of Feedwater Pumps and Main Turbine. It will 
not change the probability of occurrence of any accidents, because 
this instrumentation is not an accident initiator. This 
instrumentation resolves a potential concern regarding the results 
of an instrument line break in conjunction with a Feedwater 
Controller Failure Maximum Demand, which has been postulated and 
analyzed separately, but are not required to be analyzed in 
combination, as is described in Chapter 15 of the LaSalle UFSAR. 
There will not be any increase in probability of feedwater transient 
(postulated feedwater controller failure with assumed simultaneous 
failure of one high level trip channel of Feedwater/Main Turbine 
Trip Actuation Instrumentation), nor an instrument line break. The 
design change associated with this TS change will prevent the 
failure of the level 8 trip of Feedwater Pumps and Main Turbine due 
to loss of common variable water leg of level instrument channels 
``B'' and ``C''. Thus there is a slight increase [in] the 
reliability of the high level trip by assuring that a single 
instrument failure, including a failure of a sensing line, will not 
prevent a level 8 trip. The Feedwater/Main Turbine Trip on Reactor 
Vessel Water Level-High, Level 8, mitigates the consequences of the 
transient, Feedwater Controller Failure Maximum Demand, due to the 
main turbine trip with subsequent Turbine Stop Valve closure scram 
and Reactor Recirculation Pump Trip. This limits the neutron flux 
peak and fuel thermal transients so that no fuel damage occurs. MCPR 
remains at or above the operating limit and peak centerline fuel 
temperature increase is small. The consequences of an accident will 
not increase, because the redundancy of the instrumentation portion 
of the Trip Function is somewhat increased.
    TS 3.3.8 limiting Condition for Operation (LCO) Actions b and c 
are proposed to be changed to be similar to the LCO for TS 3.3.1, 
Reactor Protection System Action b.1 to assure trip capability, 
while being consistent with the allowed outage times of current TS 
3.3.8. Also, the proposed action statements and allowed outage times 
are consistent with LCO 3.3.2.2, ``Feedwater and Main Turbine High 
Water Level Trip Instrumentation'', of NUREG 1433, Revision 1, 
Standard Technical Specifications, General Electric Plants, BWR4, 
dated April 1995. The limit on continued plant operation of 72 hours 
in current Action c.1, is overly restrictive, since with one 
inoperable channel tripped and one Operable channel, the Trip 
Function is restored to the same status as current Action b.1 (one 
more instrument failure will cause a failure to actuate on high 
reactor water level). Therefore, although the proposed Actions are 
increasing the allowed outage time for the case with only one 
remaining Operable channel, from 72 hours to 7 days, the level of 
protection for automatic trip capability is maintained except for a 
2 hour period during which trip capability may not exist. In 
addition, like current Action b.1, the proposed Actions assure that 
the longest time that automatic trip capability failure due to 
another instrument failure will exist is 7 days. Therefore, the 
potential for failure of the Feedwater/Main Turbine trip on reactor 
vessel high water level may be slightly increased, but is not 
significant considering the non-safety-related Feedwater Pump and 
Main Turbine trips are not and are not required to be single-failure 
proof.
    Based on the above, the proposed amendments will not increase 
the probability or consequences of any accident previously 
evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The Feedwater/Main Turbine trip is a non-safety function in the 
non-safety-related feedwater system. The high water level trip is an 
equipment protective action preventing main steam carry over in the 
main steam from damaging the main turbine and preventing high 
pressure liquid discharge through the safety relief valve discharge 
lines in case of a feedwater transient due to a controller failure 
to maximum demand. The trip system is not designed to any applicable 
standards or regulatory guides or 10CFR50 Appendix A General Design 
Criteria per UFSAR Table 7.1-2. The trip system is not designed nor 
required to meet the single failure criteria. This is a non-safety/
non-divisional trip actuation required in Operating Condition 1, Run 
Mode, such that high integrity of the trip is maintained. The 
feedwater system is not required to mitigate the consequences of 
accidents.
    The design change associated with this TS change will increase 
the reliability of the trip logic. This is accomplished by assuring 
that a failure of a sensing line will not prevent or cause a level 8 
trip. The failure of Feedwater/Main Turbine channel ``C'' trip 
channel will not have any impact on the RCIC system nor Feedwater/
Main Turbine channels ``A'' & ``B'', because the added signal is 
isolated by a safety-related relay. The 2 out of 3 logic for the 
trip is maintained.
    In addition, the changes to the action statements of the 
specification do not allow a condition that could cause the 
actuation instrumentation to fail in a different manner.
    Based on the above, the proposed change will not create the 
possibility of a new or different kind [of accident] from any 
accident or transient previously evaluated.
    (3) Involve a significant reduction in the margin of safety 
because:
    The proposed TS change will not prevent tripping of Feedwater/
Main Turbine or cause false trips. The existing 2 out of 3 logic 
trip is maintained and does not affect existing failure modes or 
introduce new failure modes. This change will prevent failure of 
level 8 trip of Feedwater Pumps and Main Turbine upon loss of common 
variable water leg for Reactor Vessel Water Level-High, Level 8, 
instrument channels ``B'' & ``C'' and will slightly increase 
reliability of the trip logic. Failure of the non-safety-related 
trip logic will not impact any safety-related system, structure, or 
component.
    The changes to the TS LCO Action statements is consistent with 
the existing actions, while minimizing the time that automatic trip 
capability is not maintained. The change from 72 hours allowed 
operation with one channel Operable and only one channel tripped to 
7 days is consistent with the current allowed outage time for only 
one channel inoperable and not tripped, so any change to the margin 
of safety provided by the current action requirements is minor.
    Based on the above, the proposed TS 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.


[[Page 33121]]


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: May 1, 1997

    Description of amendment request: This request changes Technical 
Specification (TS) Surveillance Requirement (SR) 4.9.A.8.b by 
clarifying the load value for the emergency diesel generator to be 
equal to or greater than the largest single load and revise the 
frequency and voltage requirements during performance of the test.
    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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    The proposed changes represent a clarification of the intent of 
the performance of the largest single emergency load rejection 
surveillance for the diesel generator. These changes allow for 
simulated testing that will more closely duplicate actual emergency 
loading conditions. By removing the specific load value requirement 
from the surveillance, the test can be performed using the actual 
largest load in the same plant configuration that would exist during 
an actual accident scenario. Verification of the steady-state 
voltage and frequency within the required time limits provides 
confidence that the diesel generator can successfully recover from 
this transient. This provides greater assurance that the diesel 
generator is capable of performing its intended design function 
during an accident and the subsequent recovery. The changes to the 
surveillance requirement will not significantly increase the 
consequences of an accident previously evaluated.
    The diesel generator's design function is to mitigate the 
consequences of an accident by providing an independent onsite 
source of alternate AC power with the capacity for operation of 
systems required to shutdown the reactor and maintain it in a safe 
shutdown condition until offsite power is restored. The diesel 
generator and its associated subsystems are not assumed in any 
safety analysis to initiate any accident sequence for Quad Cities 
Station; therefore, the probability of an accident previously 
evaluated is not increased by the proposed amendment.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The proposed changes do not create the possibility of a new or 
different kind of accident previously evaluated for Quad Cities 
Station. The changes revise the largest single emergency load 
rejection surveillance test acceptance criteria for the diesel 
generator. This load rejection transient for the diesel generator is 
bounded by a previously performed accident analysis. This analysis 
assumes the loss of one diesel generator due to loss of 125 VDC 
control power for the duration of a LOCA combined with a LOOP. The 
diesel generator's design function is to mitigate the consequences 
of an accident by providing an independent onsite source of 
alternate AC power with the capacity for operation of systems 
required to shutdown the reactor and maintain it in a safe shutdown 
condition until offsite power is restored. Only one diesel generator 
is required to perform this function per unit. Performance of the 
Surveillance Requirement as proposed provides greater assurance that 
the diesel generator is capable of performing its intended design 
function during an accident and the subsequent recovery. No 
significant changes to existing testing or new modes of facility 
operation are proposed by this change. 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.
    (3) Involve a significant reduction in the margin of safety 
because:
    The proposed amendment is required to ensure the diesel 
generator is tested in accordance with the design basis 
requirements. The changes represent a revision to the test 
acceptance criteria for performance of the largest single emergency 
load rejection surveillance for the diesel generator. This is a 
possible transient for the diesel generator that is bounded by a 
previously performed accident analysis. The proposed changes do not 
adversely affect the capability of the diesel generator to perform 
its design function. This function is to mitigate the consequences 
of an accident by providing an independent onsite source of 
alternate AC power with the capacity for operation of systems 
required to shutdown the reactor and maintain it in a safe shutdown 
condition until offsite power is restored. Performance of the 
Surveillance Requirement as proposed provides greater assurance that 
the diesel generator is capable of performing its intended design 
function during an accident and the subsequent recovery. Existing 
plant safety margins or the reliability of the equipment assumed to 
operate in the safety analysis are not changed. The proposed changes 
have been evaluated at Quad Cities and found to be acceptable for 
use based on system design, safety analysis requirements and 
operational performance. Since the changes maintain the necessary 
levels of system reliability, 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: 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 60603.
    NRC Project Director: Robert A. Capra.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 27, 1997.

    Description of amendment request: The proposed amendments would 
delete from the Technical Specifications (TS) of each unit the 
specified minimum volume of borated water available to the Standby 
Makeup Pump; the minimum volume is already specified in other parts of 
the 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:

    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This amendment to the Catawba TS maintains the necessary 
minimum volume of borated water available to mitigate a design basis 
SSS [standby shutdown system] event through a 72 hour period. 
Eliminating TS Surveillance 4.7.13.3a.2 does not increase the 
probability or consequences of any previously evaluated accident, 
since an adequate borated water source for the SMP [standby makeup 
pump] is continued to be required by other existing TS.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. This amendment to the Catawba TS continues to ensure that 
the necessary minimum volume of borated water is available to 
mitigate an SSS event. The SSS is required to mitigate certain 
previously evaluated design basis fire, security, and other events. 
This amendment does not create the possibility of a new or different 
kind of accident from any accident previously evaluated. This 
amendment changes the TS applicable to an accident mitigating 
function and does not impact any accident initiator, either new, 
different, or previously evaluated.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    No. This amendment continues to ensure that the necessary 
minimum volume of borated water is available to mitigate an SSS 
design basis event. The available minimum volume is maintained well 
above the design basis requirement. Since the source of borated 
water that is available to supply the SMP continues to be controlled 
by existing TS (TS 3.7.13.3a.1 and 3.9.10), which both envelope the 
current 112,320 gallons, sufficient volume has been and will 
continue

[[Page 33122]]

to be present to meet design basis requirements. Therefore, no 
reduction in a margin of safety will result from the changes 
proposed in this amendment.

    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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North 
Carolina 28242-0001.
    NRC Project Director: Herbert N. Berkow.

Duke Power Company, et al., Docket No. 50-414, Catawba Nuclear Station, 
Unit 2, York County, South Carolina
Date of amendment request: May 27, 1997

    Description of amendment request: The proposed amendment would 
delete from the Technical Specification of Unit 2 requirements 
regarding steam generator tube sleeving and repair. These requirements 
are not applicable to the Westinghouse Model D5 steam generators used 
by Unit 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) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This amendment to the Catawba Unit 2 Technical 
Specifications will have no impact on operation of the facility 
since the change will delete steam generator repair methods that are 
not applicable to the Catawba Unit 2 steam generators and have not 
been used to repair the Catawba Unit 2 steam generators.
    (2) Will the change create the possibility of a new or different 
type of accident from any accident previously evaluated?
    No. This amendment will delete steam generator repair methods 
that are not applicable and have not been used. Therefore, the 
proposed changes will not create the possibility of a new or 
different accident.
    (3) Will the change involve a significant reduction in the 
margin of safety?
    No. This amendment will delete steam generator repair methods 
that are not applicable and have not been used. There will be no 
impact on safety margins as a result of these changes.

    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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North 
Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf Nuclear 
Station, Unit 1, Claiborne County, Mississippi.
Date of amendment request: May 7, 1997.

    Description of amendment request: The amendment request would 
eliminate selected response time testing (RTT) surveillance 
requirements (SRs) from the Technical Specifications (TSs) for certain 
components of the following systems: reactor protection system (SR 
3.3.1.1.15), primary containment and drywell isolation instrumentation 
(SR 3.3.6.1.8), and emergency core cooling system (SRs 3.5.1.8 and 
3.5.2.7).
    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. No significant increase in the probability or consequences of 
an accident previously evaluated results from this change.
    The purpose of the proposed Technical Specification (TS) change 
is to eliminate response time testing (RTT) requirements for 
selected components in the Reactor Protection System (RPS), Primary 
Containment and Drywell Isolation Instrumentation, and Emergency 
Core Cooling System (ECCS) actuation instrumentation. The Boiling 
Water Reactor Owners' Group (BWROG) has completed an evaluation 
which demonstrates that [RTT] is redundant to the other TS-required 
testing. These other tests, in conjunction with actions taken in 
response to NRC Bulletin 90-01, ``Loss of Fill-Oil in Transmitters 
Manufactured by Rosemount,'' and Supplement 1 [to the bulletin], are 
sufficient to identify failure modes or degradations in instrument 
response time and ensure operation of the associated systems within 
acceptable limits. There are no known failure modes that can be 
detected by [RTT] that cannot also be detected by the other TS-
required testing. This evaluation was documented in NEDO-32291-A, 
``System Analyses for Elimination of Selected Response Time Testing 
Requirements,'' October 1995. EOI [The licensee] has confirmed the 
applicability of this evaluation to Grand Gulf Nuclear Power Station 
(GGNS). In addition, EOI will complete the actions identified in the 
NRC staff's Safety Evaluation of NEDO-32291-A.
    Elimination of [ECCS] RTT during MODES 4 and 5 [(i.e., cold 
shutdown and refueling, respectively)] is acceptable since there are 
no design basis accidents in MODES 4 and 5 for which the ECCS High 
Pressure Core Spray (HPCS) system is required to initiate within a 
specified period of time. The requirement to maintain [ECCS] 
OPERABLE during Modes 4 and 5 is preserved in the affected Technical 
Specification. The ECCS RTT required by SR 3.5.1.8 (applicable 
during MODES 1, 2, and 3, [or power operation, startup, and hot 
shutdown, respectively]) is adequate to identify any operability 
problems with the ECCS HPCS system. In addition, during MODES 4 and 
5, the probability and consequences of accidents are reduced due to 
the pressure and temperature limitations of these MODES.
    Because of the continued application of other TS-required tests 
such as channel calibrations, channel checks, channel functional 
tests, and logic system functional tests, the response time of these 
systems [listed in the first paragraph] will be maintained within 
the acceptance limits assumed in the plant [(GGNS)] safety analyses 
and required for successful mitigation of an initiating event. The 
proposed changes do not affect the capability of the associated 
systems to perform their intended function within their required 
response time, nor do the proposed changes themselves affect the 
operation of any equipment.
    As a result, EOI has concluded that the proposed changes do not 
involve a significant increase in the probability or the 
consequences of an accident previously evaluated.
    2. This change would not create the possibility of a new or 
different kind of accident from any [accident] previously evaluated.
    The proposed changes only apply to the testing requirements for 
the components [in the systems] identified above and do not result 
in any physical change to these or other components [in other 
systems] or their operation. As a result, no new failure modes are 
introduced. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. This change would not involve a significant reduction in a 
margin of safety.
    The current TS-required response times are based on the minimum 
allowable values assumed in the plant [(GGNS)] safety analyses. 
These analyses conservatively establish the margin of safety. As 
described above, the proposed changes do not affect the capability 
of the associated systems to perform their intended function within 
the allowable response time used as the basis for the plant safety 
analyses. The potential failure modes for the components within the 
scope of this request were evaluated for

[[Page 33123]]

impact on instrument response time. This evaluation confirmed that, 
with the exception of loss of fill-oil of Rosemount transmitters, 
the remaining TS-required testing is sufficient to identify failure 
modes or degradations in instrument response times and ensure 
operation of the instrumentation within the scope of this request is 
within acceptable limits. The actions taken in response to NRC 
Bulletin 90-01 and Supplement 1 [to the bulletin] are adequate to 
identify loss of fill-oil failures of Rosemount transmitters. As a 
result, it has been concluded that plant and system response to an 
initiating event will remain in compliance with the assumptions of 
the [GGNS] safety analysis. Elimination of RTT for ECCS HPCS system 
in MODES 4 and 5 does not reduce the margin of safety since there 
are no design basis events in MODES 4 and 5 requiring this system to 
respond in [a] specified period of time from onset of the event. 
Response time testing required by SR 3.5.1.8 (applicable during 
MODES 1, 2, and 3) is adequate to identify any equipment or 
operability concerns).
    Further, although not explicitly evaluated, the proposed changes 
will provide an improvement to plant safety and operation by 
reducing the time safety systems are unavailable, reducing the 
potential for inadvertent safety system actuation, reducing plant 
shutdown risk, limiting radiation exposure to plant personnel [that 
would be due to the RTT], and eliminating the diversion of key 
personnel resources to conduct unnecessary testing. Therefore, EOI 
concluded that this request will result in an overall increase in 
the margin of safety.

[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: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 24, 1997.

    Description of amendment request: The proposed amendment will 
modify Technical Specification (TS) 3/4.7.4, Ultimate Heat Sink (UHS), 
Table 3.7-3, by incorporating more restrictive dry cooling tower (DCT) 
fan requirements, and it will change the wet cooling tower water 
consumption in the TS Bases. This proposed amendment seeks to modify 
the TS to be consistent with revised design basis calculations.
    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 operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies the UHS TS by not allowing 
operation with less than 12 DCT fans per DCT. This change is 
necessary to adequately preserve the assumptions and limits of the 
revised UHS design basis calculations. These calculations conclude 
that the UHS is capable of dissipating the maximum peak heat load 
resulting from the limiting design bases accident (i.e., large break 
LOCA [large break loss of coolant accident]). The proposed change 
does not directly affect any material condition of the plant that 
could directly contribute to causing an accident or that could 
contribute to the consequences of an accident. The proposed change 
ensures that the mitigating effects of the UHS will be consistent 
with the design basis analysis. Therefore, the proposed change will 
not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change modifies the UHS TS to be consistent with 
revised design basis calculations. The UHS TS is being modified to 
eliminate operation with less than 12 DCT fans per DCT. The proposed 
change will not alter the operation of the plant or the manner in 
which the plant is operated. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change modifies the UHS TS by not allowing 
operation with less than 12 DCT fans per DCT. The proposed change 
preserves the margin of safety by ensuring that the UHS will be 
capable of dissipating the maximum design basis accident heat load 
with adequate margin. Therefore, the proposed change 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: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 24, 1997

    Description of amendment request: The proposed amendment will 
modify Technical Specifications (TS) 3.1.1.1, 3.1.1.2, 3.10.1 and 
Figure 3.1-1 by removing the cycle dependent boron concentration and 
boration flow rate from the Action Statements and removing the ``RWSP 
at 1720 ppm'' curve from the figure. A change to TS Bases 3/4.1.1.1 and 
3/4.1.1.2 has been included to support this change.
    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 operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The Shutdown Margin requirements are determined by the reload 
analysis performed every cycle. The Cycle 9 reload analysis has 
determined that the current Shutdown Margin requirements are 
acceptable. The proposed change eliminates the reference to 1720 ppm 
in the Action Statement because 1720 is not adequate to ensure that 
the Shutdown Margin requirements are met at the beginning of cycle. 
The proposed Action Statement will continue to ensure that in the 
event the Shutdown Margin requirements are not met, boration will be 
immediately initiated to restore the Shutdown Margin to within 
limits.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change does not change the design or configuration 
of the plant nor does it change how boration systems are operated 
during normal or accident conditions. It

[[Page 33124]]

ensures that the Shutdown Margin requirements for accidents already 
evaluated are promptly restored in the event that the requirements 
are not met.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change has not decreased the amount of Shutdown 
Margin required. The current Shutdown Margin requirements have been 
validated by the Reload Analysis for Cycle 9 and are adequate to 
ensure that the reactor can be made subcritical from all operating 
conditions, transients, and design basis events. The proposed change 
ensures that the Shutdown Margin requirements are promptly restored 
in the event that they are not met. As such, the proposed change 
ensures that the current margin of safety is maintained.
    Therefore, the proposed change 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: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 3, 1997

    Description of amendment request: The proposed amendment requests a 
change to the ACTION Requirements for Technical Specification 3/4.3.2 
for the Safety Injection System Sump Recirculation Actuation Signal 
(RAS). The proposed change will revise the allowed outage time for a 
channel of RAS to be in the tripped condition from ``prior to entry 
into the applicable MODE(S) following the next COLD SHUTDOWN'' to the 
more restrictive time limit of 48 hours and adds a shutdown 
requirement. Additionally, the 3.0.4 exemption is being removed from 
the ACTION for the tripped condition. A change to the Technical 
Specification Basis Section 3/4.3.2 has also been included.
    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 operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed revision to the TS changes the allowed outage time 
that a channel of RAS can be in the tripped condition from a maximum 
of approximately 18 months when one channel is inoperable and 92 
days when two channels are inoperable to 48 hours. If a channel were 
in the tripped condition and a single failure occurred (that of one 
other channel of RAS), a premature [refueling water storage pool] 
RWSP low level signal would be generated. During a Design Basis 
Accident with a containment high pressure condition causing the RWSP 
outlet check valves to seat, this single failure would prevent the 
contents of the RWSP from being injected into the reactor coolant 
system and possibly resulting in failure of both trains of 
[Emergency Core Cooling System] ECCS and [Containment Spray] CS. 
Additionally, this would cause the [Low Pressure Safety Injection] 
LPSI pumps to stop. Reducing the time that a channel of RAS can be 
placed in the tripped condition will reduce the probability of this 
scenario occurring during a Design Basis Accident. Since the allowed 
outage time for a channel of RAS is being limited to 48 hours, this 
is considered an off-normal operation and a single failure is not 
required to be postulated during a Design Basis Accident in the 
accident analysis. Reducing the time the channel can be placed in 
the tripped condition and thus, the exposure time to this scenario, 
would not be an accident initiator. The proposed change of being 
more conservative in the time and condition limits in the TS will 
not affect the assumptions, design parameters, or results of any 
accident previously evaluated.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change does not change the design or configuration 
of the plant. The proposed change provides a more conservative 
allowed outage time for the channel to be in the tripped condition. 
There has been no physical change to plant systems, structures or 
components nor will the proposed change reduce the ability of any of 
the safety-related equipment required to mitigate Anticipated 
Operational Occurrences or accidents. In fact, this change will 
potentially increase the ability of safety related equipment to 
perform its functions. The configuration required by the proposed 
specification is permitted by the existing specification.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change provides a more conservative allowed outage 
time for the channel to be in the tripped condition. By reducing the 
allowed outage time, the probability is reduced that a single 
failure (that of a failure of one channel of RAS with one channel in 
the tripped condition) would occur that would cause the suction to 
be prematurely supplied by the Safety Injection System Sump, 
potentially disabling the [High Pressure Safety Injection] HPSI and 
CS pumps, and stopping of the LPSI pumps. Therefore, the only change 
to the margin of safety would be an increase. Since the allowed 
outage time for a channel of RAS is being limited to 48 hours, this 
is considered an off-normal operation and a single failure is not 
required to be postulated during a Design Basis Accident in the 
accident analysis. The proposed changes do not affect the limiting 
conditions for operation or their bases.
    Therefore, the proposed change 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: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa
Date of amendment request: May 9, 1997

    Description of amendment request: The proposed amendment would 
revise the definitions of Limiting Safety System Setting (LSSS) and 
Instrument/Channel Calibration to reference a new program being added 
to the Technical Specification (TS) (Section 6.13) for the control of 
instrument setpoints.
    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

[[Page 33125]]

consideration, which is presented below:

    1. The proposed TS amendment will not significantly increase the 
probability or consequences of any previously-evaluated accidents.
    The proposed changes will not result in any direct hardware 
changes. The change only adds a program to the TS for the 
establishment and control of instrumentation setpoints that is 
consistent with current DAEC [Duane Arnold Energy Center] practice. 
The Instrument Setpoint Control Program is based upon a methodology 
for the calculation of instrument setpoints that conforms to the 
guidelines of Regulatory Guide 1.105, Rev. 2. The methodology 
ensures that adequate margin exists between the normal plant 
operating conditions and actual instrument setpoints to preclude 
spurious plant/equipment trips. As a result, the proposed program 
establishes the criteria for changes in instrument setpoints to 
ensure that such changes will not result in unnecessary plant 
transients. Consequently, the probability of any previously-analyzed 
event is not increased by this change.
    The role of the instrumentation and their associated setpoints 
is in detecting and mitigating plant events and thereby limiting the 
consequences of any previously-analyzed event. The LSSS[NTSP] and 
corresponding LTPO[AV] have been developed in accordance with the 
DAEC Instrument Setpoint Control Program criteria to ensure that the 
instrumentation remains capable of mitigating events as described in 
the safety analyses and that the results and consequences described 
in the safety analyses remain bounding. Therefore, these changes do 
not involve a significant increase in the consequences of an 
accident previously evaluated.
    2. The proposed changes will not create a new or different kind 
of accident from those previously evaluated.
    The proposed changes will not change the method or manner of 
plant operation, in particular, calibration of TS-required 
instrumentation. The use of the proposed TS program for the control 
of changes to instrument setpoints does not impact safe operation of 
the DAEC in that the design and safety analysis limits will continue 
to be satisfied. The proposed TS program involves no system 
additions or physical modifications, other than setpoint changes. 
Any setpoint changes must conform to the criteria set forth in the 
TS Instrument Setpoint Control Program. The instrument setpoints are 
developed using a methodology that conforms to the guidelines 
contained in Regulatory Guide 1.105, Rev. 2 to ensure the affected 
instrumentation remains capable of mitigating accidents and 
transients. Since operational methods remain unchanged and the 
instrument setpoints have been evaluated to maintain the plant 
within existing design basis criteria, no new or different type of 
accident is created.
    3. The proposed change will not result in a significant 
reduction in any margin of safety.
    The proposed TS program establishes the DAEC Instrument Setpoint 
Control Program, which is based upon an NRC-approved methodology. 
The program establishes the controls and criteria used to establish 
and revise instrument setpoints. The setpoint calculations use the 
uncertainties associated with the DAEC instrumentation and actual 
DAEC physical data and operating practices to ensure the validity of 
the resulting LTPO[AV] and LSSS[NTSP]. The methodology is based upon 
combining the uncertainties of the associated channels and takes 
into account calibration accuracy, instrument uncertainties, drift, 
etc. The use of this methodology for establishing these setpoints 
ensures that the design and/or safety analysis limits are not 
exceeded in any transient or accident. Therefore, the proposed 
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: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Gail H. Marcus.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa
Date of amendment request: May 9, 1997

    Description of amendment request: The proposed amendment would 
revise the definition of Limiting Condition for Operation (LCO) to 
address the situation when systems, components, etc., are removed from 
service or otherwise made inoperable during secondary modes of 
operation, without requiring entry into the LCO actions.
    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 TS amendment will not significantly increase the 
probability or consequences of any previously evaluated accidents.
    The proposed change merely adds criteria to the TS that are 
consistent with the original design and licensing basis assumptions. 
Operation in secondary modes of operation (such as surveillance 
testing, torus cooling mode (test line-up) or Residual Heat Removal 
system, and use of High Pressure Coolant Injection system or Reactor 
Core Isolation Cooling system in test line-up for reactor pressure 
control during transients) is assumed in the safety analysis report 
(Ref. UFSAR Section 6.3.4.2.1 and 7.3.4.2). Because no changes in 
actual equipment operation or testing are being made as part of this 
change, the probability of any event which could be induced by such 
operation or testing is not increased. Also, the change will ensure 
that the time such equipment is removed from service is kept very 
short in duration, either through existing TS Allowed Outage Time 
(AOT) notes or administratively by procedures. This is consistent 
with the assumption that the time in such secondary modes of 
operation (i.e., safe test interval) is much shorter than the 
allowable repair time (i.e., LCO time). Therefore, the proposed 
change will not significantly increase the probability of any 
previously evaluated accident.
    The uniform application of the new TS criteria will further 
ensure that the plant remains within the original design and 
licensing basis assumptions for equipment removed from service 
during secondary modes of operation. In particular, in the special 
case where testing also removes the redundant system, train, 
component, etc., from service, these criteria ensure that both 
affected systems, trains, etc., are properly controlled. This is 
acceptable because the time in such secondary modes of operation is 
very short in duration, such that the impact on the overall 
availability/reliability is insignificant. Therefore, the 
consequences of any previously analyzed accident are not 
significantly increased by this change.
    2. The proposed changes will not create a new or different kind 
of accident from those previously evaluated.
    The proposed changes will not add a new or different kind of 
accident because the plant will not be operated in a different way. 
Operation in secondary modes has been previously evaluated and found 
to be acceptable (Ref. General Electric reports APED-5736: Guideline 
for Determining Safe Test Intervals and Repair Times for Engineered 
Safeguards, and NEDO-10739: Methods for Calculating Safe Test 
Intervals and Allowable Repair Times for Engineered Safeguard 
Systems). The proposed change merely adds criteria to the TS that 
are consistent with the assumptions contained within these 
evaluations. Consequently, no new or different accidents are 
postulated as a result of this proposed change.
    3. The proposed change will not result in a significant 
reduction in any margin of safety.
    Because the criteria being added to the TS enforce the 
assumptions of the evaluations that form the basis of the existing 
TS (Ref. TS Bases 4.1, 4.2, and 3.5), the proposed change will not 
result in a significant reduction in any 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: Cedar Rapids Public Library,

[[Page 33126]]

00 First Street, SE., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Gail H. Marcus.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: December 20, 1996

    Description of amendment requests: The proposed amendments would 
reduce the frequency and scope of reactor coolant pump flywheel 
inspections.
    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:

    We have evaluated the proposed T/S changes and have determined 
they do not represent a significant hazards consideration based on 
the criteria established in 10 CFR 50.92(c). Operation of Cook 
Nuclear Plant in accordance with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    This change will reduce the frequency and scope of the 
surveillance testing on the reactor coolant pump flywheels. 
Operating power plants have been inspecting their flywheels for over 
20 years with no flaws identified which affect flywheel integrity. 
Past examinations performed to satisfy T/S 4.4.10.1 have not 
revealed any cracking of flywheel plates at Cook Nuclear Plant. 
Crack extension over a 60 year service life is negligible. 
Structural reliability studies have shown that eliminating 
inspections after 10 years of plant life will not significantly 
change the probability of failure. Most flaws which could lead to 
failure would be detected during preservice inspection or, at worst, 
early in plant life, and crack growth over plant life is negligible. 
As stated in the SER associated with WCAP-14535, assuming an initial 
crack of 10% of the distance from the keyway to the flywheel outer 
radius and a maximum fatigue crack growth, ASME margins would be 
maintained during the 10-year inspection period. Therefore, the 
change in test frequency will not endanger public health or safety. 
For these reasons, it is our belief the proposed changes do not 
involve a significant increase in the probability or consequences of 
a previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The changes will not introduce any new modes of plant operation, 
nor will any physical changes to the plant be required. Thus, the 
changes will not create the possibility of a new or different kind 
of accident from any accident previously analyzed or evaluated.
    3. Involve a significant reduction in a margin of safety.
    This change will reduce the frequency and scope of the 
surveillance testing on the reactor coolant pump flywheels. 
Operating power plants have been inspecting their flywheels for over 
20 years with no flaws identified which affect flywheel integrity. 
Past examinations performed to satisfy T/S 4.4.10.1 have not 
revealed any cracking of flywheel plates at Cook Nuclear Plant. 
Crack extension over a 60 year service life is negligible. 
Structural reliability studies have shown that eliminating 
inspections after 10 years of plant life will not significantly 
change the probability of failure. Most flaws which could lead to 
failure would be detected during preservice inspection or at worst 
early in plant life, and crack growth over plant life is negligible. 
As stated in the SER associated with WCAP-14535, assuming an initial 
crack of 10% of the distance from the keyway to the flywheel outer 
radius and a maximum fatigue crack growth, ASME margins would be 
maintained during the 10-year inspection period. For these reasons, 
it is our belief 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 requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: April 30, 1997

    Description of amendment request: The proposed amendment would 
remove Technical Specifications (TSs) regarding meteorological 
monitoring instrumentation in accordance with NRC Generic Letter (GL) 
95-10, ``Relocation of Selected Technical Specification Requirements 
Related to Instrumentation.'' Specifically, the amendment would delete 
TS 3/4.3.7.3, ``Meteorological Monitoring Instrumentation,'' including 
associated TS Tables 3/4.3.7.3-1, and TS Bases 3/4.3.7.3. The TS Index 
would be revised to show these deletions. The deletion of TS 3.3.7.3 
would also eliminate the requirement that a Special Report to be 
submitted to the NRC pursuant to TS 6.9.2 when one or more 
meteorological monitoring instrumentation channels is inoperable for 
more than 7 days. The licensee states that the deleted requirements 
would be relocated to the Updated Safety Analysis Report (USAR), except 
that the special reporting requirement would be discontinued as the 
licensee would continue to evaluate future inoperability of 
meteorological instrumentation for reportability in accordance with 10 
CFR 50.72 and 10 CFR 50.73. The licensee will also insert the word 
``nominal'' in the relocated tables in the USAR to indicate that the 
meteorological instrumentation elevations of 30 and 200 feet are 
nominal elevations (this change would be made because, as the licensee 
reported in LER 96-14, the actual locations of the air temperature 
monitoring instruments are 26.8 feet and 194.8 feet and the actual 
locations of the wind indicator (speed and direction) monitoring 
instruments are 30.9 feet and 199.4 feet). As stated in GL 95-10, the 
NRC staff has determined that meteorological monitoring instrumentation 
does not serve such a primary protective function as to warrant 
inclusion in the TS in accordance with 10 CFR 50.36 criteria. Thus, in 
GL 95-10, the NRC staff established that relocation of the 
meteorological instrumentation requirements to the USAR (whereby 
changes are controlled by the licensee pursuant to 10 CFR 50.59) is 
acceptable.
    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 operation of Nine Mile Point Unit 2 [NMP2], in accordance 
with the proposed amendment, will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The NMP2 meteorological monitoring instrumentation is used to 
provide data for use in radioactive dose assessment with respect to 
routine or accidental releases of radioactive materials to the 
atmosphere. The deletion of the special reporting requirements is an 
administrative change. The subject special reporting requirements 
serve no nuclear related protective function. The relocation of the 
meteorological monitoring instrumentation requirements from the TSs 
to the USAR, and the addition of the word nominal to the USAR and 
tables, will not increase the probability of an accident since the 
specification applies only to monitoring instrumentation. This also 
is an administrative change and does not reduce the effectiveness of 
the current instrumentation requirements. The meteorological 
monitoring instrumentation

[[Page 33127]]

requirements are not precursors to any accident previously 
evaluated. According to the NRC Staff (GL 95-10), the meteorological 
monitoring instrumentation does not serve to ensure the plant is 
operated within the bounds of initial conditions assumed in any 
design basis accidents or transients previously evaluated, or that 
the plant will be operated to preclude transients or accidents. In 
addition, the meteorological monitoring instrumentation does not 
function as part of the primary success path of a safety sequence 
analysis used to demonstrate that the consequences of these events 
are within the appropriate acceptance criteria. Therefore, the 
proposed changes do not significantly increase the probability or 
consequences of an accident previously evaluated.
    2. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed deletion of the special reporting requirements is 
an administrative change. The subject special reporting requirements 
serve no nuclear related protective function. The proposed change 
also removes meteorological monitoring instrumentation 
specifications from the NMP2 TSs. This also is an administrative 
change and does not reduce the effectiveness of the current 
instrumentation requirements. The relocation of the meteorological 
instrumentation requirements to the USAR, and the addition of the 
word nominal to the USAR and tables, will not create the possibility 
of a new or different kind of accident since the specification only 
applies to monitoring instrumentation. The NRC Staff has concluded 
in GL 95-10 that the provisions of the meteorological monitoring 
instrumentation specifications are not related to dominant 
contributors to plant risk. The NMP2 meteorological instrumentation 
is used to provide data for use in radioactive dose assessment with 
respect to routine or accidental releases of radioactive materials 
to the atmosphere. Since no physical modification to the plant is 
being performed, and no changes to actual plant operations are 
required by the change, removal of the specifications from the NMP2 
TSs will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The proposed deletion of the special reporting requirements is 
an administrative change. The subject special reporting requirements 
serve no nuclear related protective function. The proposed removal 
of the instrumentation requirements from the NMP2 TSs is also an 
administrative change and does not reduce the effectiveness of the 
current instrumentation requirements. The relocation of the 
meteorological instrumentation requirements to the USAR, and the 
addition of the word nominal to the USAR and tables, will not 
involve a reduction in a margin of safety since the specification 
only applies to monitoring instrumentation. The instrumentation will 
continue to meet the requirements of Regulatory Guide 1.23, and the 
offsite dose calculations will continue to use the actual measured 
elevation differences. In GL 95-10, the NRC Staff concluded (1) That 
the meteorological monitoring instrumentation does not function as 
part of the primary success path of a safety sequence analysis, and 
(2) that the meteorological monitoring instrumentation 
specifications are not related to dominant contributors to plant 
risk. Therefore, the removal of the meteorological monitoring 
instrumentation specifications from the NMP2 TSs will not result in 
a significant reduction in any margin of safety.

    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: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Alexander W. Dromerick, Acting Director.

Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone 
Nuclear Power Station, Unit No. 1, New London County, Connecticut
Date of amendment request: May 15, 1997

    Description of amendment request: The proposed amendment would 
revise Technical Specification Sections 3.1 and 4.1 ``Reactor 
Protection System'' and the associated Bases to remove run mode 
intermediate range monitor high flux/inoperative with the associated 
average power range monitor downscale scram trip function and 
incorporate editorial revisions.
    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 operation of Millstone Nuclear Power Station, Unit No. 1, 
in accordance with the proposed amendment, will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    No physical change is being made to any systems or components 
that are credited in the safety analysis, therefore there is no 
change in the probability or consequences of any accident analyzed 
in the UFSAR [Updated Final Safety Analysis Report].
    The design basis accident applicable to the startup power region 
is the Control Rod Drop Accident (CRDA). The UFSAR does not credit 
the RUN Mode IRM [intermediate range monitor] High Flux/Inoperative 
with the associated APRM [average power range monitor] downscale 
scram Trip Function (IRM RUN Mode SCRAM) in the termination of this 
accident. Accident mitigation is provided by the APRM 120% power 
scram. Therefore, elimination of the IRM RUN Mode SCRAM function has 
no adverse affect on previously evaluated accidents.
    The Continuous Control Rod Withdrawal Error (CWE) transient is 
terminated by the Rod Block Monitor (RBM) in the RUN Mode. The APRM 
Reduced High Flux Scram provides the primary STARTUP Mode protection 
in conjunction with the IRMs and limits the consequences of this 
transient. Therefore, elimination of the IRM RUN Mode SCRAM function 
has no effect on the consequences of this transient.
    Clarification of the LCO [limiting condition for operation] RPS 
[reactor protection system] Table aligns requirements with Limiting 
Safety System Settings. Further revisions to LCO 3.1 Reactor 
Protection System Table 3.1.1 and associated TS [technical 
specification] bases to clarify APRM Trip Functions do not alter the 
required trip functions. Deletion of RUN requirement and associated 
Action B for Reduced High Flux fixes an editorial error introduced 
in a previous amendment. This trip function is not effective with 
the mode switch in the RUN position and removal does not alter the 
neutron monitoring requirements credited in the accident analyses.
    Adding a new surveillance to verify SRM [source range monitor]/
IRM/APRM overlap will enhance neutron monitoring during startups and 
shutdowns and does not have an adverse affect on previously 
evaluated accidents.
    None of the proposed changes will affect any of the rod blocks 
or other precursor events to either the CRDA or CWE. Therefore, 
there is no change in the probability of any accident previously 
analyzed.
    2. The operation of Millstone Nuclear Power Station, Unit No. 1, 
in accordance with the proposed amendment, will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes affect only the operations of neutron 
monitoring and protective systems (IRM and APRM) which provide 
indication and mitigation actions only. Operation of these systems 
does not create the possibility for new precursors (such as 
reactivity) which would introduce a new or different kind of 
accident from any accident previously evaluated.
    Additionally, the proposed changes do not affect the ability of 
those systems required to mitigate previously evaluated accidents 
during the modes they are credited.
    3. The operation of Millstone Nuclear Power Station, Unit No. 1, 
in accordance with the proposed amendment, will not involve a 
significant reduction in a margin of safety.
    The only scram function that the UFSAR takes credit for in the 
mitigation of the limiting accident (control rod drop accident) is 
the APRM 120% power scram which is not

[[Page 33128]]

affected by this change. Only the IRM RUN Mode SCRAM, for which the 
UFSAR takes no credit in the termination of any analyzed event, is 
removed by this change. Removal of the IRM RUN Mode SCRAM will avoid 
the need to operate the plant in a ``half scram'' condition with the 
potential for an inadvertent plant transient. For these reasons, the 
change does not involve a significant reduction in a margin of 
safety.
    The Continuous Control Rod Withdrawal Error (CWE) transient is 
terminated by the Rod Block Monitor (RBM) in the RUN Mode. When 
initiated from the STARTUP Mode, the consequences of a CWE are 
limited by the APRM Reduced High Flux scram in conjunction with the 
IRM scram function. Therefore eliminating the TS requirement for the 
IRM RUN Mode SCRAM will not reduce the margin of safety for this 
transient.
    Clarification of the LCO RPS Table aligns requirements with 
Limiting Safety System Settings. Further revisions to LCO 3.1 
Reactor Protection System Table 3.1.1 and associated TS bases to 
clarify APRM Trip Functions do not alter the required trip 
functions. Deletion of the RUN requirement and associated Action B 
for Reduced High Flux corrects an editorial error introduced in a 
previous amendment. This trip function is not effective with the 
mode switch in the RUN position and removal does not alter the 
neutron monitoring requirements credited in the accident analyses.
    Adding a new surveillance to verify SRM/IRM/APRM overlap will 
enhance neutron monitoring during startups and shutdowns and 
consequently 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: Learning Resources Center, 
Three Rivers Community--Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Deputy 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: May 20, 1997

    Description of amendment request: This submittal supersedes the 
January 22, 1996, submittal which was previously noticed on February 
28, 1996 (61 FR 7554). The proposed change would relocate the 
containment isolation valve (CIV) list, Table 3.6-2, from the Technical 
Specifications to the Technical Requirements Manual (TRM). This change 
would affect Technical Specification Sections 1.8.1.b, 4.6.1.1.a, 
3.6.3.1, 4.6.3.1.1, and 4.6.3.1.2, and Basis Section 3/4.6.3. A note at 
the bottom of Table 3.6-2 regarding the CIVs that are subject to 
administrative controls is retained in the Technical Specifications by 
relocating it to Sections 1.8.1.b and 3.6.3.1. This change is being 
performed in accordance with Generic Letter 91-08, which provides 
guidance for removal of component lists from the Technical 
Specifications.
    Additionally, a change to provide relief in the surveillance 
requirement in Section 4.6.1.1.a is included. The change allows valves, 
blind flanges, and deactivated automatic valves located inside the 
containment and are locked, sealed, or otherwise secured in the closed 
position to be verified closed prior to entering Mode 4 from Mode 5, if 
not performed within the previous 92 days. The current requirements 
check the valve position once per 31 days.
    TS Bases Section 3/4.6.3 is updated to reflect the removal and 
relocation of the CIV list to the TRM. Also, details of the 
administrative controls for operating CIVs while in Modes 1 through 4 
are added to Bases Section 3/4.6.3.
    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:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to relocate the containment isolation valve 
(CIV) list will not result in any hardware or equipment operating 
changes. The proposed change is based on Generic Letter (GL) 91-08 
and merely relocates the CIV table and removes all references to the 
table. The relocation of the CIV table from the Technical 
Specifications does not affect the operability requirements of any 
of the listed valves. Technical Specifications will still continue 
to require the CIVs to be operable. The LCO [limiting condition for 
operation] and surveillance requirements for the valves will remain 
in Technical Specifications. The CIV table will be relocated to the 
Millstone Unit No. 2 Technical Requirements Manual (TRM), which is 
controlled in accordance with 10 CFR 50.59. This change does not 
alter the design, function, or operation of the valves involved. 
Thus, there is no significant affect on the possibility or 
consequences of any previously evaluated accident.
    The change to Surveillance Requirement (SR) 4.6.1.1.a will allow 
the valves, blind flanges and deactivated automatic valves located 
inside the containment that are locked, sealed, or otherwise secured 
in the closed position to be verified closed prior to entering Mode 
4 from Mode 5, if not performed within the previous 92 days, instead 
of the current 31 day requirement. This means that the surveillance 
interval could be as long as the entire operating cycle, depending 
on whether entry into Mode 5 is required during the cycle. The 
change in the surveillance frequency (increase in time from 31 days 
to not less than 92 days and only prior to entering Mode 4 from Mode 
5) recognizes that these valves are operated under administrative 
controls and probability of misalignment is low. This provides 
adequate assurance that the containment function assumed in the 
accident analysis will be maintained. Therefore, there is no 
significant affect on the probability or consequences of any 
previously evaluated accident. This proposed change to SR 4.6.1.1.a 
is consistent with NUREG-1432 Standard Technical Specifications for 
Combustion Engineering Pressurized Water Reactors Revision 1 (SR 
3.6.3.4).
    The information added to the Bases will provide additional 
guidance to ensure the plant is operated correctly. This information 
will not result in any new approaches to plant operation. Therefore, 
there is not significant affect on the probability or consequences 
of any previously evaluated accident.
    These proposed changes do not alter the design, function, or 
operation of the valves involved. Therefore, there is no significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The change to relocate the CIV list from the Technical 
Specifications to the TRM will not impose any different operational 
or surveillance requirements, nor will the change remove any such 
requirements. Adequate control will be maintained. Furthermore, as 
stated above, the proposed change does not alter the design, 
function, or operation of the valves involved, and therefore does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The change to SR 4.6.1.1.a reduces the surveillance frequency 
for valves, blind flanges and deactivated automatic valves located 
inside the containment. It does not alter the design, function, or 
operation of the valves. Therefore, it does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The information added to the Bases will provide additional 
guidance to ensure the plant is operated correctly. This information 
does not alter the design, function, or operation of the valves 
involved. Therefore, it does 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 will not reduce the margin of safety since 
they have no impact on any safety analysis assumption. The proposed 
changes do not decrease the scope

[[Page 33129]]

of equipment currently required to be operable or subject to 
surveillance testing, nor do the proposed changes affect any 
instrument setpoints or equipment safety functions.
    The effectiveness of Technical Specifications will be maintained 
since the change will not alter function or operability requirements 
for any CIV. In addition, the relocation of the valve list is 
consistent with the guidance provided in GL 91-08, and the change to 
the surveillance interval is consistent with NUREG-0212 Standard 
Technical Specifications for Combustion Engineering Pressurized 
Water Reactors Revision 2 (LCO 3.6.1.1) and NUREG-1432 Standard 
Technical Specifications for Combustion Engineering Pressurized 
Water Reactors Revision 1 (LCO 3.6.3).
    The information added to the Bases is consistent with the 
guidance provided in GL 91-08 and NUREG-1432 Standard Technical 
Specifications for Combustion Engineering Pressurized Water Reactors 
Revision 1. The intent of the Technical Specifications will be met 
since this information will not result in any new approaches to 
plant operation.
    Therefore, there is no 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, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Deputy Director: Phillip F. McKee.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut
Date of amendment request: May 9, 1997

    Description of amendment request: The proposed amendment would 
revise the shutdown margin requirements and add Technical Specification 
3/4.3.5 to provide the limiting condition for operation (LCO) and 
surveillance requirements for the shutdown margin monitors. The 
proposed amendment would also make administrative changes and revise 
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 changes in accordance with 10 
CFR 50.92 and has concluded that the change does not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10 CFR 50.92(c) are not 
satisfied. The proposed changes do not involve [an] SHC because the 
changes would not: 
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed Technical Specification changes will revise the 
current shutdown margin requirements for Modes 3, 4 and 5 in Figures 
3.1-1, 3.1-2, 3.1-3, 3.1-4 and 3.1-5 and allow for additional 
boration of the RCS [reactor coolant system] as directed by 
Specification 3.3.5. The new Shutdown Margin requirements are based 
on re-analyses of the Boron Dilution Event provided by Westinghouse. 
In the re-analyses, assumptions were modified in order to justify 
the operability of the Shutdown Margin Monitor for count rates which 
are lower than currently allowed. The proposed Shutdown Margin 
requirements for Modes 3, 4 and 5 will continue to assure that the 
operator has a minimum of 15 minutes from the alarm to loss of 
shutdown margin during an assumed Boron Dilution Event.
    The proposed change also adds Technical Specification 3/4.3.5 to 
provide the LCO and Surveillance Requirements for the Shutdown 
Margin Monitors. LCO 3.3.5 refers to the Core Operating Limits 
Report (COLR) which will specify the minimum count rate/alarm ratio 
requirements in order to consider the Shutdown Margin Monitors 
operable. The LCO also directs the additional boration of the RCS in 
order to allow the Shutdown Margin Monitors to be considered 
operable for lower count rates. Also, a footnote (**) is included in 
Specification 3/4.3.5 to make the Specification treatment of the 
valves consistent with the Mode 6 and Mode 5-loops drained 
requirements.
    Due to the addition of Technical Specification 3/4.3.5, the 
related Bases information is added as BASES Section 3/4.3.5. 
Additionally, the Bases information for the Shutdown Margin Monitors 
which is currently in BASES Section 3/4.3.1 is moved to the added 
BASES Section 3/4.3.5. This Bases information is also revised to be 
consistent with the added Specification 3/4.3.5.
    Also, due to the addition of Technical Specification 3/4.3.5, 
the guidance related to the Shutdown Margin Monitor in Tables 3.3-1 
and 4.3-1 is deleted to avoid redundancy.
    Additionally, Section 3/4.1.2 of the Bases is revised so that it 
refers to Figure 3.1-4 (Shutdown Margin for Mode 5/filled) instead 
of Figure 3.1-5 (Shutdown Margin for Mode 5/drained). This change 
will make the Bases consistent with the ACTION statement 
requirements of Technical Specifications 3.1.2.2 and 3.1.2.6.
    Finally, Reference 12 (NUSCO-152, Addendum 4) is added to the 
list of references in Section 6.9.1.6.b. The addition of this 
reference is considered administrative and is not related to or 
required by the changes proposed for the Shutdown Margin 
requirements or Shutdown Margin Monitors.
    The new requirements for increased Shutdown Margin (Figures 3.1-
1 to 3.1-5) and additional boration (LCO 3.3.5) continue to assure 
that the operator will have a response time of at least 15 minutes 
to mitigate the consequences of a Boron Dilution Event. The 
implementation of the new requirements does not alter the alignment 
of any plant equipment and therefore, the change cannot increase the 
probability or consequences of any previously analyzed accident.
    The proposed changes will not adversely affect the assumptions 
or results of other FSAR [Final Safety Analysis Report] accident 
analysis and it is concluded that this change is safe. The changes 
do not adversely affect any equipment credited in the safety 
analysis.
    Based upon the re-analyses of the boron dilution event, revised 
plant operating requirements (shutdown margin) are generated to 
maintain the required operator action time. Therefore, there is no 
effect on the probability of occurrence or consequences of 
previously evaluated accidents.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed Shutdown Margin requirements for Modes 3, 4 and 5 
(Figures 3.1-1 to 3.1-5 and additional boration as per Specification 
3.3.5) will continue to assure that the operator has a minimum of 15 
minutes from the alarm to loss of shutdown margin during an assumed 
Boron Dilution Event. Additionally, the use of these revised 
requirements allows the Shutdown Margin Monitor to be considered 
operable for count rates which are lower than currently allowed.
    The changes do not introduce any new failure modes or 
malfunctions since the changes implement revised, more conservative 
plant operating requirements (shutdown margin) which are based on 
re-analyses of the Boron Dilution Event. Also, the changes do not 
eliminate any existing requirements.
    Therefore, 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 Shutdown Margin requirements for Modes 3, 4 and 5 
(Figures 3.1-1 to 3.1-5 and additional boration as per Specification 
3.3.5) will continue to assure that the operator has a minimum of 15 
minutes from the alarm to loss of shutdown margin during an assumed 
Boron Dilution Event. Additionally, the use of these revised 
requirements allows the Shutdown Margin Monitor to be considered 
operable for count rates...which are lower than currently allowed.
    The re-analyses of the Boron Dilution Event demonstrated that 
the required

[[Page 33130]]

operator action time is maintained. As such, the re-analyses will 
become the ``analysis of record'' for the Boron Dilution Event in 
Modes 3, 4 and 5. The Boron Dilution Event analysis is documented in 
FSAR Chapter 15.4.6.
    The re-analyses of the Boron Dilution Event and the proposed 
revisions to the Technical Specifications do not adversely affect 
the results of the current FSAR accident analysis and therefore, it 
is concluded that this change is safe. Additionally, the change does 
not adversely affect any equipment credited in the safety analysis.
    The changes do not have an adverse impact on the protective 
boundaries and there is no reduction in the margin of safety as 
specified in the Technical Specifications. Thus, this proposed 
change does not involve a significant reduction in the margin of 
safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed changes do not involve an SHC.

    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, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Deputy Director: Phillip F. McKee.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut
Date of amendment request: May 14, 1997

    Description of amendment request: Technical Specification 
Surveillance Requirement 4.8.2.1.c.4 requires that each battery charger 
be tested to verify that it can supply a specified current at 125 
volts. The proposed amendment would increase the required test voltage.
    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 revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed revision does not involve [an] SHC because 
the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes to Technical Specification Surveillance 
4.8.2.1.c.4 to increase the required test voltage for the battery 
chargers from 125 volts to greater than or equal to 132 volts is 
consistent with the design criteria of the chargers and performing 
battery charger surveillance testing does not significantly increase 
the probability of an accident previously evaluated. The proposed 
changes to increase the required test voltage for the battery 
chargers provides the necessary assurance that the battery chargers 
will function as required in previous evaluations and does not 
significantly increase the consequence of an accident previously 
evaluated.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence 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 to Technical Specification Surveillance 
4.8.2.1.c.4 to increase the required test voltage for the battery 
chargers from 125 volts to greater than or equal to 132 volts does 
not change the operation of the battery chargers during normal or 
accident evaluations.
    Therefore, the proposed revision does not create the possibility 
or a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to Technical Specification Surveillance 
4.8.2.1.c.4 to increase the required test voltage for the battery 
chargers from 125 volts to greater than or equal to 132 volts 
provides assurance that the battery chargers are capable of 
supplying the largest combined demands of the various steady state 
loads, plus the current required to recharge its battery, which has 
undergone a duty cycle discharge, to its fully charged condition in 
less than 24 hours.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve an SHC.

    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, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Deputy Director: Phillip F. McKee.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek 
Generating Station, Salem County, New Jersey
Date of amendment request: March 3, 1997 as supplemented by letter 
dated May 5, 1997. The May 5, 1997, supplement revised the proposed no 
significant hazards consideration entirely

    Description of amendment request: The proposed changes to the Hope 
Creek (HC) Technical Specifications (TSs) would: (1) Change TS 3/4.3.1, 
``Reactor Protection System Instrumentation,'' TS 3/4.3.2, ``Isolation 
Actuation Instrumentation,'' and TS 3/4.3.3, ``Emergency Core Cooling 
System Actuation Instrumentation'' to include additional information 
concerning response time testing; (2) Change TS 4.0.5 to reference 
inservice inspection and test requirements; (3) Change TS 3/4.6.1, 
``Primary Containment,'' and associated Bases to reflect a design 
modification; (4) Change TS 3/4.7.7, ``Main Turbine Bypass System,'' to 
specify a new operability requirement; and (5) Change the Bases for TS 
3/4.8, ``Electrical Power Systems.''
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes for the TS related to response time testing 
reflect testing methodologies that were approved by the NRC in 
Amendment No. 85 to the Hope Creek TS. These proposed TS revisions 
involve: (1) no hardware changes; (2) no significant changes to the 
operation of any systems or components in normal or accident 
operating conditions; and (3) no changes to existing structures, 
systems or components. Therefore, these changes will not increase 
the probability of an accident previously evaluated. Since the plant 
systems associated with these proposed changes will still be capable 
of: (1) meeting all applicable design

[[Page 33131]]

basis requirements; and (2) retain the capability to mitigate the 
consequences of accidents described in the HC [Updated Final Safety 
Analysis Report] UFSAR, the proposed changes were determined to be 
justified. As a result, these changes will not involve a significant 
increase in the consequences of an accident previously evaluated.
    The proposed changes to Surveillance Requirement 4.0.5 do not 
alter the current requirements for the Hope Creek inservice 
inspection and inservice testing programs and are considered to be 
editorial in nature. These proposed TS revisions involve: (1) no 
hardware changes; (2) no significant changes to the operation of any 
systems or components in normal or accident operating conditions; 
and (3) no changes to existing structures, systems or components. 
Therefore, these changes will not increase the probability of an 
accident previously evaluated. Since the plant systems associated 
with these proposed changes will still be capable of: (1) Meeting 
all applicable design basis requirements; and (2) retain the 
capability to mitigate the consequences of accidents described in 
the HC UFSAR, the proposed changes were determined to be justified. 
As a result, these changes will not involve a significant increase 
in the consequences of an accident previously evaluated.
    The proposed changes to the drywell and suppression chamber 
purge system are being made to justify design modifications to that 
system. As discussed in NRC Notice of Violation 50-354/96-10-01, 
this design modification replaced isolation valves containing 
resilient material seals with metal seated valves under 10CFR50.59. 
As a result of this modification, a 24 month frequency has been 
implemented to perform Type C tests on these new metal seated 
valves. PSE&G has concluded that the 24 month frequency is 
appropriate for the new valves since: (1) This frequency is imposed 
by Surveillance Requirement 4.6.1.2.d, which is applicable to 
similar containment isolation valves in Table 3.6.3-1 that penetrate 
the primary containment; and (2) concerns raised about severe 
environment-induced degradation and frequent use for the previously 
installed resilient seal material valves are not applicable to the 
replacement metal seat valves. PSE&G has concluded that the valve 
modification was an enhancement to the Hope Creek design that did 
not impact the isolation capability of the drywell and suppression 
chamber purge system. No significant changes were made to the 
operation of these valves in normal or accident operating 
conditions. As a result, these changes will not increase the 
probability of an accident previously evaluated. Since the plant 
systems associated with these proposed changes will still be capable 
of: (1) Meeting all applicable design basis requirements; and (2) 
retain the capability to mitigate the consequences of accidents 
described in the HC UFSAR, the proposed changes were determined to 
be justified. As a result, these changes will not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    The proposed changes to [Limiting Condition for Operation] LCO 
3.7.7 establish consistent and appropriate requirements for main 
turbine bypass valve operability requirements. These changes do not 
impact the assumptions contained in these UFSAR analyses since they 
do not change the manner in which Hope Creek is currently permitted 
to operate. Since the ACTION Statement for LCO 3.7.7 already allows 
indefinite continued operation below 25% of RATED THERMAL POWER with 
an inoperable main turbine bypass valve system, the proposed 
modification to the APPLICABILITY statement for this LCO does not 
involve: (1) Hardware changes; (2) significant changes to the 
operation of any systems or components in normal or accident 
operating conditions; or (3) changes to existing structures, systems 
or components. Therefore these changes will not increase the 
probability of an accident previously evaluated. Since the plant 
systems associated with these proposed changes will still be capable 
of: (1) meeting all applicable design basis requirements; and (2) 
retain the capability to mitigate the consequences of accidents 
described in the HC UFSAR, the proposed changes were determined to 
be justified. As a result, these changes will not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    The proposed changes to the HC emergency diesel generator (EDG) 
TS Bases [Change 5--Bases for TS 3/4.8, ``Electrical Power 
Systems''] include information contained in the Safety Evaluation 
Report for Technical Specification Amendment No. 75. This 
information concerns the bases for the allowed-outage-time (AOT) for 
the C and D EDGs. Concerning the revisions to planned C and D EDG 
outages, PSE&G believes that implementation of 10CFR50.65 
requirements to monitor EDG unavailability will provide an 
acceptable and more clearly defined method for maintaining EDG 
availability within acceptable limits. As stated in PSE&G's letter 
LR-N97167, dated March 21, 1997, Hope Creek will not plan C or D EDG 
outages that exceed 72 hours if the total unavailability of the EDG 
will be greater than 720 hours on a 12 month rolling basis. The 
proposed TS revisions involve: (1) no hardware changes; (2) no 
significant changes to the operation of any systems or components in 
normal or accident operating conditions; and (3) no changes to 
existing structures, systems or components. Therefore these changes 
will not increase the probability of an accident previously 
evaluated. Since the plant systems associated with these proposed 
changes will still be capable of: (1) Meeting all applicable design 
basis requirements; and (2) retain the capability to mitigate the 
consequences of accidents described in the HC UFSAR, the proposed 
changes were determined to be justified. As a result, these changes 
will not involve a significant increase in the 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.
    The proposed changes for the TS related to response time testing 
reflect testing methodologies that were approved by the NRC in 
Amendment No. 85 to the Hope Creek TS and are being made to clarify 
the licensing basis for performing response time testing. The 
proposed changes will not adversely impact the operation of any 
safety related component or equipment. Since the proposed changes 
involve: (1) No hardware changes; (2) no significant changes to the 
operation of any systems or components; and (3) no changes to 
existing structures, systems or components, there can be no impact 
on the occurrence of an accident previously evaluated. Furthermore, 
there is no change in plant testing proposed in this change request 
that could initiate an event. Therefore, these changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed changes to Surveillance Requirement 4.0.5 do not 
alter the current requirements for the Hope Creek inservice 
inspection and inservice testing programs and are considered to be 
editorial in nature. The proposed changes will not adversely impact 
the operation of any safety related component or equipment. Since 
the proposed changes involve: (1) No hardware changes; (2) no 
changes to the operation of any systems or components; and (3) no 
changes to existing structures, systems or components, there can be 
no impact on the occurrence of an accident previously evaluated. 
Furthermore, there is no change in plant testing proposed in this 
change request that could initiate an event. Therefore, these 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes to the drywell and suppression chamber 
purge system are being made to justify design modifications to that 
system. As discussed in NRC Notice of Violation 50-354/96-10-01, 
this design modification replaced isolation valves containing 
resilient material seals with metal seated valves under 10 CFR 
50.59. As a result of this modification, a 24 month frequency has 
been implemented to perform Type C tests on these new metal seated 
valves. PSE&G has concluded that the 24 month frequency is 
appropriate for the new valves since: (1) This frequency is imposed 
by Surveillance Requirement 4.6.1.2.d, which is applicable to 
similar containment isolation valves in Table 3.6.3-1 that penetrate 
the primary containment; and (2) concerns raised about severe 
environment-induced degradation and frequent use for the previously 
installed resilient seal material valves are not applicable to the 
replacement metal seat valves. PSE&G has concluded that the valve 
modification was an enhancement to the Hope Creek design that did 
not impact the isolation capability of the drywell and suppression 
chamber purge system. Since the proposed changes will not adversely 
impact the operation of any safety related component or equipment, 
there can be no impact on the occurrence of any accident. 
Furthermore, there is no change in plant testing proposed in this 
change request that could initiate an event. Therefore, these 
changes will not create the possibility of a

[[Page 33132]]

new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to LCO 3.7.7 establish consistent and 
appropriate requirements for main turbine bypass valve operability 
requirements. These changes do not impact the assumptions contained 
in these UFSAR analyses since they do not change the manner in which 
Hope Creek is currently permitted to operate. Since the ACTION 
Statement for LCO 3.7.7 already allows indefinite continued 
operation below 25% of RATED THERMAL POWER with an inoperable main 
turbine bypass valve system, the proposed modification to the 
APPLICABILITY statement for this LCO does not involve: (1) hardware 
changes; (2) significant changes to the operation of any systems or 
components in normal or accident operating conditions; or (3) 
changes to existing structures, systems or components. The proposed 
changes will not adversely impact the operation of any safety 
related component or equipment. Since the proposed changes involve: 
(1) no significant hardware changes; (2) no significant changes to 
the operation of any systems or components; and (3) no changes to 
existing structures, systems or components, there can be no impact 
on the occurrence of any accident. Furthermore, there is no change 
in plant testing proposed in this change request that could initiate 
an event. Therefore, these changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the HC emergency diesel generator (EDG) 
TS Bases [Change 5--Bases for TS \3/4\.8, ``Electrical Power 
Systems''] include information contained in the Safety Evaluation 
Report for Technical Specification Amendment No. 75. This 
information concerns the bases for the allowed-outage-time (AOT) for 
the C and D EDGs. Concerning the revisions to planned C and D EDG 
outages, PSE&G believes that implementation of 10CFR50.65 
requirements to monitor EDG unavailability will provide an 
acceptable and more clearly defined method for maintaining EDG 
availability within acceptable limits. As stated in PSE&G's letter 
LR-N97167, dated March 21, 1997, Hope Creek will not plan C or D EDG 
outages that exceed 72 hours if the total unavailability of the EDG 
will be greater than 720 hours on a 12 month rolling basis. The 
proposed changes will not adversely impact the operation of any 
safety related component or equipment. Since the proposed changes 
involve: (1) No hardware changes; (2) no significant changes to the 
operation of any systems or components; and (3) no changes to 
existing structures, systems or components, there can be no impact 
on the occurrence of any accident. Furthermore, there is no change 
in plant testing proposed in this change request which could 
initiate an event. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes for the TS related to response time testing 
reflect testing methodologies that were approved by the NRC in 
Amendment No. 85 to the Hope Creek TS. No changes are being made to 
methodologies with this proposal. Therefore, the changes contained 
in this request do not result in a significant reduction in a margin 
of safety.
    The proposed changes to Surveillance Requirement 4.0.5 do not 
alter the current requirements for the Hope Creek inservice 
inspection and inservice testing programs and are considered to be 
editorial in nature. Therefore, the changes contained in this 
request do not result in a significant reduction in a margin of 
safety.
    The proposed changes to the drywell and suppression chamber 
purge system are being made to reflect design modifications that 
have been installed. This design modification replaced isolation 
valves containing resilient material seals with metal seated valves 
under 10 CFR 50.59. PSE&G has concluded that the 24 month frequency 
is appropriate for the new valves since: (1) this frequency is 
imposed by Surveillance Requirement 4.6.1.2.d, which is applicable 
to other containment isolation valves in Table 3.6.3-1 that 
penetrate the primary containment; and (2) concerns raised about 
severe environment-induced degradation and frequent use for the 
previously installed resilient seal material valves are not 
applicable to the replacement metal seat valves. The valve 
modification was an enhancement to the Hope Creek design that did 
not impact the isolation capability of the drywell and suppression 
chamber purge system, and does not result in a significant reduction 
in a margin of safety.
    The proposed changes to LCO 3.7.7 establish consistent and 
appropriate requirements for main turbine bypass valve operability 
requirements. These changes do not impact the assumptions contained 
in these UFSAR analyses since they do not change the manner in which 
Hope Creek is currently permitted to operate. Since the ACTION 
Statement for LCO 3.7.7 already allows indefinite continued 
operation below 25% of RATED THERMAL POWER with an inoperable main 
turbine bypass valve system, the proposed modification to the 
APPLICABILITY statement for this LCO would be editorial in nature. 
Therefore, the changes contained in this request do not result in a 
significant reduction in a margin of safety.
    The HC TS Bases [Change 5--Bases for TS \3/4\.8, ``Electrical 
Power Systems''] will be revised to include information contained in 
the Safety Evaluation Report for Technical Specification Amendment 
No. 75. This information concerns the bases for the allowed-outage-
time (AOT) for the C and D emergency diesel generators (EDGs). PSE&G 
believes that implementation of 10 CFR 50.65 requirements to monitor 
EDG unavailability limits will provide an acceptable and more 
clearly defined method for maintaining EDG availability within 
acceptable limits and not result in 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: 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: May 19, 1997

Description of amendment request: The proposed amendment would change 
Technical Specification (TS) 3.7.1.3, ``Ultimate Heat Sink'' to reflect 
that continued plant operation depends upon the association of ultimate 
heat sink (UHS) temperature and safety system availability. The 
requirements of TS 3.7.1.1, ``Safety Auxiliaries Cooling System 
(SACS)'', TS 3.7.1.2, ``Station Service Water System (SSWS)'' and TS 
3.8.1.1, ``Electrical Power Systems'' would be revised to reflect the 
revised TS 3.7.1.3. In addition, the Bases for \3/4\.7.1, ``Service 
Water Systems'' would be appropriately revised.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS revisions related to SSWS/SACS and the emergency 
diesel generators (EDGs) [TS 3.7.1.1, TS 3.7.1.2, and TS 3.8.1.1] 
involve no hardware changes and no changes to existing structures, 
systems or components. The additional system configuration limits and 
changes to the operation of SSWS/SACS/EDGs are being made to ensure 
that SSWS/SACS can remove required heat loads during design basis 
accidents and transients with the proposed UHS river water temperature 
and level limits. The link to the UHS LCO in the proposed SSWS/SACS/EDG 
TS ACTION Statements and the proposed revisions to the SACS ACTION 
Statement for one inoperable SACS subsystem ensure that the plant is 
directed to enter a safe shutdown condition whenever the capability to

[[Page 33133]]

mitigate design basis accidents and transients is lost. Since the SSWS/
SACS/EDGs will still remain capable of meeting all applicable design 
basis requirements and retaining the capability to mitigate the 
consequences of accidents described in the HC UFSAR, the proposed 
changes were determined to be justified. As a result, these changes 
will not increase the probability of an accident previously evaluated 
nor significantly increase in the consequences of an accident 
previously evaluated.
    The proposed TS revisions related to UHS [TS 3.7.1.3] involve no 
hardware changes and no changes to existing structures, systems or 
components. The additional system configuration limits and changes 
to the operation of UHS supported systems are being made to ensure 
that the UHS can remove required heat loads during design basis 
accidents and transients with the proposed UHS river water 
temperature and level limits. The proposed UHS TS ACTION Statements 
ensure that the plant is directed to enter a safe shutdown condition 
whenever the capability to mitigate design basis accidents and 
transients is lost. The proposed changes to the UHS TS surveillance 
requirements to increase monitoring of the river water temperature 
at 82 deg.F adequately ensures that the actions required when river 
temperatures exceed 85 deg.F are taken as appropriate. Since the UHS 
will still remain capable of meeting all applicable design basis 
requirements and retaining the capability to mitigate the 
consequences of accidents described in the HC UFSAR, the proposed 
changes were determined to be justified. As a result, these changes 
will not increase the probability of an accident previously 
evaluated nor significantly increase in the consequences of an 
accident previously evaluated.
    With the approval of the proposed changes to the SSWS/SACS/EDG/
UHS TS, the proposed TS Bases changes are considered to be editorial 
in nature. As a result, the proposed Bases changes will not increase 
the probability of an accident previously evaluated nor 
significantly increase in the 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.
    The proposed changes to the SSWS/SACS/EDG TS contained in this 
submittal will not adversely impact the operation of any safety 
related component or equipment. Since the proposed changes involve 
no hardware changes and no changes to existing structures, systems 
or components, there can be no impact on the potential occurrence of 
any accident due to new equipment failure modes. The additional 
system configuration limits and changes to the operation of SSWS/
SACS/EDGs imposed by the proposed changes ensure that SSWS/SACS and 
the UHS can remove required heat loads during design basis accidents 
and transients with the proposed UHS river water temperature and 
level limits. Furthermore, there is no change in plant testing 
proposed in this change request which could initiate an event. 
Therefore, these changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes to the UHS TS contained in this submittal 
will not adversely impact the operation of any safety related 
component or equipment. Since the proposed changes involve no 
hardware changes and no changes to existing structures, systems or 
components, there can be no impact on the potential occurrence of 
any accident due to new equipment failure modes. The additional 
system configuration limits imposed by the proposed UHS LCO ensure 
that supported systems can remove required heat loads during design 
basis accidents and transients with the proposed UHS river water 
temperature and level limits. Furthermore, there is no change in 
plant testing proposed in this change request which could initiate 
an event. The proposed changes to the UHS TS surveillance 
requirements to increase monitoring of the river water temperature 
at 82  deg.F adequately ensures that the actions required when river 
temperatures exceed 85  deg.F are taken as appropriate. Therefore, 
these changes will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    With the approval of the proposed changes to the SSWS/SACS/EDG 
UHS TS, the proposed TS Bases changes are considered to be editorial 
in nature. As a result, the proposed Bases changes will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes for the TS related to the SSWS/SACS/EDGs 
establish consistent and appropriate requirements for SSWS/SACS/EDG 
and UHS operability requirements. The additional system 
configuration limits and changes to the operation of SSWS/SACS/EDG 
are being made to ensure that SSWS/SACS can remove required heat 
loads during design basis accidents and transients with the proposed 
UHS river water temperature and level limits. The link to the UHS 
LCO in the proposed SSWS/SACS/EDG TS ACTION Statements and the 
revision to the SACS ACTION Statement for one inoperable SACS 
subsystem ensure that the plant is directed to: (1) enter a safe 
shutdown condition whenever the capability to mitigate design basis 
accidents and transients is lost; or (2) enter a conservatively 
short period of continued operation when system redundancy is 
reduced. Since the SSWS/SACS/EDG will still remain capable of 
meeting all applicable design basis requirements and retaining the 
capability to mitigate the consequences of accidents described in 
the HC UFSAR, the proposed changes contained in this submittal were 
determined to not result in a significant reduction in a margin of 
safety.
    The proposed changes for the TS related to the UHS ensure 
continued capability of the UHS to mitigate the consequences of 
design basis accidents and transients. The additional SSWS/SACS 
configuration limits and changes to the operating limits of the UHS 
ensure that the UHS can remove required heat loads during design 
basis accidents and transients with the proposed river water 
temperature and level limits. The proposed UHS TS ACTION Statements 
ensure that the plant is directed to: (1) enter a safe shutdown 
condition whenever the capability to mitigate design basis accidents 
and transients is lost; or (2) enter a conservatively short period 
of continued operation when supported system redundancy is reduced. 
Since the UHS will still remain capable of meeting all applicable 
design basis requirements and retaining the capability to mitigate 
the consequences of accidents described in the HC UFSAR, the 
proposed changes contained were determined to not result in a 
significant reduction in a margin of safety.
    With the approval of the proposed changes to the SSWS/SACS/UHS 
TS, the proposed TS Bases changes are considered to be editorial in 
nature. As a result, the proposed bases changes will not result in 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 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.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, 
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: May 21, 1997

    Description of amendment request: The proposed amendment would 
revise the Virgil C. Summer Nuclear Station Technical Specifications 
(TS), Surveillance Requirements (SRs), to change the methodology for 
testing the charcoal adsorbers in (1) the control room normal and 
emergency air handling system (TS 3/4.7.6), and (2) the spent fuel pool 
ventilation system (TS 3/4.9.11), by reference to the methodology of 
ASTM D 3803-1989 from the ANSI STD N509-1980.
    The proposed reference testing methodology to ASTM D 3803-1989 for 
the control room is at a relative humidity of 70% and 30 degrees C with 
methyl iodide penetration of < 2.5%. The proposed reference testing 
methodology to ASTM D 3803-1989 for

[[Page 33134]]

the spent fuel pool is at a relative humidity of 95% and 30 degrees C 
with a methyl iodide penetration of < 2.5%.
    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:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change revises the methodology for testing the 
charcoal adsorbers in the Control Room Normal and Emergency Air 
Handling System and the Spent fuel Pool Ventilation System 
(Engineered Safeguards Feature [ESF] air handling units) to the 
updated Standard Test Method for Nuclear-Grade Carbon.* * *. The 
charcoal adsorbers are not initiators of any analyzed event.* * * 
The charcoal adsorbers will be tested to the updated version of the 
approved standard, which generally contains more stringent testing 
requirements. The change does not affect the operation of the ESF 
air handling units. The new testing requirements will continue to 
ensure that the ESF air handling units will be capable of performing 
their safety function and meeting the assumptions in the safety 
analysis [Final Safety Analysis Report (FSAR)]. The change does not 
affect the mitigation capabilities of any component or system nor 
does it affect the assumptions relative to the mitigation of 
accidents or transients. Therefore, the 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 revises the methodology for testing the 
charcoal adsorbers in the Control Room Normal and Emergency Air 
Handling System and the Spent fuel Pool Ventilation System * * * to 
the updated Standard Test Method for Nuclear-Grade Carbon. The 
change does not involve a significant change in the design or 
operation of the plant. The changes do not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed), or new or unusual operator actions. No new or 
different accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this change. Therefore, the 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 margin of 
safety?
    The proposed change revises the methodology for testing the 
charcoal adsorbers in the Control Room Normal and Emergency Air 
Handling System and the Spent fuel Pool Ventilation System * * * to 
the updated Standard Test Method for Nuclear-Grade Carbon. Testing 
of the charcoal adsorbers in the ESF air handling units to the new 
standard will continue to ensure the systems perform their design 
function. The increase in the allowed penetration percentage does 
not affect the accident analysis because testing requirements are 
more stringent and the higher allowed percentages continue to be 
below the assumptions of the safety analysis [FSAR]. Therefore, the 
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: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Project Director: Gordon Edison, Acting.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama
Date of amendments request: May 27, 1997

    Description of amendments request: The proposed amendments would 
revise the applicable Modes for Source Range Nuclear Instrumentation 
(Technical Specification 3/4.3.1, ``Reactor Trip System 
Instrumentation''), provide allowances for an exception to the 
requirements for the state of the power supplies for Residual Heat 
Removal System discharge to charging pump suction valves following Mode 
changes (Technical Specification 3/4.5.2, ``ECCS Subsystems--
Tavg greater than 350 deg.F'' and 3/4.5.3, ``ECCS 
Subsystems--Tavg less than 350 deg.F''), and delete cycle-
specific guidance concerning manual emergency engineered safety feature 
function input checks.
    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 changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the FSAR [Final Safety Analysis Report]. The purposes for 
repositioning the breakers/disconnects for MOVs [motor-operated 
valves] 8706A and 8706B are to ensure that the ECCS [Emergency Core 
Cooling System] System is aligned properly such that the assumptions 
used in the safety analyses are met and to prevent possible 
overpressurization of the charging pump suction line piping. The 
likelihood of a severe transient occurring in this time frame is 
very small and has to be weighed against the possibility of over 
pressurizing the CVCS [Chemical and Volume Control System] charging 
pump suction piping. The allowance of a 4 hour time period to 
perform the required alignment appropriately weighs this risk. 
Changing the applicability of the requirement to have indication 
from a Source Range Nuclear Instrument available to agree with the 
design of the plant does not change the physical design of the plant 
or affect any assumptions used in accident analyses and, therefore, 
has no effect on the probability or consequences of an accident 
previously evaluated in the FSAR. The allowance of 1 hour to perform 
the Source Range Channel Check upon reaching P-6 from Mode 2 is 
consistent with the current basis for a source range channel 
inoperable. Therefore, these changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    (2) The proposed changes to the Technical Specifications do not 
increase the possibility of a new or different kind of accident than 
any accident already evaluated in the FSAR. No new limiting single 
failure or accident scenario has been created or identified due to 
the proposed changes. Safety-related systems will continue to 
perform as designed. Therefore, the proposed changes do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed changes do not involve a significant reduction 
in the margin of safety. The margin of safety is not significantly 
reduced due to the proposed changes to the breaker/disconnect 
positioning requirements of TS [Technical Specifications] 3/4.5.2 
and 3/4.5.3 when transitioning between Modes 3 and 4. The likelihood 
of either a severe transient occurring in Mode 3 or the possible 
overpressurization of the CVCS charging pump suction line by the RHR 
[residual heat removal] system in Mode 4 is very small. Changing the 
Applicability of the requirement to have indication from a Source 
Range Nuclear Instrument available to agree with the design of the 
plant does not change the physical design of the plant or affect any 
assumptions used in accident analyses and, therefore, has no effect 
on the margin of safety. These proposed changes are technically 
consistent with the requirements and standard format of NUREG-1431, 
Revision 1. Performing the source range channel check within 1 hour 
upon reaching P-6 from Mode 2 does not change the physical design of 
the plant or affect any assumptions used in accident analyses and, 
therefore, also does not [a]ffect the margin of safety. Thus, 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

[[Page 33135]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Project Director: Herbert N. Berkow.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama
Date of amendments request: May 28, 1997

    Description of amendments request: The proposed amendments would 
insert a footnote in Technical Specification (TS) Surveillance 
Requirement 4.8.1.1.2.e, to clarify that load rejection testing of the 
shared emergency diesel generator set on either unit may be used to 
satisfy TS 4.8.1.1.2.e surveillance requirements for both units.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes clarify that load rejection testing of the 
shared emergency diesel generator set is only required once per five 
years, and that testing of the shared EDG [emergency diesel 
generator] set on one unit may be used to satisfy SR [Surveillance 
Requirement] 4.8.1.1.2.e requirements for both units. These changes 
do not affect the probability or consequences of an accident. There 
are no changes being made to the emergency diesel generator testing 
program. These changes simply clarify the existing test program and 
the intent of the test requirements.
    Therefore, the proposed TS changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes clarify that load rejection testing of the 
shared emergency diesel generator set is only required once per five 
years, and that testing of the shared EDG set on one unit may be 
used to satisfy SR 4.8.1.1.2.e requirements for both units. No new 
testing configuration is being proposed that could create the 
possibility of any new or different kind of accident from any 
accident previously evaluated. There are no changes being made to 
the emergency diesel generator testing program. These changes simply 
clarify the existing test program and the intent of the test 
requirements.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes clarify that load rejection testing of the 
shared emergency diesel generator set is only required once per five 
years, and that testing of the shared EDG set on one unit may be 
used to satisfy SR 4.8.1.1.2.e requirements for both units. A 
similar technical specification change has been previously approved 
by the NRC for Hatch Nuclear Plant. The technical specification 
bases and the Final Safety Analysis Report have been reviewed. 
Clarification of the testing requirements has no effect on the 
margin of plant safety since no reduction in the test program is 
involved.
    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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Project Director: Herbert N. Berkow.

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 1, Lake County, Ohio
Date of amendment request: May 2, 1997.

    Description of amendment request: The proposed change would 
continue to allow entry into Operational Conditions 1, 2, and 3 with 
the inboard main steam isolation valve (MSIV) leakage control subsystem 
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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This License Amendment application proposes a revision to the 
exception to Limiting Condition for Operation (LCO) 3.0.4 as it 
applies to the Technical Specification (TS) for the MSIV Leakage 
Control System (LCS). This revision is proposed to permit completion 
of activities necessary to implement the most appropriate permanent 
resolution for the issues that resulted from the elimination of the 
secondary containment bypass leakage path through the Main Steam 
Line drains. In addition, the revision clarifies that the exception 
only applies to the Inboard MSIV LCS subsystem. The drains will 
remain in their current configuration, which seals off the secondary 
containment bypass leakage path. The sealed drain path results in a 
temporary inoperability of the Inboard MSIV LCS subsystem when the 
plant is operated below 50 percent rated thermal power (RTP), due to 
condensate build-up in the bottom of the steam lines between the 
MSIVs. The requested 3.0.4 exception is necessary to permit plant 
startups with this temporary inoperability. The exception to LCO 
3.0.4 simply permits use of the existing Action statement (Condition 
A of LCO 3.6.1.9) during MODE changes.
    The probability of occurrence of a previously evaluated accident 
is not affected by the proposed revision of the LCO 3.0.4 exception 
since no change to the plant or to the manner in which the plant is 
operated is involved. The existing plant configuration will be 
maintained, and possible concerns resulting from that configuration 
have been analyzed. The extra weight of the water pooled between the 
MSIVs was analyzed with respect to piping supports and seismic 
considerations and was found to be acceptable, and condensate that 
is carried past the outboard MSIVs will be drained to the condenser 
by drain connections downstream of the outboard MSIVs before it can 
reach the turbine. The temporary inoperability of the Inboard MSIV 
LCS subsystem when below 50 percent RTP has no impact on accident 
initiation probability, since the MSIV LCS does not serve to prevent 
accidents, but is only used in mitigating the consequences of Loss 
of Coolant Accidents (LOCAs) that have already occurred.
    The consequences of an accident are not affected in that the 
Outboard MSIV LCS subsystem will be available to perform the MSIV 
LCS function by mitigating the consequences of a LOCA during the 
temporary period in which the Inboard MSIV LCS subsystem is 
unavailable. Condensate that is carried past the outboard MSIVs will 
be drained to the condenser by drain connections downstream of the 
outboard MSIVs; therefore, no impairment of the Outboard MSIV LCS 
subsystem will result from condensed water. The Required Action and 
Completion Time for one inoperable MSIV LCS subsystem remains the 
same, and limits plant operation to the previously established 30-
day Allowable Outage Time. The Required Action if both subsystems of 
MSIV LCS were to become inoperable also remains the same. The MSIV 
function of isolating the Main Steam Lines is also unaffected by the 
existing plant

[[Page 33136]]

configuration, since MSIV performance will not be affected by the 
existence of accumulated water in the bottom of the steam lines 
between the MSIVs during plant operation below 50 percent RTP. 
Therefore, if necessary, the Main Steam Lines will be isolated, and 
leakage past the MSIVs will be routed for filtration as in the 
design-basis radiological analyses, and the safety and radiological 
consequences of previously evaluated accidents will remain 
unaffected.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to permit inoperability of the Inboard MSIV 
LCS subsystem during periods of startup and power ascension to 50 
percent RTP and during shutdown below 50 percent RTP does not create 
the possibility of a new or different kind of accident from any 
previously evaluated. The Inboard MSIV LCS subsystem is only 
credited during a large-break LOCA wherein Reactor Coolant System 
depressurization occurs. The temporary unavailability of the Inboard 
MSIV LCS subsystem can be mitigated by operation of the Outboard 
MSIV LCS subsystem. The amendment to the TS is an administrative 
change that does not involve change to the current plant design or 
methods of operation. No new plant equipment failure modes or 
accident initiators are introduced by the LCO 3.0.4 exception.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The response to a large-break LOCA will not be affected since 
the Outboard MSIV LCS subsystem can be assumed to be available 
during the limited period of time that the Technical Specifications 
permit the Inboard subsystem to be unavailable. Allowing entry into 
MODES 1, 2, and 3 while utilizing the existing Condition A and 
Required Action A.1 does not reduce the margin of safety since the 
Completion Time allowed for that Condition is not increased. The 
proposed change will have no adverse impact on the reactor coolant 
system pressure boundary nor will other system protective boundaries 
or safety limits be affected.

    The NRC staff has reviewed the licensees' 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.

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 1, Lake County, Ohio
Date of amendment request: May 2, 1997

    Description of amendment request: The proposed change would allow 
the leakage rate of one or more main steam lines to be up to 35 
standard cubic feet per hour (scfh), as long as the total leakage rate 
through all four main steam lines is less than or equal to 100 scfh.
    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 involves the deletion of the portion of 
Technical Specification Surveillance Requirement (SR) 3.6.1.3.10 
that states the increased leakage rate of less than or equal to 35 
scfh for an individual main steam line is only acceptable for 
Operating Cycle 6, and a deletion of the restriction that a main 
steam line leakage rate of less than or equal to 35 scfh is 
acceptable for only one main steam line. The overall main steam line 
leakage limit of less than or equal to 100 scfh for all four main 
steam lines is not being revised.
    The MSIV [main steam isolation valve] leakage is not an 
initiator of an accident, including the steam line rupture accident. 
Therefore, the probability of an accident previously evaluated has 
not changed.
    The consequences of interest are the radiological dose 
consequences following a large-break Loss of Coolant Accident 
(LOCA). This is the event which the regulatory guidance requires to 
be evaluated using the extremely conservative source term 
assumptions of Regulatory Guide 1.3, ``Assumptions Used for 
Evaluating the Potential Radiological Consequences of a Loss of 
Coolant Accident for Boiling Water Reactors.'' Since the overall 
main steam line leakage rate of less than or equal to 100 scfh for 
all four main steam lines is not being revised, the radiological 
consequences of an accident previously evaluated has not changed.
    Therefore, the probability or consequences of an accident 
previously evaluated have not significantly increased.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change does not physically alter the plant or 
systems or equipment in the plant, or the method for operation of 
the plant. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.
    The proposed change does not revise the overall combined leakage 
rate of less than or equal to 100 scfh for all four main steam lines 
that is permitted in the present Specification. It is the combined 
main steam line penetration leakage rate that is assumed in the 
radiological accident analyses. Thus, although individual steam line 
leakage rates may be less than or equal to 35 scfh, as long as 
overall leakage of the four main steam lines is maintained at its 
current value of less than or equal to 100 scfh, the proposed change 
does not reduce the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensees' 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.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia
Date of amendment request: November 9, 1987, as supplemented March 31, 
1988, June 8, 1992 and February 4, 1997

    Description of amendment request: The proposed changes would revise 
the Technical Specifications (TS) for the North Anna Power Station (NA 
1&2). The changes would reformat the operability and surveillance 
requirements for the intermediate range (IR) channels to be consistent 
with NUREG-0452, Revision 4, ``Standard Technical Specifications (STS) 
for Westinghouse Pressurized Water Reactors'' (Fall 1981), which is 
applicable to NA 1&2. Also, the proposed changes would revise the 
nominal IR high flux trip setpoint. The IR nuclear flux trips provide 
backup reactor core protection during reactor startup. There is no 
operating condition under which the IR trip provides sole overpower 
protection. It is a backup trip only, and no credit is taken for the 
trip in the NA 1&2 Updated Final Safety Analysis Report (UFSAR). 
Operating experience at NA 1&2 has shown the IR channel response to be 
sensitive to core loading patterns, varying core burnups, and control 
rod positions, and the variability in the channel response had made it 
difficult to maintain the channels in proper calibration. Therefore, 
the proposed change would

[[Page 33137]]

elevate the nominal IR high flux trip setpoint from a current 
equivalent to 25% of rated thermal power to a current equivalent to 35% 
of 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:

[The proposed changes would not:]
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. There is no 
adverse impact on the safety analysis (since no credit is taken for 
the trips in the existing analyses), and no degradation of the 
protection system redundancy or reliability. This latter conclusion 
is based on sensitivity studies which show that the effectiveness of 
the flux trip system in protecting against the low power reactivity 
excursions examined in the FSAR is not sensitive to realistic 
variations in the actual flux trip setpoint.
    2. Create the probability of a new or different kind of accident 
from any accident previously identified, since the severity of the 
analyzed accidents is unchanged, and since only a change to a 
setpoint and the associated surveillance requirements for the 
reactor protection system is involved.
    3. Involve a significant reduction in a margin of safety, since 
none of the safety analysis input or assumptions are changed, nor 
are the probability nor the consequences of any previously analyzed 
accidents increased.

    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: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Brenda Mozafari (Acting).

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 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.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York 
Date of application for amendment: March 31, 1997

    Brief description of amendment: The proposed amendment would remove 
containment isolation valve 863 from Technical Specification Table 3.6-
1, ``Non-Automatic Containment Isolation Valves Open Continuously or 
Intermittently for Plant Operation.''
    Date of publication of individual notice in Federal Register: May 
15, 1997 (62 FR 26823).
    Expiration date of individual notice: June 16, 1997.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey
Date of amendment request: April 25, 1997

    Brief description of amendment request: The proposed amendment 
changes to revise Technical Specification 3.5.2 to eliminate the flow 
path from the residual heat removal system to the reactor coolant 
system hot legs that is specified in Limiting Condition for Operation 
3.5.2.c.2.
    Date of publication of individual notice in Federal Register: May 
14, 1997 (62 FR 26574).
    Expiration date of individual notice: June 13, 1997.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear 
Generating Station, Unit No. 2, Salem County, New Jersey
Date of amendment request: May 1, 1997

    Brief description of amendment request: The proposed amendment 
would revise Technical Specification (TS) 3/4.7.7, ``Auxiliary Building 
Exhaust Air Filtration System,'' and add a new TS Section 3/4.7.11, 
``Switchgear and Penetration Area Ventilation System.'' The change to 
TS 3/4.7.7 would allow for an increase in the allowed outage time from 
7 to 14 days when one auxiliary building exhaust fan is inoperable. The 
new TS 3/4.7.11 addresses the support function this system provides to 
other necessary safety support components.
    Date of publication of individual notice in Federal Register: May 
15, 1997 (62 FR 26826).
    Expiration date of individual notice: June 16, 1997.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey
Date of amendment request: May 14, 1997

    Brief description of amendment request: Your application proposes 
changes to revise Technical Specification Surveillance Requirement 
4.7.6.1.d.1 to indicate that the specified acceptable filter 
differential pressure (DP) is to be measured across the filter housing 
and to change the filter DP acceptance value from less than or equal to 
3.5 inches water gauge to less than or equal to 2.70 inches water 
gauge.
    Date of publication of individual notice in Federal Register: May 
29, 1997 (62 FR 29158).
    Expiration date of individual notice: June 30, 1997.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

[[Page 33138]]

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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts
Date of application for amendment: November 26, 1997

    Brief description of amendment: The amendment revises Technical 
Specifications Definition 1.M, ``Primary Containment Integrity,'' Note 
6 on Table 3.2.A for the high flow main steam line instrumentation, 
Table 3.2.D for a typographical error, Table 3.2.F to reflect a change 
made in instrument type for the suppression chamber water temperature 
instrumentation, Table 3.2.F to reflect modifications made to 
suppression chamber bulk and local temperature instrumentation, Bases 
Section 3/4.6G to remove an obsolete reference to Group I welds, and 
Bases Section 3/4.7.A to remove ``high radiation'' from the description 
of Primary Containment Group 1 initiation signals. In addition, this 
amendment includes changes made to the Bases Section 3.10, ``Core 
Alterations,'' as noted by BECo letter dated March 7, 1997.
    Date of issuance: May 28, 1997.
    Effective date: May 28, 1997.
    Amendment No.: 172.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6568). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 28, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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 application for amendments: June 10, 1996, as supplemented by 
letter dated February 17, 1997

    Brief description of amendments: The amendments change the 
Technical Specifications to reflect the transition from General 
Electric Company (GE) to Siemens Power Corporation (SPC) as the fuel 
supplier for the Quad Cities Nuclear Power Station, Units 1 and 2. In 
addition, as an administrative action by the Commission that only 
involves the format of the licenses and does not authorize any 
activities outside the scope of the application and supplement, the NRC 
has amended the licenses to include an Appendix C that lists additional 
license conditions. The additional license condition as a result of the 
review of this application reflects the relocation of the contents of 
TS 5.4 to the Updated Final Safety Analysis Report.
    Date of issuance: May 23, 1997.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 177 and 175.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Licenses, Technical Specifications and Updated Final Safety 
Analaysis Report.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44355). The February 17, 1997, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 23, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: August 29, 1995, as supplemented 
August 7, 1996, and January 10, 1997

    Brief description of amendment: The amendment revises Technical 
Specifications to incorporate the commitments made in connection with 
Amendment No. 183, which allowed the installation of laser welded 
sleeves inside of defective steam generator tubes.
    Date of issuance: May 20, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 192.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56365) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 20, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut
Dates of application for amendment: December 24, 1996, and January 31, 
1997

    Brief description of amendment: Changes Administrative Controls 
Section of the Technical Specifications to implement revised management 
responsibilities and titles that reflect the permanently shut down 
status of the plant.
    Date of issuance: May 22, 1997.
    Effective date: Effective May 22, 1997, to be implemented within 60 
days of issuance.
    Amendment No.: 191.
    Operating License No. DPR-61: Amendment revised the Technical 
Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14460) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 22, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of application for amendments: March 10, 1997

    Brief description of amendments: These amendments modify Unit No. 1 
Technical Specification (TS) 5.2.1 to add ZIRLO as fuel assembly 
material

[[Page 33139]]

and add reference to the Nuclear Regulatory Commission approved Topical 
Report WCAP-12610, ``Vantage+ Fuel Assembly Reference Core Report,'' to 
TS 6.9.1.12 for both units.
    Date of issuance: May 23, 1997.
    Effective date: Both units, as of date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 203 and 84.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17231) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 23, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 5, 1997, as supplemented by letter 
dated March 26, 1997

    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications for Waterford Steam Electric Station, Unit 
3, by revising Technical Specifications 3.1.2.7, 3.1.2.8, 3.5.1, 3.5.4, 
3.9.1, and Bases 3/4.1.2. The changes will increase the minimum boron 
concentration in the Safety Injection Tanks and the Refueling Water 
Storage Pool from 1720 to 2050 ppm.
    Date of issuance: May 29, 1997, to be implemented within 60 days.
    Effective date: May 29, 1997.
    Amendment No.: 129.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997, (62 FR 
14461) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 29, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: June 28, 1996, as supplemented March 
11, 1997

    Brief description of amendment: The amendment revises Three Mile 
Island, Unit 1, Technical Specifications to permit the use of 10 CFR 
50, Appendix J, Option B, Performance-Based Containment Leakage 
Testing.
    Date of issuance: May 27, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 201.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40019) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 27, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY), Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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: August 8, 1996

    Brief description of amendments: The amendments allowed the 
transition from Mode 4 to Mode 3 with the turbine-driven auxiliary 
feedwater pump inoperable and allowed a 72-hour period after the entry 
into Mode 3 to complete all necessary operability testing.
    Date of issuance: May 27, 1997.
    Effective date: May 27, 1997, to be implemented within 30 days.
    Amendment Nos.: Unit 1--Amendment No. 87; Unit 2--Amendment No. 74.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44359) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 27, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear 
Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: March 6, 1997

    Brief description of amendment: The amendment revises the Technical 
Specifications on allowed outage times for certain protective 
instrumentation and also for reactor building access control. The 
amendment adopts, in part, guidance from NUREG-0123, ``Standard 
Technical Specifications for General Electric Boiling Water Reactors 
(BWR/5),'' Revision 3, and NUREG-1433, ``Standard Technical 
Specifications General Electric Plants BWR/4,'' Revision 1.
    Date of issuance: May 28, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 90 days.
    Amendment No.: 101.
    Facility Operating License No. DPR-21: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14462) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 28, 1997.
    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, Connecticut 06360 and at the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut
    Date of application for amendment: March 31, 1997

    Brief description of amendment: The amendment modifies Technical 
Specification Surveillance 4.7.1.2.1.b, which requires the testing of 
the auxiliary feedwater motor-driven and turbine-driven pumps on 
recirculation flow at least once per 92 days. The amendment also makes 
changes to the appropriate Bases section.
    Date of issuance: May 29, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 139.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19832) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 29, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center,

[[Page 33140]]

Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: March 31, 1997

    Brief description of amendment: The amendment separates the 
required testing of motor-operated valve thermal overload protection 
into two new surveillances.
    Date of issuance: May 29, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 140.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19833) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 29, 1997.
    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, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.

Portland General Electric Company, et al., Docket No. 50-344, Trojan 
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 16, 1997, as supplemented on 
February 24, 1997

    Brief description of amendment: This amendment revises the license 
to delete the prohibition on moving a spent fuel assembly shipping cask 
into the Fuel Building.
    Date of issuance: May 19, 1997.
    Effective date: This license amendment is effective as of the date 
of issuance (May 19, 1997), but shall be implemented within 30 days of 
issuance.
    Amendment No.: 196.
    Facility Operating License No. NPF-1: The amendment revised the 
license.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14467).
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.

Portland General Electric Company, et al., Docket No. 50-344, Trojan 
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 28, 1997

    Brief description of amendment: This amendment changes the 
Permanently Defueled Technical Specifications to delete the requirement 
for NRC prior approval to changes in the Certified Fuel Handler's 
Training Program.
    Date of issuance: May 23, 1997.
    Effective date: May 23, 1997.
    Amendment No.: 197.
    Possession-Only License No. NPF-1: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17241).
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.

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 15, 1997

    Brief description of amendments: These amendments revise 
Surviellance Requirement 3.8.1.8 of Technical Specifications (TS) 
3.8.1, ``AC Sources--Operating,'' for San Onofre Nuclear Generating 
Station (SONGS), Units 2 and 3. The TS change will allow the licensee 
to credit overlap testing to validate the capability of the alternate 
offsite power source.
    Date of issuance: June 2, 1997.
    Effective date: June 2, 1997.
    Amendment Nos.: Unit 2--136; Unit 3--128.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 1, 1997 (62 FR 
23811) The Commission's related evaluation of the amendments is 
contained in a Safety E valuation dated June 2, 1997.
    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 No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee
Date of application for amendment: January 10, 1997, as supplemented 
May 2 and May 15, 1997

    Brief description of amendment: The amendment modifies the Watts 
Bar Nuclear Plant (WBN) Unit 1 Technical Specifications (TS) in order 
to implement 10 CFR Part 50, Appendix J, Option B, by referring to 
Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test 
Program.''
    Date of issuance: May 27, 1997.
    Effective date: May 27, 1997.
    Amendment No.: 5.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4356) The May 2 and May 15, 1997 letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 27, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas
Date of amendment request: March 18, 1997

    Brief description of amendment: This amendment revises Technical 
Specification Surveillance Requirement 4.5.2.c to clarify when a 
containment entry visual inspection is required. This change reduces 
the visual inspection requirement to at least once daily and is in 
accordance with the guidance provided in Generic Letter 93-05, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation.''
    Date of issuance: May 28, 1997.
    Effective date: May 28, 1997, to be implemented within 30 days of 
the date of issuance.
    Amendment No.: 105.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19839) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 28, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University,

[[Page 33141]]

William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

Notice of Issuance of Amendments To Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By July 18, 1997, 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 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

[[Page 33142]]

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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 20555-0001, 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. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, 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 the 
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket No. STN 50-456, Braidwood Station, 
Unit No. 1, Will County, Illinois
Date of application for amendment: Two submittals dated May 23, 1997
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 4.5.2.b.1 to include the use of ultrasonic testing 
(UT) to verify that the emergency core cooling system (ECCS) is 
completely filled with water. For the ECCS subsystems with high point 
vent valves in direct communication with the operating systems, UT is 
acceptable in lieu of physically opening the vents.
    Date of Issuance: May 23, 1997.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 83.
    Facility Operating License No. NPF-72: The amendment revised the 
TSs.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated May 
23, 1997.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690.
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.
    NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket No. STN 50-454, Byron Station, Unit 
No. 1, Ogle County, Illinois
Date of application for amendment: May 24, 1997, as supplemented on May 
31, 1997

    Brief description of amendment: The amendment revises Technical 
Specification 4.5.2.b.1 to include the use of ultrasonic testing (UT) 
to verify that the emergency core cooling system (ECCS) is completely 
filled with water. For the ECCS subsystems with high point vent valves 
in direct communication with the operating systems, UT is acceptable in 
lieu of physically opening the vents. This amendment supersedes NOED 
No. 97-6-010 for Byron, Unit 1, which was granted on May 23, 1997.
    Date of Issuance: June 1, 1997.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 90.
    Facility Operating License No. NPF-37: The amendment revised the 
TS. Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated June 
1, 1997.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690.
    Local Public Document Room location: Byron Public Library District, 
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
    NRC Project Director: Robert A. Capra.

    Dated at Rockville, Maryland, this 11th day of June, 1997.

    For The Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation.
[FR Doc. 97-15827 Filed 6-17-97; 8:45 am]
BILLING CODE 7590-01-P