[Federal Register Volume 60, Number 108 (Tuesday, June 6, 1995)]
[Notices]
[Pages 29869-29896]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-13759]



[[Page 29869]]

NUCLEAR REGULATORY COMMISSION


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

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 15, 1995, through May 25, 1995. The last 
biweekly notice was published on Tuesday, May 23, 1995 (60 FR 27334).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By July 7, 1995, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be 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 [[Page 29870]] limitations in the order granting leave 
to intervene, and have the opportunity to participate fully in the 
conduct of the hearing, including the opportunity to present evidence 
and cross-examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment requests: March 24, 1995.
    Description of amendment requests: The proposed amendments would 
make numerous changes to Technical Specification (TS) 3/4.8.1, ``A.C. 
Sources,'' and the associated TS Bases, for Palo Verde Units 1, 2, and 
3. The proposed amendments would implement recommended changes from 
NUREG-1432, ``Standard Technical Specifications: Combustion Engineering 
Plants''; Generic Letter (GL) 94-01, ``Removal of Accelerated Testing 
and Special Reporting Requirements for Emergency Diesel Generators''; 
and GL 93-05, ``Line-Item Technical Specification Improvements to 
Reduce Surveillance Requirements for Testing During Power Operation.'' 
The proposed changes are intended to increase emergency diesel 
generator (EDG) reliability by reducing the stresses on the EDGs from 
unnecessary testing. Additional changes have also been proposed to TS 
3/4.8.1 to further enhance EDG reliability, to achieve consistency with 
NUREG-1432, Combustion Engineering Standard TS, and to improve the TS 
presentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to TS 3/4.8.1 and the associated Bases 
affect the required actions in response to inoperable offsite and 
onsite AC sources, surveillance requirements for the EDG, and 
reporting requirements for EDG failures. The majority of the 
proposed changes are based on the recommendations of NUREG 1432, GL 
94-01, and GL 93-05. These proposed changes have been extensively 
reviewed by the NRC during the preparation of these documents, and 
by APS during the development of this request for TS amendment. The 
proposed changes are expected to result in improvements in EDG 
performance and reduce EDG aging due to excessive testing. The 
proposed changes will permit the elimination of the unnecessary 
mechanical stress and wear on the EDGs while ensuring that the EDGs 
will perform their design function. The elimination of mechanical 
stress and wear will improve reliability and availability of the 
EDGs which will have a positive effect on the ability of the EDGs to 
perform their design function. The proposed changes to [do] not 
affect the availability or the testing requirements of the offsite 
circuits.
    Because the proposed changes do not affect the design or 
performance of the EDGs or their ability to perform their design 
function, the changes are expected to result in a decrease in the 
probability or consequences of an accident previously evaluated. The 
proposed changes will increase EDG reliability, thereby increasing 
overall plant safety. Because these changes do not affect the 
probability of accident precursors (EDGs do not initiate any 
accidents), the proposed type license amendment does not involve a 
significant increase in the probability or consequence 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 TS 3/4.8.1 and the associated Bases do 
not introduce any new modes of plant operation or new accident 
precursors, involve any physical alterations to plant 
configurations, or make any changes to system setpoints which could 
initiate a new or different kind of accident. The proposed changes 
do not affect the design or performance characteristics of any EDG 
or its ability to perform its design function. No new failure modes 
have been defined nor new system interactions introduced for any 
plant system or component, nor has any new limiting failure been 
identified as a result of the proposed changes. The proposed changes 
will eliminate unnecessary EDG testing, increasing EDG reliability 
and availability, and thereby having an overall positive affect on 
plant safety. Accidents concerning loss of offsite power and a 
single failure (e.g., loss of an EDG) have previously been 
evaluated. These changes are intended to improve plant safety, 
decrease equipment degradation, and remove unnecessary burden on 
personnel resources by reducing the amount of testing that the TS 
requires during power operation. Therefore, the proposed license 
amendment does not create the possibility of a new or different kind 
of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Under the proposed changes to TS 3/4.8.1 and the associated 
Bases, the EDGs will remain capable of performing their safety 
function. The changes do not affect the design or performance of any 
EDG, but will increase EDG reliability and availability by reducing 
the stresses and the effects of aging on the EDG by eliminating 
unnecessary testing. This will result in an overall increase in 
plant safety. Since the ability of the EDGs to perform their safety 
function will not be degraded, the proposed license amendment does 
not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that [[Page 29871]] review, it appears that the three standards of 
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: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.
    Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

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

    Date of amendment requests: March 31, 1995.
    Description of amendment requests: The proposed amendment would 
clarify the shutdown margin definition, change the shutdown margin 
applicability and surveillance requirements to comply with safety 
analysis assumptions for subcritical inadvertent control element 
assembly withdrawal (UFSAR Section 15.4), and expand the applicability 
for core protection calculator (CPC) operability. In addition, the 
proposed amendment would add a reference to the Core Operating Limits 
Report (COLR) for the MODE 6 refueling boron concentration limit. The 
proposed amendment would also change the power calibration requirements 
for the linear power level, the CPC delta T power, and CPC nuclear 
power signals to allow more conservative settings than presently 
required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis about 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.
    These changes are being made to ensure compliance with the 
safety analysis assumptions for subcritical inadvertent CEA [control 
element assembly] withdrawal. These changes also ensure that the 
boron concentration in the reactor is sufficient to prevent 
criticality if an inadvertent withdrawal of a shutdown CEA bank were 
to occur with all other CEAs inserted. Therefore, the consequences 
of the inadvertent CEA withdrawal is no greater than those of the 
event previously evaluated. This change also has no affect on the 
probability of an accident since it is not introducing or changing 
any accident initiating mechanism.
    The analysis of uncontrolled CEA withdrawal from MODES 2 and 3 
subcritical with four RCPs [reactor coolant pumps] running is 
presented in UFSAR Section 15.4.1 as an anticipated operational 
occurrence. The consequences of this event are that the acceptable 
fuel design limits are not exceeded (General Design Criterion 25 as 
specified in the NRC Standard Review Plan). The proposed change to 
TS requiring that either the CPCs or Logarithmic Power Level--High 
trip (trip setpoint lowered to 10-4% of Rated Thermal Power) 
are Operable in MODES 3, 4, and 5, ensures that an inadvertent CEA 
withdrawal with less than four pumps operating, results in 
consequences no greater than those of the previously evaluated 
uncontrolled CEA withdrawal event.
    The revised TS will also ensure that the reactivity worth of any 
full-length CEAs not capable of being inserted is accounted for in 
the determination of the shutdown margin. This change will ensure 
the shutdown margin will continue to be within safety analysis 
assumptions for previously evaluated accidents.
    The proposed changes to TS, replacing the MODE 6 boron 
concentration specification with the requirement to maintain the 
boron concentration within the limit specified in the COLR, will not 
affect the probability or consequences of an accident, because it is 
not changing the MODE 6 reactivity requirement of Keff less 
than or equal to 0.95, but provides a specific boron concentration 
value in the COLR to ensure the MODE 6 required Keff value of 
less than or equal to 0.95 is met.
    The proposed changes will reduce the amount of non-conservatism 
presently allowed for the linear power level, the CPC delta T power 
and CPC nuclear power signals. Changing the tolerance range from 
plus or minus 2% to between -0.5% and 10% between 15% and 80% RATED 
THERMAL POWER, except during initial post refueling power ascension 
and restricting recalibration, will allow more conservative settings 
than currently required.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    The changes revising the mode applicabilities are being made to 
comply with safety analysis assumptions for subcritical CEA 
withdrawal. The SR [surveillance requirement] ensures that the 
shutdown margin is within the safety analysis assumptions when the 
reactor trip breakers are open and any full-length CEA is not fully 
inserted. No new or different kind of accident will be initiated 
since this change will ensure that the required shutdown margin is 
maintained when the reactor trip breakers are closed.
    The proposed change to TS, requiring either the CPCs or 
Logarithmic Power Level--High trip to be operable, will provide 
protection from inadvertent CEA withdrawal when less than four RCPs 
are operating. No new or different kind of accident will be 
initiated by this change, since this change incorporates TS 
limitations to ensure protection for an existing accident scenario.
    The revised TS shutdown margin definition ensures that the 
reactivity worth of any full-length CEAs not capable of being 
inserted is accounted for in the determination of the shutdown 
margin. This ensures the shutdown margin will continue to be within 
safety analysis assumptions. Maintaining the shutdown margin within 
the safety analyses assumption will not create any new or different 
kind of accident.
    The proposed changes to TS power calibration tolerance limits 
are conservative relative to the current TS requirements and 
therefore will not create any new or different kind of accident.
    The proposed change to TSs replacing the MODE 6 boron 
concentration specification with the requirement to maintain the 
boron concentration within the limit specified in the COLR does not 
create the possibility of a new or different kind of accident from 
any accident previously analyzed. The proposed change is not 
changing the MODE 6 reactivity requirements of less than or equal to 
0.95 while providing a specific boron concentration value in the 
COLR to ensure the MODE 6 required Keff value of less than or 
equal to 0.95.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed change to TS adds an additional requirement for the 
CPCs or Logarithmic Power Level--High trip to be operable in MODES 
3, 4, and 5. This change maintains the margin of safety in the 
safety analysis by providing a TS that will ensure appropriate 
protection is provided in the event of an inadvertent CEA withdrawal 
with less than four RCPs operating.
    The proposed changes to TS (Boration Control, Shutdown Margin), 
revising the mode applicabilities, maintains the margin of safety 
provided in the TS by ensuring that the safety analysis assumptions 
for subcritical CEA withdrawal are met. The new SR does not reduce 
the margin of safety since the shutdown margin assumed in the safety 
analysis will be maintained by this TS.
    The revised TS shutdown margin definition ensures that the 
reactivity worth of any full length CEAs not capable of being 
inserted is accounted for in the determination of shutdown margin. 
This ensures shutdown margin will continue to be within safety 
analysis assumptions. This change maintains the margin of safety 
that is currently provided by TS.
    The proposed changes to TS, reducing the amount of non-
conservatism in the safety system power indications, maintains the 
margin of safety for design basis events which take credit for the 
linear power level, the CPC delta T power, and CPC nuclear power 
signals.
    The proposed change to TS moves the specific MODE 6 boron 
concentration value to COLR. The proposed change does not change the 
MODE 6 reactivity requirement of Keff of less than or equal to 
0.95, but provides a specific boron concentration value in the COLR 
to ensure the MODE 6 required Keff value of less than or equal 
to 0.95 is met. Therefore, the margin of safety is not affected by 
the proposed change.

     [[Page 29872]] The NRC staff has reviewed the licensees' analysis 
and, based on that review, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment requests involve no significant hazards 
consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.
    Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

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

    Date of application for amendment request: December 8, 1992, as 
supplemented on September 10, 1993, and May 17, 1995.
    Description of amendment request: As a result of findings by a 
Diagnostic Evaluation Team inspection performed by the NRC staff at the 
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company 
(ComEd, the licensee) made a decision that both Dresden Nuclear Power 
Station and sister site Quad Cities Nuclear Power Station, needed 
attention focused on the existing custom Technical Specifications (TS) 
used.
    The licensee made the decision to initiate a Technical 
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities. 
The licensee evaluated the current TS for both Dresden and Quad Cities 
against the Standard Technical Specifications (STS) contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' The licensee's evaluation identified numerous potential 
improvements such as clarifying requirements, changing TS to make them 
more understandable and to eliminate interpretation, and deleting 
requirements that are no longer considered current with industry 
practice. As a result of the evaluation, ComEd has elected to upgrade 
both Dresden and Quad Cities TS to the STS contained in NUREG-0123.
    The TSUP for Dresden and Quad Cities is not a complete adaptation 
of the STS. The TSUP focuses on (1) Integrating additional information 
such as equipment operability requirements during shutdown conditions, 
(2) clarifying requirements such as limiting conditions for operations 
and action statements utilizing STS terminology, (3) deleting 
superseded requirements and modifications to the TS based on the 
licensee's responses to Generic Letters (GLs), and (4) relocating 
specific items to more appropriate TS locations.
    The December 8, 1992, application, as supplemented on September 10, 
1993, and May 17, 1995, proposed to upgrade only Section 3/4.1 (Reactor 
Protection System) of the Dresden and Quad Cities TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. 
Implementation of these changes will provide increased reliability 
of equipment assumed to operate in the current safety analysis, or 
provide continued assurance that specified parameters remain within 
their acceptance limits, and as such, will not significantly 
increase the probability or consequences of a previously evaluated 
accident.
    Some of the proposed changes to the current Technical 
Specifications (CTS) represent minor curtailments of the current 
requirements which are based on generic guidance or previously 
approved provisions for other stations. The proposed amendment for 
Dresden and Quad Cities Station's Technical Specification Section 3/
4.1 are based on BWR-STS (NUREG-0123, Revision 4 ``Standard 
Technical Specifications General Electric Plants BWR/4) guidance or 
NRC accepted changes at later operating BWR plants. Any deviations 
from BWR-STS and CTS requirements do not significantly increase the 
probability or consequences of any previously evaluated accident for 
Dresden and Quad Cities Station. These proposed changes are 
consistent with the current safety analyses and have been previously 
determined to represent sufficient requirements for the assurance 
and reliability of equipment assumed to operate in the safety 
analysis, or provide continued assurance that specified parameters 
remain within their acceptance limits. As such, these changes will 
not significantly increase the probability or consequences of a 
previously evaluated accident.
    The associated systems that make up the Reactor Protection 
System are not assumed in any safety analysis to initiate any 
accident sequence for both Dresden and Quad Cities Stations; 
therefore, the probability of any accident previously evaluated is 
not increased by the proposed amendment. In addition, the proposed 
surveillance requirements for the proposed amendments to these 
systems are generally more prescriptive than the current 
requirements specified within the Technical Specifications. These 
more prescriptive surveillance requirements increase the probability 
that the Reactor Protection System will perform its intended 
function. Therefore, the proposed TS will improve the reliability 
and availability of all affected systems and reduce the consequences 
of any accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. These changes do not involve revisions to the design 
of the station. Some of the changes may involve revision in the 
operation of the station; however, these changes provide for 
additional restrictions which are in accordance with the current 
safety analyses, or are to provide for additional testing or 
surveillances which will not introduce new failure mechanisms beyond 
those already considered in the current safety analyses. Therefore, 
these changes will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment for Dresden and Quad Cities Station's 
Technical Specification Section 3/4.1 is based on BWR-STS guidelines 
or NRC accepted changes at later operating BWR plants. The proposed 
amendment has been reviewed for acceptability at the Dresden and 
Quad Cities Nuclear Power Stations considering similarity of system 
or component design versus the BWR-STS or later operating BWRs. Any 
deviations from BWR-STS or CTS requirements do not create the 
possibility of a new or different kind of accident than previously 
evaluated for Dresden and Quad Cities Stations. No new modes of 
operation are introduced by the proposed changes. Surveillance 
requirements are changed to reflect improvements in technique, 
frequency of performance or operating experience at later plants. 
Proposed changes to action statements in many places add 
requirements that are not in the present technical specifications or 
adopt requirements that have been used at other operating BWRs with 
design similar to Dresden and Quad Cities. The proposed changes 
maintain at least the present level of operability. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The associated systems that make up the Reactor Protection 
System are not assumed in any safety analysis to initiate any 
accident sequence for Dresden and Quad Cities Stations. In addition, 
the proposed surveillance requirements for affected 
[[Page 29873]] systems associated with the Reactor Protection System 
are generally more prescriptive than the current requirements 
specified within the Technical Specifications; therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    (3) Involve a significant reduction in the margin of safety 
because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. Some of the later individual items may introduce 
minor reductions in the margin of safety when compared to the 
current requirements. However, other individual changes are the 
adoption of new requirements which will provide significant 
enhancement of the reliability of the equipment assumed to operate 
in the safety analysis, or provide enhanced assurance that specified 
parameters remain within their acceptance limits. These enhancements 
compensate for the individual minor reductions, such that taken 
together, the proposed changes will not significantly reduce the 
margin of safety.
    The proposed amendment to Technical Specification Section 3/4.1 
implements present requirements, or the intent of present 
requirements in accordance with the guidelines set forth in the BWR-
STS. Any deviations from BWR-STS and CTS requirements do not 
significantly reduce the margin of safety for Dresden and Quad 
Cities Stations. The proposed changes are intended to improve 
reliability, usability, and the understanding of technical 
specification requirements while maintaining acceptable levels of 
safe operation. The proposed changes have been evaluated and found 
acceptable for use at Dresden and Quad Cities based on system 
design, safety analysis requirements and operational performance. 
Since the proposed changes are based on NRC accepted provisions at 
other operating plants that are applicable at Dresden and Quad 
Cities and maintain necessary levels of system or component 
readability, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The proposed amendment for Dresden and Quad Cities Stations will 
not reduce the availability of systems associated with the Reactor 
Protection System when required to mitigate accident conditions; 
therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: For Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of amendment request: July 29, 1993.
    Description of amendment request: The proposed amendment would 
extend the instrument calibration intervals for selected plant 
instrumentation from 18 months to 36 months.
    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 change to extend to 36 months the calibration 
interval of selected instrumentation does not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The purpose of the 
proposed Technical Specification change is to extend calibration 
interval testing requirements for selected instrumentation. However, 
because of the continued application of redundant Technical 
Specification requirements such as channel checks, channel 
functional tests, and logic system functional tests, the performance 
of these instruments will be maintained within the acceptance limits 
assumed in plant safety analyses and required for the successful 
mitigation of an initiating event. The proposed Technical 
Specification changes do not affect the capability of the associated 
systems to perform their intended function within their instrument 
settings.
    These other tests are sufficient to identify failure modes or 
degradations in instrument performance and ensure operation of the 
associated systems within acceptance limits. There are no credible 
failure modes that can be detected by instrument calibration that 
cannot also be detected by other Technical Specification tests.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated. As discussed above, 
the proposed Technical Specification changes do not affect the 
capability of the associated systems to perform their intended 
function within the acceptance limits assumed in plant safety 
analyses and required for successful mitigation of an initiating 
event. All plant systems continue to operate in an identical manner. 
No new accident modes are created.
    (3) Involve a significant reduction in a margin of safety. The 
current Technical Specification allowable values are based on the 
maximum analytical limits assumed in the plant safety analyses. 
These analyses conservatively establish the margin of safety. The 
proposed Technical Specification changes do not affect the 
capability of the associated systems to perform their function 
within the instrument settings used as the basis for the plant 
safety analyses. Plant and system settings to an initiating events 
will remain in compliance within the assumptions of the safety 
analyses, and therefore the margin of safety is not affected.

    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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter, Acting.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of amendment request: December 15, 1994.
    Description of amendment request: The proposed amendment would 
relocate, revise, or delete various Technical Specification (TS) 
provisions. Administrative controls on working hours in TS 6.2.2.f, the 
Independent Safety Engineering Group requirements in TS 6.2.3, the unit 
staff qualification requirements in TS 6.3, the reportable event 
requirement for the Onsite Review Organization (OSRO) in TS 6.6.1.b, 
the radiation protection program requirements in TS 6.11, the record 
retention requirements in TS 6.10, and the review and audit functions 
in TS 6.5 (with the exception of TS 6.5.2.8), would be relocated to 
Chapter 13 of the Updated Final Safety Analysis Report (UFSAR). The 
review and approval process for temporary changes to each TS 6.8.1 
plant procedure listed in TS 6.8.4 would also be relocated to Chapter 
13 of the UFSAR.
    The requirements of TS 6.5.2.8, the review and approval process for 
administrative procedures in TS 6.8.2, and the review and approval 
process for plant procedures in TS 6.8.3, would be relocated to the 
Fermi 2 Quality Assurance program. The in-plant radiation monitoring 
program requirements in TS 6.8.5.b, and the high radiation area 
requirements in TS 6.12 would be relocated to Chapter 12 of the 
[[Page 29874]] UFSAR. The radiological environmental monitoring program 
requirements in TS 6.8.5.f would be relocated to Chapter 11 of the 
UFSAR. The Process Control Program (PCP) requirements in TS 6.13 would 
be relocated to the PCP.
    The requirements for OSRO to review the Security Plan in TS 
6.5.1.6.j and to have Security Plan implementing procedures in TS 
6.8.1.e would be relocated to the Fermi 2 Security Plan. The 
requirements for OSRO to review the Emergency Plan in TS 6.5.1.6.k and 
to have Emergency Plan implementing procedures in TS 6.8.1.f would be 
relocated to the Fermi 2 Emergency Plan.
    The unit staff qualification requirements, as specified in the H. 
R. Denton (NRC) letter of March 29, 1980, in TS 6.3, would be deleted. 
The licensee states these have been superseded by 10 CFR Part 55 and 
Generic Letter (GL) 87-07. The training requirements in TS 6.4 would be 
deleted. The licensee states that other Section 6.0 TS and NRC 
regulations provide sufficient control of these training requirements. 
The submittal requirement of the annual radioactive effluent release 
report in TS 6.9.1.8 would be revised from ``within 90 days after 
January 1 * * *'' to ``prior to May 1. * * *''
    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 because the proposed changes are administrative in nature. 
None of the proposed changes involve a physical modification to the 
plant, a new mode of operation or a change to the UFSAR transient 
analyses. No Limiting Condition for Operation, ACTION statement or 
Surveillance Requirement is affected by any of the proposed changes. 
Also, these proposed changes, in themselves, do not reduce the level 
of qualification or training such that personnel requirements would 
be decreased. Therefore, this change is administrative in nature and 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. Further, the 
proposed changes do not alter the design, function, or operation of 
any plant component and therefore, do not affect the consequences of 
any previously evaluated accident.
    (2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because the proposed changes do not introduce a new mode of plant 
operation, surveillance requirement or involve a physical 
modification to the plant. The proposed changes are administrative 
in nature. The changes propose to revise, delete or relocate the 
stated administrative control provisions from the TS to the UFSAR, 
plant procedures or the QA Program whereby, adequate control of 
information is maintained. Further, as stated above, the proposed 
changes do not alter the design, function, or operation of any plant 
components and therefore, no new accident scenarios are created.
    (3) The proposed changes do not involve a significant reduction 
in a margin of safety because they are administrative in nature. 
None of the proposed changes involve a physical modification to the 
plant, a new mode of operation or a change to the UFSAR transient 
analyses. No Limiting Condition for Operation, ACTION statement or 
Surveillance Requirement is affected. The proposed changes do not 
involve a significant reduction in a margin of safety. Additionally, 
the proposed change does not alter the scope of equipment currently 
required to be OPERABLE or subject to surveillance testing nor does 
the proposed change affect any instrument setpoints or equipment 
safety functions. 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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter, Acting.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit 2, Shippingport, Pennsylvania

    Date of amendment request: April 26, 1995.
    Description of amendment request: The proposed amendment would add 
a requirement to Technical Specification (TS) 4.5.2.a to periodically 
verify that the High Head Safety Injection (HHSI) pump minimum flow 
valve, 2CHS*MOV373, is maintained open during plant operation in Modes 
1, 2, and 3. Valve 2CHS*MOV373 must be maintained open to provide a 
minimum flowpath for the HHSI pumps and thereby minimize the likelihood 
of HHSI pump damage due to operating the pumps with insufficient flow. 
The proposed change would allow flexibility for local verification of 
valve position or flow indication if the control room indication is not 
available. The proposed amendment would also make several editorial 
changes to TS 3/4.5.2 for consistent format with other TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Maintaining 2CHS*MOV373 in a de-energized locked open position 
ensures charging/High Head Safety Injection pump (HHSI pump) minimum 
flow remains available for normal operation and design basis 
accidents. It has been determined that with 2CHS*MOV373 in the open 
position there is no significant increase in radiation levels and no 
change to the existing environmental qualification or personnel 
access routes. Sufficient injection flow to the core will be 
maintained during events requiring a Safety Injection (SI) 
actuation. Potential HHSI pump damage due to low flow will be 
prevented during periods of high Reactor Coolant System (RCS) 
pressure following a steam line break and SI. It has also been 
determined that the HHSI pumps will remain capable of performing 
their safety function with a continuous minimum flow. There is no 
impact on analysis assumptions or radiological consequences of an 
accident.
    There are no postulated events in the Updated Final Safety 
Analysis Report (UFSAR) which require that 2CHS*MOV373 be closed. 
Thus, the decision to de-energize and lock open the valve ensures 
adequate minimum flow for the HHSI pumps.
    The proposed addition of 2CHS*MOV373 to Technical Specification 
3.5.2 enhances the operator's ability to verify the valve position. 
The proposed surveillances and footnote will be used to monitor the 
valve position, the status of motor operator, and the valve position 
indicating lights. Therefore, the proposed change to the technical 
specification will ensure that the HHSI pump minimum flow is always 
available.
    Several editorial changes were also made to Technical 
Specification 3.5.2. These changes do not alter the intent of the 
technical specification and as such have no impact on previously 
evaluated accident scenarios.
    Therefore, the proposed change will 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 addition of 2CHS*MOV373 to the technical 
specifications does not involve changes to the physical plant. The 
proposed change adds surveillance requirements and a footnote which 
monitor the valve position, the lack of power to the 
[[Page 29875]] motor operator, and the valve position indicating 
lights. This assures that the minimum flow valve is open to maintain 
the HHSI pumps operable under all conditions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change provides additional action to ensure that 
2CHS*MOV373 remains open and minimum HHSI pump flow remains 
available. Safety limits and limiting safety system settings are not 
affected by this proposed change. There are no changes to the 
offsite dose consequences resulting from this request.
    Therefore, use of the proposed technical specification would 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: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg, 
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., 
Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 18, 1991, as supplemented by 
letters dated March 16, and December 2, 1994, and March 9, 1995.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) on control Room Air 
Conditioning System (CRACS) by separating the current composite 
requirements of TS 3.7.6 into four TSs covering three separate 
functions; control room emergency air filtration system (two mode 
sets), control room air temperature, and control room isolation and 
pressurization. The changes also increase the allowed outage time to 
identify and correct breaches to the control room envelope, adds 
requirements for make-up air flow rate to be used in conjunction with 
existing differential pressure requirements, and adds toxic gas 
specifications for Modes 5 and 6. The amendment is related to a 
revision to the Technical Specification Bases approved by the NRC in a 
letter dated August 9, 1988. The March 16, and December 2, 1994, and 
March 9, 1995 submittals provided additional information and included 
some additional restrictions in proposed changes by original 
application dated July 18, 1991. The original notice was published on 
September 4, 1991 (56 FR 43808). The additional submittals do not 
change the no significant hazard consideration determination previously 
made by the licensee.
    Basis for proposed no significant hazards consideration 
determination: The proposed change would create new Specifications as 
follows: 3/4.7.6.1 Emergency Air Filtration, Modes 1-4; 3/4.7.6.2 
Emergency Air Filtration, Modes 5 and 6; 3/4.7.6.3 Control Room Air 
Temperature; 3/4.7.6.4 Control Room Isolation and Pressurization. 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 limiting accidents against which the CRACS protects are:
     All Chapter 15 scenarios involving a release of 
radiation to the environment outside the containment,
     Toxic gas releases, and
     Smoke resulting from control room envelope fires.
    Limiting accidents against which the emergency air filtration 
system protects are all Chapter 15 scenarios involving release of 
radiation to the environment outside the containment.
    The probability and consequences of any of the limiting 
accidents listed above are unchanged by the specialization of the 
plant TSs. As pointed out in the description of the change, TSs 3/
4.7.6.1 and 3/4.7.6.2 have retained all requirements from the 
existing TS with the addition of one action statement based on the 
inoperability of both trains, and the exception of one action 
statement based on one inoperable train in Modes 5 or 6. This action 
statement is unnecessary since it is only applicable in a mode 
unlikely to experience the limiting design basis accidents against 
which this system protects. Therefore, the protection of the 
original specification is uncompromised for the function of 
emergency air filtration.
    There are two differences between the existing TS and the 
proposed TS 3/4.7.6.3 regarding control room air temperature. The 
first is the three hour outage allowed when both air conditioning 
units are inoperable [this was withdrawn by licensee's March 9, 
1995, letter].
    This corrects most types of failures. Although three hours are 
less restrictive than TS 3.0.3, it is not significantly less and 
therefore, does not seriously reduce the protection of the original 
specification. The other change is the reduction of the surveillance 
temperature from 110 deg.F to 80 deg.F. This is more restrictive 
than the existing version. All other requirements for air 
conditioning are retained in the proposed TS.
    Proposed TS 3/4.7.6.4, which concerns control room isolation and 
pressurization, allows more limited continued plant operation than 
the existing TS. When compared to existing actions required for 
continued operation with a known breach, the proposed specification 
recognizes the potential consequences that could arise from 
operation with an unidentified breach in the envelope and imposes 
more restrictive actions.
    Engineering analysis also shows that, for most of the time, 
toxic chemical concentrations in the control room envelope after a 
postulated release are largely the result of in-leakage from the RAB 
[reactor auxiliary building] after isolation. This has the effect of 
reducing the chemical concentration of gas leaking into the control 
room by at least an order of magnitude and ultimately results in a 
control room chemical concentration buildup rate slower than 
previously assumed. These characteristics make it likely that the 
operators would have sufficient time to don the breathing apparatus 
installed in the control room. It is also noteworthy that this 
emergency breathing apparatus is considered by Regulatory Guide 1.78 
to provide sufficient operator protection for those cases where 
chemical toxicity limits might be exceeded.
    The limited continued operation allowed by the proposed change, 
the design characteristics of the control room, and the installed 
breathing apparatus provides a reasonable level of protection for 
plant personnel. Some new restrictions are identified for the 
control room isolation and pressurization. These were not previously 
identified and therefore offer enhanced protection to the TS. All 
existing requirements specific to the isolation and pressurization 
function are retained in the proposed version. As such, the proposed 
specification offers more protection than the existing TS.
    Based on the above, these revisions to the TS will not adversely 
affect the reliability or performance of any installed equipment. 
There are no design changes associated with this proposed amendment, 
consequently, all aspects of the safety analysis will remain 
unchanged and there will be no physical change to the facility, and 
operation of Waterford 3 in accordance with these proposed changes 
will not involve a significant increase in the probability or 
consequence of any accident previously evaluated.
    To create a new or different kind of accident, these changes 
must introduce a new failure path. In this regard, these revisions 
are benign since they do not alter the system or its operation. With 
a few exceptions, all existing TS restrictions have been retained. 
The exceptions have been shown to have insignificant impact. 
Furthermore, several additional restrictions, not in the existing 
specification, have been added.
    Based on the above information, these changes do not introduce a 
new failure path and therefore, cannot create a new, unevaluated 
sequence of events. The current plant safety analyses are bounding 
and this revision will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
[[Page 29876]] 
    Safety margins related to the control room envelope air systems 
are established for control room temperature and the habitability of 
the control room following all credible accidents. This change does 
not modify the equipment installed in the plant or its operation. 
Therefore, existing margins of safety are retained, and the 
operation of Waterford 3 in accordance with this 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 12, 1995.
    Description of amendment request: The proposed change modifies 
surveillance requirements associated with containment leakage Technical 
Specification (TS) 3.6.1.2 by removing scheduler requirement for Type A 
tests to be performed specifically at 40 plus or minus 10 month 
intervals and, instead, reference Type A testing in accordance with 10 
CFR part 50, appendix J. The proposed change adopts the wording for 
primary containment integrated leak rate testing that is consistent 
with the requirements of the Combustion Engineering Improved Standard 
Technical Specifications (NUREG-1432). The proposed change also 
includes several administrative changes. The May 12, 1995, submittal 
superseded the November 16, 1993, submittal in its entirety. The 
November 16, 1993, submittal was noticed in Federal Register on January 
5, 1994
(59 FR 619).
    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 change will not affect the assumptions, design 
parameters, or results of any accident previously evaluated. The 
proposed change does not add or modify any existing equipment. The 
proposed Type A test schedule will continue to be consistent with 10 
CFR 50 Appendix J. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    The proposed change does not involve modifications to any 
existing equipment. The proposed change will not affect 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.
    The margin of safety for the containment barrier is, in part, 
preserved by compliance with 10 CFR 50 Appendix J. Although the 
proposed change will allow greater flexibility in meeting Appendix J 
requirements, the TS will continue to preserve compliance with 10 
CFR Appendix J. 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, Louisiana 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.

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

    Date of amendment request: May 1, 1995.
    Description of amendment request: The proposed amendment would 
provide a special test exception that would allow an extension of the 
standby diesel generator (SDG) allowed outage time for a cumulative 21 
days on each SDG once per fuel cycle, and it would also allow an 
extension of the essential cooling water (ECW) loop allowed outage time 
for a cumulative 7 days on each ECW loop once per fuel cycle. These 
extended allowed outage times will be used to perform required 
inspections and maintenance on the SDGs and the ECW system during power 
operation.
    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 Standby Diesel Generators are not accident initiators, 
therefore the increase in Allowed Outage Times for this system does 
not increase the probability of an accident previously evaluated. 
The three train design of the South Texas Project ensures that even 
during the seven days the Essential Cooling Water loop is inoperable 
there are still two complete trains available to mitigate the 
consequences of any accident. If the Essential Cooling Water loop is 
not operable during the 21 days the Standby Diesel Generator is 
inoperable, the Standby Diesel Generator's Engineered Safety 
Features bus and equipment in the train will be operable. This 
ensures that all three redundant safety trains of the South Texas 
Project design are operable. In addition the Emergency Transformer 
will be available to supply the Engineered Safety Features bus 
normally supplied by the inoperable Standby Diesel Generator. These 
actions will ensure that the changes do not involve a significant 
increase in the consequences of previously evaluated accidents.
    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 affect only the magnitude of the Standby 
Diesel Generator and Essential Cooling Water Allowed Outage Times 
once per fuel cycle as identified by the marked-up Technical 
Specification. As indicated above, the proposed change does not 
involve the alteration of any equipment nor does it allow modes of 
operation beyond those currently allowed. Therefore, implementation 
of these proposed changes does 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 result in no significant increase in core 
damage or large early release frequencies.
    Three sets of PSA results have been presented to the NRC for the 
South Texas Project. One submitted in 1989 from the initial Level 1 
PSA of internal and external events with a mean annual average CDF 
estimate of 1.7 x 10(-4), a second one submitted in 1992 to meet the 
IPE requirements from the Level 2 PSA/IPE with a CDF estimate of 4.4 
x 10(-5), and an update of the PSA that was reported in the August 
1993 Technical Specifications submittal with a variety of CDF 
estimates for different assumptions regarding the rolling 
maintenance profile and different combinations of modified Technical 
Specifications. The South Texas Project PSA was updated in March of 
1995 to include the NRC approved Risk-Based Technical 
Specifications, Plant Specific Data and incorporate the Emergency 
Transformer into the model. This update resulted in a CDF 
[[Page 29877]] estimate of 2.07 x 10(-5). When the requested changes 
are modeled along with the compensatory actions, the resulting CDF 
estimate is 2.30 x 10(-5). While this is slightly higher (approx. 
11%) than the updated results, it is still significantly lower 
(approx. 46%) than the previous Risk-Based Evaluation of Technical 
Specification submitted in 1993. Therefore, it is concluded that 
there is no significant reduction in the margin of safety.
    Based on the above evaluation, Houston Lighting & Power has 
concluded that these changes do not involve any significant hazards 
considerations.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: May 2, 1995.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications 3.4.2.2 and 3.7.1.1 (Table 3.7-2) by 
relaxing the lift setting tolerances of the pressurizer safety valves 
from plus or minus 1% to plus or minus 2% and the main steam safety 
valves from plus or minus 1% to plus or minus 3%, respectively. In 
addition, a footnote would be added to require that the pressurizer 
safety valves and main steam safety valves setpoint tolerances be 
restored to within plus or minus 1% whenever a lift setting is 
determined to be outside plus or minus 1% following valve testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because:
    The proposed changes increase the ``as-found'' setpoint 
tolerances for the Pressurizer Safety valves from plus or minus 1% 
to plus or minus 2% and the Main Steam Safety valves from plus or 
minus 1% to plus or minus 3%. The proposed changes do not involve 
any hardware modifications to plant structures, systems, or 
components. An evaluation has determined that the proposed changes 
do not significantly affect the structural integrity of either the 
reactor coolant system or the main steam system.
    The proposed setpoint tolerance of plus or minus 2% for the 
Pressurizer Safety valves and plus or minus 3% for the Main Steam 
Safety valve ``as-found'' condition was previously evaluated as part 
of the evaluation for the transition to VANTAGE 5H fuel. The 
evaluation was reviewed and approved by the NRC Staff as part of 
License Amendment Nos. 61 and 50 to Operating License NPF-76 and 
NPF-80. Since the VANTAGE 5H fuel evaluation incorporated these 
proposed changes, the calculated radiological release associated 
with that evaluation is unaffected. Similarly, this applies to the 
radiological dose associated with a steam generator tube rupture.
    Additionally, the proposed change [sic] are consistent with the 
guidance provided by Section III and XI of the ASME Code.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because:
    Since the lift setting of a Pressurizer Safety valve or Main 
Steam Safety valve will be restored to plus or minus 1% whenever it 
is determined to be outside plus or minus 1%, the ``as-left'' 
setpoint tolerances for the Pressurizer Safety valves and Main Steam 
Safety valves are unchanged. The ``as-left'' setpoint will continue 
to satisfy the current technical specification requirement on lift 
setting tolerance. As such, there is no change in plant operation or 
equipment performance. Since neither plant operation or equipment 
performance is affected by the proposed changes, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    3. The proposed change does not involve a significant reduction 
in a margin of safety because:
    Since the proposed changes are consistent with the guidance 
provided by Section III and XI of the ASME Code, and the proposed 
lift setting tolerance of plus or minus 2% for the Pressurizer 
Safety valves and plus or minus 3% for the Main Steam Safety valves 
has been incorporated into the design basis accident analyses, the 
proposed changes do not involve a significant reduction in the 
margin of safety.
    Based on the safety evaluation presented above for the proposed 
changes, Houston Lighting & Power has determined that the health and 
safety of the public will not be jeopardized. Therefore, the 
proposed changes do not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 
77488.
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036.
    NRC Project Director: William D. Beckner.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit No. 1, Berrien County, Michigan

    Date of amendment request: April 13, 1995.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications to allow use of laser-welded 
sleeves to repair defective steam generator tubes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Conformance of the proposed amendments to the standards for a 
determination of no significant hazard as defined in 10 CFR 50.92 
(three factor test) is shown in the following:
    (1) Operation of CNP [Cook Nuclear Plant] Unit 1 in accordance 
with the proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The TS [tubesheet] or TSP [tube support plate] intersection LWS 
[laser-welded sleeve] configuration has been designed and analyzed 
in accordance with the requirements of the ASME [American Society of 
Mechanical Engineers] Code and RG [Regulatory Guide] 1.121. Fatigue 
and stress analyses of the sleeved tube assemblies produced 
acceptable results. Mechanical testing has shown that the structural 
strength of the Alloy 690 sleeves under normal faulted and upset 
conditions is within acceptable limits. Leak testing has 
demonstrated that primary to secondary leakage is not expected 
during all plant conditions, including the case where the seal weld 
is not produced in the lower joint of the TS sleeve. Testing shows 
that non-welding TS sleeve lower joints remained leaktight at 
temperature and pressure conditions representative of normal and 
accident conditions. Since laser welding produces a hermetic seal 
between the tube and sleeve, no leak path can be realized under any 
condition. Therefore, installation of LWSs will not influence 
offsite dose [[Page 29878]] calculation for a postulated steam line 
break event.
    The proposed technical specification change to support the 
installation of Alloy 690 LWSs does not adversely impact any 
previously evaluated design basis accident or the results of 
accident analyses for the current technical specification minimum 
reactor coolant system flow rate. The results of the qualification 
testing, analyses, and plant operating experience demonstrate that 
the sleeve assembly is an acceptable means of maintaining tube 
integrity. These aforementioned analyses and tests demonstrate that 
installation of sleeves spanning degraded areas of the tube will 
restore the tube to a condition consistent with its original design 
basis. Plugging limit criteria are established using the guidance of 
RG 1.121. Furthermore per RG 1.83 recommendations, the sleeved tube 
can be monitored through periodic inspections with present eddy 
current techniques.
    Conformance of the sleeve design with the applicable sections of 
the ASME Code and results of the leakage and mechanical tests, 
support the conclusion that installation of laser-welded tube 
sleeves will not increase the probability or consequences of an 
accident previously evaluated. Depending upon the break location for 
a postulated steam generator tube rupture event, implementation of 
tube sleeving could act to reduce the radiological consequences to 
the public due to reduced flow rate through a sleeved tube compared 
to a non-sleeved tube based on the restriction afforded by the 
sleeve wall thickness.
    (2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of laser-welded sleeving will not introduce 
significant or adverse changes to the plant design basis. Stress and 
fatigue analysis of the repair has shown the ASME Code and RG 1.121 
allowable values are met. Implementation of laser-weld sleeving 
maintains overall tube bundle structural and leakage integrity 
during all plant conditions at a level consistent to that of the 
originally supplied tubing. Leak and mechanical testing of sleeves 
supports the conclusions of the calculations that the sleeve retains 
both structural and leakage integrity during all conditions. 
Sleeving of tubes does not provide a mechanism resulting in an 
accident outside of the area affected by the sleeves. Any 
hypothetical accident as a result of potential tube or sleeve 
degradation in the repaired portion of the tube is bounded by the 
existing tube rupture accident analysis. Since the sleeve design 
does not affect any other component or location of the tube outside 
of the immediate area repaired, in addition to the fact that the 
installation of sleeves and the impact on current plugging level 
analyses is accounted for, the possibility that laser-weld sleeving 
creates a new or different type of accident is not supported.
    The design of thermally treated Alloy 600 and 690 sleeved tube 
assemblies have performed well historically with regard to 
corrosion. There are no reported instances of Alloy 600 thermally 
treated or Alloy 690 sleeve degradation for the greater than 35,000 
sleeves that Westinghouse has installed in the U.S. Accelerated 
corrosion test results show the free span laser-weld joint (LWJ) 
(with post weld heat treatment) is capable of exhibiting a 
resistance to corrosion of greater that 10 times that of rolled tube 
transitions. Most LWS corrosion specimens did not experience 
degradation and were subsequently removed from the corrosion test 
media after a substantial testing period (supporting the 10x factor 
compared to roll transitions) was achieved. Several mill annealed 
Alloy 600 material heats were used for corrosion specimen 
preparation. All were documented by previous test to have been 
highly susceptible to PWSCC. The post weld heat treatment process 
applied to LWS free span joints is designed to achieve a minimum 
tube OD wall temperature of 1400 deg.F adjacent to the weld and 
within the laser weld heat affected zone. Since the target 
temperature of 1400 deg.F is achieved on the tube OD, a slightly 
higher temperature is achieved at the tube ID surface, where the 
weld cooling stresses are concentrated. Also, since the axial length 
of the laser weld and laser weld heat affected zone are relatively 
narrow compared to other sleeve welding processes, a narrower 
section of tube is required to be heat treated. Since the length of 
tube required to be heat treated is shorter in the LWS process than 
with other sleeving processes, lower residual stresses in the tube 
can be expected. Accelerated corrosion tests also show that non-heat 
treated laser-weld free span joints exhibit resistance to stress 
corrosion cracking equal to or greater than rolled tube transitions. 
An extensive data base exists on LWS joint performance in foreign 
plants in which the free span joints are not heat treated. Of the 
approximately 18,000 non-heat treated joints in service, none has 
exhibited a rapid corrosion potential. Corrosion testing of the TS 
sleeve lower joint LWJs exhibit a resistance to corrosion cracking 
of three to four times that of rolled tube transitions. These 
factors suggest postulated sleeve/tube assembly degradation would 
occur at a rate less than rolled transitions, and the potential for 
a sleeve/tube assembly with accelerated degradation rate 
characteristics more severe than rolled transitions, and the 
potential for a sleeve/tube assembly with accelerated degradation 
rate characteristics more severe than roll transitions is 
negligible.
    Approximately 800 LWSs are currently in operation in the U.S. 
Some of these have been in service since April 1992. The plants in 
which these sleeves are installed have not experienced any adverse 
operational issues (such as primary to secondary leakage) as has 
been detected at other plants with sleeves which have experienced 
rapid corrosion of the parent tube.
    (3) The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The laser-welded sleeving repair of degraded steam generator 
tubes as identified in WCAP-13088 Rev. 3 has been demonstrated to 
restore the integrity of the tube bundle under normal and postulated 
accident conditions. The safety factors used in the design of 
sleeves for the repair of degraded tubes are consistent with the 
safety factors the ASME Boiler and Pressure Vessel Code used in 
steam generator design. The plugging limit criteria for the sleeve 
has been established using the methodology of RG 1.121. The design 
of the sleeve joints have been verified by testing to preclude 
leakage during normal and postulated accident conditions. 
Implementation of laser-welded sleeving will reduce the potential 
for primary to secondary leakage during a postulated steam line 
break while maintaining available primary coolant flow area in the 
event of a LOCA. By removing from service degraded intersections 
through repair, the potential for tube leakage during a steam line 
break is reduced. These degraded intersections now are returned to a 
condition consistent with the design basis. While the installation 
of a sleeve causes a reduction in flow, the reduction is far below 
the reduction incurred by plugging. Therefore, far greater primary 
coolant flow area is maintained through sleeving. Use of RG 1.121 
criteria assures that the margin of safety with respect to 
structural integrity is the same for the sleeves as for the original 
steam generator tubes.
    The portions of the installed sleeve assembly which represent 
the reactor coolant pressure boundary can be monitored for the 
initiation and progression of sleeve/tube wall degradation, thus 
satisfying the requirements of RG 1.83. Portions of the tube bridged 
by the sleeve joints are effectively isolated from the pressure 
boundary, and the sleeve then forms the pressure boundary in these 
areas. The areas of the sleeved tube assembly which require 
inspection are defined in Attachment 4 [WCAP-13088, ``Westinghouse 
Series 44 and 51 Steam Generator Generic Sleeving Report, Laser 
Welded Sleeves,'' January 1994].
    In addition, since the installed sleeve represents a portion of 
the pressure boundary, a baseline inspection of these areas is 
required prior to operation with sleeves installed. As stated 
previously, weld fusion zone width is established using UT testing. 
The minimum acceptable weld width as determined by UT examination is 
approximately 50% wider than the minimum weld width which satisfies 
the stress conditions of the ASME Code.
    The generic evaluation uses the pressure stress equation of 
Section NB 3224.1 of the ASME Code which is used to establish the 
minimum required wall thickness for the sleeve design and 
subsequently used to determine the level of sleeve wall degradation 
(depth by eddy current determination) that would require the sleeve 
to be removed from service. Using the [Delta]PNorm. Op. value 
of 1530 psi from Attachment 4 [WCAP-13088, ``Westinghouse Series 44 
and 51 Steam Generator Generic Sleeving Report, Laser Welded 
Sleeves,'' January 1994] the limiting minimum required sleeve wall 
thickness is established. The sleeve wall plugging limit (using 
Attachment 4 [WCAP-13088, ``Westinghouse Series 44 and 51 Steam 
Generator Generic Sleeving Report, Laser Welded Sleeves,'' January 
1994]) of 25% is subsequently established, and includes an allowance 
of 10% for eddy current uncertainty and 10% for growth, although 
sleeve wall degradation has not been observed to date in 
Westinghouse [[Page 29879]] sleeves. The generic evaluation used the 
ASME Code minimum property values to establish the sleeve plugging 
limit. Certified material test reports indicate that the sleeve 
material properties are significantly higher than the ASME Code 
minimum values. The generic evaluation considered a primary to 
secondary pressure differential of 1530 psia, with a steam pressure 
of 720 psia, for normal operating conditions. CNP Units 1 can 
operate at full power with a reduced Thot value and RCS 
pressure of 2250 psi. Steam pressure can be maintained as low as 650 
psi (to keep Thot as low as possible), but cannot go lower than 
650 psi or the steam generator operating requirement of a primary to 
secondary [Delta]P of 1600 psi (max) will be exceeded. At this 
[Delta]PNorm. Op. value of 1600 psi, the sleeve minimum wall 
thickness requirement (and subsequently sleeve pressure boundary 
plugging limit) using ASME Code minimum material properties can be 
recalculated. For this condition (normal operating [Delta]P equal to 
1600 psi), the sleeve minimum wall plugging limit is defined to be 
23%. An allowance for eddy current uncertainty and continued 
degradation are included in this value. The minimum required wall 
thickness is determined by examining plant conditions at normal, 
upset, faulted, and test conditions. For Model 51 steam generators, 
the normal operating condition results in the limiting minimum wall 
thickness requirement.

    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: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting.

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

    Date of amendment requests: December 20, 1993, as supplemented July 
19, 1994, and February 28, 1995.
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications to change Train A and B emergency 
loads from 8 hour to composite 4 hour, delete a load on the Train B 
batteries load list, and revise the operational loads on the Train N 
batteries. The supplemental submittals, made in response to NRC staff 
concerns, would also add surveillance requirements for a battery with 
signs of degradation and modify performance testing requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which was published in the Federal Register on February 
2, 1994 (59 FR 4939). This analysis was not changed by the supplemental 
submittals.
    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, including the supplemental submittals, involve no 
significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting.

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

    Date of amendment requests: March 31, 1995.
    Description of amendment requests: The proposed amendments would 
revise the technical specifications to provide increased flexibility in 
the operation of the containment personnel airlocks during core 
alterations.
    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:

    Per 10 CFR 50.92, a proposed change does not involve a 
significant hazards consideration if the change does not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety.

Criterion 1

    The design basis fuel handling accident is the rupture of the 
highest rated fuel assembly. As discussed previously [in the 
application], the consequences of an accident inside containment 
(i.e., site boundary dose) with both airlock doors are bounded by 
the existing fuel handling accident currently presented in our UFSAR 
[Updated Final Safety Analysis Report].
    Since the containment airlock doors do not affect the failure 
mechanism of a fuel assembly during a fuel handling accident, we 
believe that this amendment request does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Additionally, no credit was taken for 
containment closure in the accident analysis. Therefore, based on 
these considerations, it is concluded that the proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Criterion 2

    As stated in response to criterion one, the position of the 
containment airlock doors in no way affects the mechanism by which a 
spent fuel assembly is damaged during a fuel handling accident. 
Thus, it is concluded that the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

Criterion 3

    The margin for safety as defined in 10 CFR 100 has not been 
reduced. As discussed previously, the existing fuel handling 
accident analysis for an event inside containment takes no credit 
for the isolation of containment. As a result, the position of the 
airlock doors has no impact on the analyzed site boundary doses 
resulting from such an accident. Based on these considerations, it 
is concluded that the 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, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: May 2, 1995.
    Description of amendment request: This license amendment request 
revises Surveillance Requirement (SR) 4.7.A.2.f.1 to allow a one-time 
schedular extension of the two year Type B Local Leak Rate Test (LLRT) 
interval required for the Drywell Head and Manport (penetrations DWH 
and X-4 respectively). This extension will allow [[Page 29880]] the 
Type B testing of penetrations DWH and X-4 to be deferred from the 
current due date of July 17, 1995, until Refueling Outage No. 16 (RE-
16), which is currently scheduled to commence in October 1995.
    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 enclosed Technical Specifications change is judged to 
involve no significant hazards based on the following:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    This license amendment request revises Surveillance Requirement 
(SR) 4.7.A.2.f.1 to allow the one-time schedular extension of the 
two year Type B Local Leak Rate Test (LLRT) interval required for 
the Drywell Head and Manport (Penetrations DWH and X-4 
respectively). This extension will allow Penetrations DWH and X-4 to 
be Type B tested during Refueling Outage No. 16 (RE-16), which is 
currently scheduled to commence October 1995. Currently, the two 
year maximum interval for these penetrations comes due July 17, 
1995. The District has concluded that a one-time extension of 
approximately six months beyond the two year limit will not result 
in a significant increase in the probability of these penetrations 
failing to perform their safety function. This conclusion is based 
on the previous LLRT surveillance history of Penetrations DWH and X-
4, which have not failed an LLRT in the last 19 years. The 
surveillance history demonstrates that these penetrations are not 
subject to leak related failures.
    Additionally, the seals associated with these penetrations will 
not have experienced significantly more radiation and heat exposure 
by the conclusion of the proposed extension than they would have 
during the current two year interval. Although some radiation and 
heat is present during plant shutdowns, the seal degradation 
resulting from these conditions is significantly slower. Because 
seal degradation is a function of heat and radiation, and is 
generally not a function of time, the District has concluded that 
the one-time extension will not result in a significant increase of 
seal degradation. Because seal failure for these penetrations is 
largely based on the rate of seal degradation, the probability of 
the failure of these penetrations is not significantly increased. 
Therefore, a significant increase in the probability or consequences 
of an accident is not created.
    This proposed change does not introduce any new modes of plant 
operation, make any physical changes, or alter any operational 
setpoints. The change does not degrade the performance of any safety 
system assumed to function in the accident analysis. Therefore, this 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility for a new or 
different kind of accident from any accident previously evaluated?
    This license amendment request involves the one-time schedular 
extension of the LLRT interval requirement for Penetrations DWH and 
X-4. SR 4.7.A.2.f.1 is being revised to extend the surveillance test 
interval for Penetrations DWH and X-4 to coincide with RE-16, 
currently scheduled to commence October 1995. A one-time extension 
of the subject surveillance interval does not involve the creation, 
deletion, or modification of the function of any structure, system, 
or component, nor does this change introduce or change any mode of 
plant operation. This proposed change does not create the 
possibility for a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change create a significant reduction in 
the margin of safety?
    This license amendment request involves the one-time extension 
of the two year maximum surveillance test interval for Penetrations 
DWH and X-4 from the current due date of July 17, 1995, to instead 
coincide with RE-16, which is scheduled to commence October 1995. By 
the time these tests are performed, the penetration seals will not 
have experienced significantly more radiation and heat than they 
would have during the previous test intervals. Therefore, the 
penetration seals will not have experienced significant degradation 
as a result of the extended interval. Furthermore, Penetrations DWH 
and X-4 have not failed an LLRT in the last 19 years. The 
surveillance history demonstrates that these penetrations are not 
subject to leak related failure. This proposed change does not 
involve any change to plant design, equipment instrument setpoints, 
or operation. Therefore, this proposed change does not create 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: Auburn Public Library, 118 
15th Street, Auburn, NE 68305.
    Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: March 29, 1995.
    Description of amendment request: The request will revise Technical 
Specification Section 3.10.5 to allow more than one control bank to be 
fully withdrawn from the core simultaneously for rod drop time response 
testing. Specifically, the change will delete, (1) the limiting 
condition for operation (LCO) 3.10.5.a and (2) a reference to the full 
length shutdown rods from LCO 3.10.5. The change will also add a 
statement that ``The SHUTDOWN MARGIN requirement of Section 3.1.1.1.2 
shall be met without credit for withdrawn control rods.'' Other 
editorial changes are to be made for consistency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    * * * The proposed changes do not involve an SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes cannot initiate an event since the plant 
will be maintained shutdown at all times. Thus, there is no increase 
in the probability of occurrence of an accident previously 
evaluated.
    The proposed changes do not degrade the performance of any 
safety system nor do they alter any assumptions made in the accident 
analyses. Currently, the technical specifications allow the rod 
position indication system to be disabled for each control bank 
while performing this test. In addition, this system is not a safety 
system credited in the accident analyses. Therefore, allowing more 
than one bank to have its indication removed during the test does 
not degrade any safety system. Since the shutdown margin will be 
maintained without crediting these rods, there is no change to the 
assumptions made in the accident analyses. Thus, there is no 
increase in the consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed changes do not position the control rods into any 
new configurations or sequence not previously analyzed. Ejected rod 
worths are evaluated for ARI-1 (all rods in with the most reactive 
rod out) and, therefore, bound the test configuration. In addition, 
the reactivity state of the system is maintained shut down by the 
margin required in Technical Specification 3.1.1.1.2 without 
crediting the control rods. Therefore, there is no possibility of a 
new or different type of accident than previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The proposed changes do not impact any of the physical 
protective boundaries, safety systems, or operating conditions. The 
plant [[Page 29881]] will be maintained shut down without crediting 
the control rods. The accident analyses is not impacted and, 
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: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L.M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: April 28, 1995.
    Description of amendment request: The request will revise the 
diesel generator (DG) fuel oil testing that is performed on new fuel 
prior to the addition of the new fuel to the storage tank.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    * * * The proposed changes do not involve an SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes: correct a typographical error by providing 
the appropriate range for the Saybolt viscosity; replace the 
qualitative clear and bright test with a quantitative water and 
sediment test for new fuel prior to adding it to the storage tank; 
and clarify that a calculated cetane index may be performed in lieu 
of obtaining the cetane number for the fuel. The water and sediment 
test provides a quantitative method for evaluating water and 
sediment, and will require a more restrictive limit of 0.05 percent 
by volume of water and sediment than the 0.10 percent recommended by 
the manufacturer. The cetane index has been shown to be 
representative of the cetane number for the fuel. The DG capability 
to start and operate is enhanced by the proposed changes. Therefore, 
the changes have no negative effect on the consequences of the 
previously evaluated accidents.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter or affect the design, 
function, failure mode, or operation of the plant. The proposed 
changes have no adverse effect on the quality of the fuel oil that 
is utilized by the DG. The proposed changes are administrative in 
nature and do not involve any physical alteration to any plant 
system or change the method by which any safety-related system 
performs its function. For these reasons, there is no possibility of 
an accident of a different type than previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will assure that the DG fuel oil meets DG 
manufacturer's quality requirements by the performance of the 
recommended testing of the DG fuel oil. The proposed changes will 
not impact the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: April 28, 1995.
    Description of amendment request: The proposed revision to the 
Action Statement of Limiting Condition for Operation (LCO) 3.7.5 would 
permit Millstone Unit No. 3 to remain in Modes 1 through 4 with the 
average water temperature of the ultimate heat sink (UHS) greater than 
75 deg.F (but lower than 77 deg.F) for 12 hours. An additional action 
would be added which would require the plant to be placed in at least 
HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 
hours upon identifying that the UHS temperature is greater than 
77 deg.F.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    * * * The proposed changes do not involve an SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed addition of a 12-hour period to monitor the UHS 
temperature to the Technical Specification LCO Action Statement does 
not involve an increase in the probability of an accident previously 
evaluated. The probability of an accident previously evaluated is 
not increased by a short-term increase in the UHS temperature. The 
probability of FSAR Chapter 15 Condition IV accidents occurring in 
conjunction with the short duration increase in service water inlet 
temperature above 75 deg.F is low enough such that they are not risk 
significant. Further, an evaluation has been performed that safe 
shutdown will be achieved and maintained for a loss of offsite power 
event and a steam generator tube rupture event with the additional 
consideration of a single failure with service water inlet 
temperatures as high as 77 deg.F. There has been no significant 
increase in the consequences of these events previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed technical specification change does not create the 
possibility of a new or different kind of accident previously 
evaluated. The addition of a 12-hour time period to monitor the UHS 
temperature increases the amount of time that is allowed for the 
plant to be in HOT STANDBY from 6 to 18 hours should the UHS 
temperature increase above 75 deg.F. This extension of the time 
allowed for the plant to be in HOT STANDBY does not change the plant 
configuration. As such, the change does not create the possibility 
of a new or different kind of accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed technical specification change does not involve a 
significant reduction in the margin of safety. The addition of a 12-
hour time period to monitor the UHS temperature increases the time 
required for the plant to be in HOT STANDBY from 6 to 18 hours 
should the UHS temperature exceed 75 deg.F. An evaluation has been 
performed to demonstrate that the risk significance associated with 
the increased action time is very low. In addition, safe shutdown 
capability has been demonstrated for service water inlet 
temperatures as high as 77 deg.F.

    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, 
[[Page 29882]] Three Rivers Community-Technical College, 574 New London 
Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L.M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: May 1, 1995.
    Description of amendment request: Technical Specifications that 
specify an 18-month surveillance will be changed to state that these 
surveillances are to be performed at least once each refueling (i.e., 
24 months).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    * * * The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed change to Surveillance Requirement 4.6.3.2 of the 
Millstone Unit No. 3 Technical Specifications extends the frequency 
for verifying that each containment isolation valve actuates to its 
required position in response to Phase A and Phase B isolation test 
signals, and for verifying that each containment purge supply and 
exhaust isolation valve actuates to its required position in 
response to a containment high radiation test signal. The proposal 
would extend the frequency from at least once per 18 months to at 
least once per refueling interval (24 months).
    The proposed change to Surveillance Requirement 4.6.3.2 does not 
alter the intent or method by which the surveillances are conducted, 
does not involve any physical changes to the plant, does not alter 
the way any structure, system, or component functions, and does not 
modify the manner in which the plant is operated. As such, the 
proposed change to the frequency of Surveillance Requirement 4.6.3.2 
will not degrade the ability of the containment isolation valves to 
perform their safety function. Also, the containment isolation valve 
arrangements are not vulnerable to single failures, because they 
provide at least two barriers between the atmosphere outside the 
containment and the atmosphere within the containment, the reactor 
coolant system, or systems that would become connected to the 
containment atmosphere or the reactor coolant system as a result of, 
or subsequent to, a DBA.
    Additional assurance of containment isolation valve operability 
is provided by Surveillance Requirements 4.6.3.1 and 4.6.3.3. 
Surveillance Requirement 4.6.3.1 requires that a containment 
isolation valve will be restored to an operable status following the 
performance of work on the containment isolation valve or its 
ancillaries. Surveillance Requirement 4.6.3.3 requires the 
confirmation of the mechanical operability of the containment 
isolation valves by the inservice inspection program. The proposed 
change does not modify these requirements.
    Additionally, Surveillance Requirements 4.3.2.1 and 4.3.3.1 
assure the operability of the automatic isolation logic (Phase A and 
Phase B isolation signals and containment high radiation signal) for 
the containment isolation valves by performing tests on a monthly 
basis. This proposed change does not modify these Surveillance 
Requirements.
    Equipment performance over the last four operating cycles was 
evaluated to determine the impact of extending the frequency of 
Surveillance Requirement 4.6.3.2. This evaluation included a review 
of surveillance results, preventive maintenance records, and the 
frequency and type of corrective maintenance. It has been concluded 
that the containment isolation valves are highly reliable, and that 
there is no indication that the proposed extension could cause 
deterioration in valve condition or performance.
    Based on the above, the proposed change to Surveillance 
Requirement 4.6.3.2 of the Millstone Unit No. 3 Technical 
Specifications does not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change to Surveillance Requirement 4.6.3.2 of the 
Millstone Unit No. 3 Technical Specifications extends the frequency 
for verifying that each containment isolation valve actuates to its 
required position in response to Phase A and Phase B isolation test 
signals, and for verifying that each containment purge supply and 
exhaust isolation valve actuates to its required position in 
response to a containment high radiation test signal. The proposal 
would extend the frequency from at least once per 18 months to at 
least once per refueling interval (24 months).
    The proposed change does not alter the intent or method by which 
the surveillances are conducted, does not involve any physical 
changes to the plant, does not alter the way any structure, system, 
or component functions, and does not modify the manner in which the 
plant is operated. As such, the proposed change in the frequency of 
Surveillance Requirement 4.6.3.2 will not degrade the ability of the 
containment isolation valves to perform their safety function. Also, 
the containment isolation valve arrangements are not vulnerable to 
single failures, because they provide at least two barriers between 
the atmosphere outside the containment and the atmosphere within the 
containment, the reactor coolant system, or systems that would 
become connected to the containment atmosphere or the reactor 
coolant system as a result of, or subsequent to, a DBA.
    Based on the above, the proposed change to Surveillance 
Requirement 4.6.3.2 of the Millstone Unit No. 3 Technical 
Specifications will 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.
    The proposed change to Surveillance Requirement 4.6.3.2 of the 
Millstone Unit No. 3 Technical Specifications extends the frequency 
for verifying that each containment isolation valve actuates to its 
required position in response to Phase A and Phase B isolation test 
signals, and for verifying that each containment purge supply and 
exhaust isolation valve actuates to its required position in 
response to a containment high radiation test signal. The proposal 
would extend the frequency from at least per 18 months to at least 
once per refueling interval (24 months).
    The proposed change does not alter the intent or method by which 
the surveillances are conducted, does not involve any physical 
changes to the plant, does not alter the way any structure, system, 
or component functions, and does not modify the manner in which the 
plant is operated. As such, the proposed change in the frequency of 
Surveillance Requirement 4.6.3.2 will not degrade the ability of the 
containment isolation valves to perform their safety function. Also, 
the containment isolation valve arrangements are not vulnerable to 
single failures, because they provide at least two barriers between 
the atmosphere outside the containment and the atmosphere within the 
containment, the reactor coolant system, or systems that would 
become connected to the containment atmosphere or the reactor 
coolant system as a result of, or subsequent to, a DBA.
    Additional assurance of the operability of the containment 
isolation valves is provided by Surveillance Requirements 4.6.3.1 
and 4.6.3.2. Also, assurance of the operability of the automatic 
actuation logic of the containment isolation valves is provided by 
Surveillance Requirements 4.3.2.1 and 4.3.3.1.
    Equipment performance over the last four operating cycles was 
evaluated to determine the impact of extending the frequency of 
Surveillance Requirement 4.6.3.2. This evaluation included a review 
of surveillance results, preventive maintenance records, and the 
frequency and type of corrective maintenance. It has been concluded 
that the containment isolation valves are highly reliable, and that 
there is no indication that the proposed extension could cause 
deterioration in valve condition or performance.
    Based on the above, the proposed change to Surveillance 
Requirement 4.6.3.2 of the Millstone Unit No. 3 Technical 
Specifications 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 
[[Page 29883]] amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: May 8, 1995.
    Description of amendment request: The proposed amendment would 
change Technical Specifications 2.3, 3.1, 3.2, 3.3 and 3.6. These 
changes are in accordance with the guidance of Generic Letter 93-05, 
``Line Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation,'' dated 
September 27, 1993.
    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.

GL 93-05, Item 4.2, Control Rod Movement Test; Specification 3.2, 
Table 3-5, Item 2

    Omaha Public Power District (OPPD) proposes to extend the 
control element assembly (CEA) partial movement surveillance test of 
Specification 3.2, Table 3-5, Item 2 from a biweekly to a quarterly 
frequency. This change is based on operating experience and the 
recommendation of Generic Letter (GL) 93-05, Item 4.2.1. A review of 
previous surveillance tests and interviews with personnel familiar 
with the test did not identify any prior surveillance test failures. 
Industry experience has shown that this test can cause reactor 
trips, dropped rods and unnecessary challenges to safety systems as 
stated in NUREG-1366, ``Improvements to Technical Specification 
Requirements,'' dated December 1992. Therefore, extending the 
frequency of conducting this surveillance test may be beneficial to 
plant operations and does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

GL 93-05, Item 5.14, Radiation Monitors; Specification 3.1, Table 
3-3, Items 3b, 4 and 5b

    OPPD proposes to replace descriptive wording in Specification 
3.1, Table 3-3, Items 3a/b and 5a/b with defined terms. OPPD also 
proposes to extend surveillance of the area, post-accident and 
primary to secondary leak-rate radiation monitors (Specification 
3.1, Table 3-3, Items 3b and 5b) from a monthly to a quarterly 
frequency as recommended by GL 93-05, Item 5.14. Most of these 
monitors are new (i.e., installed within the last two cycles) or 
contain many new components. The value of monthly testing is greatly 
reduced as the new monitors include self checking circuitry that 
will indicate monitor failure, loss of power, or loss of background. 
Although post accident radiation monitors RM-091 A/B are not new, 
Station operating experience has shown that they are reliable. In 
cases where new components interface with older components, the 
older components have a history of reliable operation.
    Readings and internal test signals are used to verify instrument 
operation on a daily basis. In addition, the proposed frequency 
(quarterly) is the same frequency currently specified for the 
containment radiation high signal (CRHS) monitors (Specification 
3.1, Table 3-2, Item 6b), which generate an engineered safeguards 
signal. Replacing descriptive words with defined terms ensures 
consistency and that the surveillance test accomplishes its purpose.
    A quarterly surveillance conserves resources, increases the 
availability of the area, post-accident and primary to secondary 
leak-rate detection radiation monitors and is consistent with CRHS 
monitor testing. These proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    OPPD proposes to delete Specification 3.1, Table 3-3, Item 4 on 
surveillance testing of the emergency plan radiation instruments. 
These are portable instruments stored in specified locations for use 
by emergency response personnel in the event of an accident. The 
instruments may be used to survey onsite/offsite areas for 
radioactivity or to facilitate the decontamination of personnel 
following an accident. No limiting condition for operation (LCO) 
action statement is associated with these instruments. As a result, 
there is no basis for the TS to contain a surveillance requirement 
for them. In addition, retaining this surveillance in the TS is 
unnecessary since it does not meet criteria 1 through 4 of the Final 
Policy Statement on Technical Specifications Improvements for 
Nuclear Power Reactors, dated July 22, 1993. Therefore, since these 
instruments are not utilized until after an accident has occurred, 
and do not assist in accident mitigation, deleting this surveillance 
requirement does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

GL 93-05, Item 6.1, Reactor Coolant System Isolation Valves; 
Specification 3.3(2)a

    The reactor coolant system (RCS) pressure isolation valves have 
proven to be very reliable. Therefore, OPPD proposes to extend the 
time that the plant can be in cold shutdown before the test is 
required (Specification 3.3(2)a) from 72 hours to 7 days, following 
the recommendation of GL 93-05, Item 6.1. A review of previous 
surveillance tests and interviews with personnel familiar with the 
test did not identify any prior surveillance test failures. This 
proposed change will reduce radiation exposure and does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
GL 93-05, Item 7.4, Accumulator Water Level and Pressure Channel 
Surveillance Requirements; Specification 2.3(2)g, Specification 
3.1, Table 3-2, Item 14a

    OPPD proposes to revise Specification 2.3(2)g following the 
recommendation of GL 93-05, Item 7.4. This revision will clarify 
that the safety injection tank (SIT) level and/or pressure 
instrumentation may be inoperable, which does not alter the intent 
of the Specification, but is more accurate in defining when the 
Specification applies. This revision also extends the time limit for 
inoperability of SIT instrumentation from 1 hour to 72 hours, which 
is justified based upon a review of historical data. As stated in 
NUREG-1366: ``While technically inoperable, the accumulator [SIT] 
would be available to fulfill its safety function during this time, 
and thus, this change would have a negligible increase on risk.''
    Currently, Specification 2.3(2)g allows only one hour for SIT 
level and pressure instrumentation to be inoperable, which is 
insufficient time to initiate repairs. A review of historical data 
determined that SIT water level stays relatively constant while 
pressure decreases slightly over time. It is unlikely that SIT 
pressure would decrease below the Specification 2.3(1)c limit of 240 
psig during the proposed 72-hour LCO, since SIT pressure is normally 
maintained around 255 psig (Updated Safety Analysis Report (USAR), 
Section 6.2.3.5).
    OPPD's proposal to revise Specification 3.1, Table 3-2, Item 14a 
to require shiftly verification that SIT level and pressure are 
within limits and remove reference to verifying ``indications are 
between independent high and low alarms for level and pressure,'' is 
consistent with the guidance of GL 93-05, Item 7.4. As stated in GL 
93-05, Item 7.4, the operability of SIT instrumentation is not 
directly related to the capability of a SIT to perform its safety 
function. OPPD proposes to suspend this surveillance on the affected 
SIT while the instrumentation is being repaired, since as stated 
above, SIT level and pressure are expected to stay within the limits 
of Specification 2.3(1)c during the proposed 72 hour LCO. Therefore, 
these proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

GL 93-05, Item 8.1, Containment Spray System; Specification 3.6(2)b

    OPPD proposes to extend the surveillance frequency for verifying 
that the containment spray nozzles are open (Specification 3.6(2)b) 
from five to ten years following the recommendation of GL 93-05, 
Item 8.1. Minor revisions to statements in the basis of 
Specification 3.6 that refer to conducting this test at five year 
intervals are proposed also. OPPD has not experienced problems with 
[[Page 29884]] obstructions in the containment spray nozzles as 
determined by a review of previous surveillance tests and personnel 
interviews. Of the three instances reported in NUREG-1366 concerning 
obstructions of containment spray nozzles, all were problems related 
to construction errors. Any construction errors in the FCS 
containment spray system would have been found by previous 
surveillance tests.
    The problem that occurred at San Onofre Unit 1 (clogging of 
several containment spray nozzles following the application of a 
coating material to the carbon steel piping) is not a concern at FCS 
since the FCS containment spray system piping and valves are 
constructed of stainless steel (USAR Table 6.3-2). Thus, extending 
the surveillance frequency of Specification 3.6(2)b from five to ten 
years does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

GL 93-05, Item 4.2, Control Rod Movement Test; Specification 3.2, 
Table 3-5, Item 2

    OPPD's proposal to extend the CEA partial movement surveillance 
test (Specification 3.2, Table 3-5, Item 2) to a quarterly frequency 
is based on operating experience and the recommendation of GL 93-05, 
Item 4.2.1. The proposed change only lengthens the time between 
surveillance tests and will not result in any physical alterations 
to the plant configuration, changes to setpoint values, or changes 
to the application of setpoints or limits. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

GL 93-05, Item 5.14, Radiation Monitors; Specification 3.1, Table 
3-3, Items 3b, 4 and 5b

    OPPD proposes to replace unnecessary wording in Specification 
3.1, Table 3-3, Items 3a/b and 5a/b with defined terms and to extend 
the surveillance frequency of Items 3b and 5b from monthly to 
quarterly based on the recommendation of GL 93-05, Item 5.14. Most 
of the area, post accident and primary to secondary leak-rate 
detection radiation monitors are new or contain new components. The 
new monitors include self checking circuitry that provides failure 
notification. Although post accident radiation monitors RM-091 A/B 
are not new, they have an excellent operating history. The proposed 
changes introduce consistent use of terminology and lengthen the 
time between surveillance tests and will not result in any physical 
alterations to the plant configuration, changes to setpoint values, 
or changes to the application of setpoints or limits. Therefore, 
these proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    OPPD's proposal to delete Specification 3.1, Table 3-3, Item 4 
on surveillance testing of the emergency plan radiation instruments 
will not result in any physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
application of setpoints or limits. Since these instruments are not 
utilized until after an accident has occurred, and do not assist in 
accident mitigation, this change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
GL 93-05, Item 6.1, Reactor Coolant System Isolation Valves; 
Specification 3.3(2)a

    The RCS pressure isolation valves have proven to be very 
reliable. As a result, OPPD proposes to extend the time that the 
plant can be in cold shutdown before the test is required 
(Specification 3.3(2)a) from 72 hours to 7 days following the 
recommendation of GL 93-05, Item 6.1. The proposed change will 
reduce radiation exposure and does not result in any physical 
alterations to the plant configuration, changes to setpoint values, 
or changes to the application of setpoints or limits. Therefore, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

GL 93-05, Item 7.4, Accumulator Water Level and Pressure Channel 
Surveillance Requirements; Specification 2.3(2)g, Specification 
3.1, Table 3-2, Item 14a

    OPPD's proposal to revise Specification 2.3(2)g following the 
guidance of GL 93-05, Item 7.4 more accurately states when the 
specification should apply and extends the time limit for 
inoperability of SIT instrumentation from 1 hour to 72 hours based 
upon a review of historical data. The proposed change will not 
result in any physical alterations to the plant configuration, 
changes to setpoint values, or changes to the application of 
setpoints or limits. As stated in NUREG-1366: ``While technically 
inoperable, the accumulator [SIT] would be available to fulfill its 
safety function during this time, and thus, this change would have a 
negligible increase on risk.''
    OPPD's proposal to revise Specification 3.1, Table 3-2, Item 14a 
to require shiftly verification that SIT level and pressure are 
within limits and remove reference to verifying ``indications are 
between independent high and low alarms for level and pressure,'' is 
consistent with the guidance of GL 93-05, Item 7.4. As stated in GL 
93-05, Item 7.4, the operability of SIT instrumentation is not 
directly related to the capability of a SIT to perform its safety 
function. OPPD proposes to suspend this surveillance on the affected 
SIT while the instrumentation is being repaired, since SIT level and 
pressure are expected to stay within the limits of Specification 
2.3(1)c during the proposed 72 hour LCO. Therefore, since these 
proposed changes do not result in any physical alterations to the 
plant configuration, changes to setpoint values, or changes to the 
application of setpoints or limits, they do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

GL 93-05, Item 8.1, Containment Spray System; Specification 3.6(2)b

    OPPD's proposal to extend the surveillance frequency for 
verifying that the containment spray nozzles are open (Specification 
3.6(2)b) from five to ten years as recommended by GL 93-05, Item 8.1 
is justified by operating experience. OPPD has not experienced 
problems with obstructions in the containment spray nozzles as 
determined by a review of previous surveillance tests and personnel 
interviews. The problem that occurred at San Onofre Unit 1 (clogging 
of several containment spray nozzles following the application of a 
coating material to the carbon steel piping) is not a concern at FCS 
since the FCS containment spray system piping and valves are 
constructed of stainless steel (USAR Table 6.3-2).
    The proposed change only extends the time between surveillance 
tests and revises associated basis statements to support the 
extension. The proposed change will not result in any physical 
alterations to the plant configuration, changes to setpoint values, 
or changes to the application of setpoints or limits. Therefore, 
OPPD's proposal to extend the surveillance frequency of 
Specification 3.6(2)b from five to ten years does 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.

GL 93-05, Item 4.2, Control Rod Movement Test; Specification 3.2, 
Table 3-5, Item 2

    OPPD's proposal to extend the CEA partial movement surveillance 
test of Specification 3.2, Table 3-5, Item 2 to a quarterly 
frequency is based on operating experience and the recommendation of 
GL 93-05, Item 4.2.1. A review of previous surveillance tests and 
interviews with personnel familiar with the test did not identify 
any prior surveillance test failures. Industry experience has shown 
that this test can occasionally cause reactor trips, dropped rods 
and unnecessary challenges to safety systems as stated in NUREG-
1366. Therefore, extending the frequency of conducting this 
surveillance test may be beneficial to plant operations and does not 
involve a significant reduction in a margin of safety.

GL 93-05, Item 5.14, Radiation Monitors; Specification 3.1, Table 
3-3, Items 3b, 4 and 5b

    OPPD proposes to replace descriptive wording in Specification 
3.1, Table 3-3, Items 3a/b and 5a/b with defined terms and to extend 
the surveillance frequency of Items 3b and 5b from monthly to 
quarterly based on the recommendation of GL 93-05, Item 5.14. Most 
of the area, post accident and primary to secondary leak-rate 
detection radiation monitors are new or contain new components. Post 
accident radiation monitors RM-091 A/B are not new but have a 
history of reliable operation. The value of monthly testing is 
greatly reduced since the new monitors include self checking 
circuitry that provides failure notification. The proposed changes 
introduce consistent use of terminology and lengthen the time 
between surveillance tests and therefore do not involve a 
significant reduction in a margin of safety.
    OPPD's proposal to delete Specification 3.1, Table 3-3, Item 4 
is justified because the [[Page 29885]] emergency plan radiation 
instruments are portable instruments that are not utilized until 
after an accident has occurred. The instruments are checked for 
proper operation before use and since these instruments do not 
assist in accident mitigation, the deletion of this surveillance 
requirement does not involve a significant reduction in a margin of 
safety.

GL 93-05, Item 6.1, Reactor Coolant System Isolation Valves; 
Specification 3.3(2)a

    The RCS pressure isolation valves have proven to be very 
reliable. Therefore, consistent with the guidance of GL 93-05, Item 
6.1, OPPD proposes to revise Specification 3.3(2)a and extend the 
time that the plant is allowed to be in cold shutdown before this 
surveillance test is required from 72 hours to 7 days. This change 
will reduce radiation exposure and does not involve a significant 
reduction in a margin of safety.

GL 93-05, Item 7.4, Accumulator Water Level and Pressure Channel 
Surveillance Requirements; Specification 2.3(2)g, Specification 
3.1, Table 3-2, Item 14a

    OPPD's proposal to revise Specification 2.3(2)g following the 
guidance of GL 93-05, Item 7.4 more accurately states when the 
specification applies and extends the time limit for inoperability 
of SIT instrumentation from 1 to 72 hours based upon historical 
data. As stated in NUREG-1366: ``While technically inoperable, the 
accumulator [SIT] would be available to fulfill its safety function 
during this time, and thus, this change would have a negligible 
increase on risk.''
    OPPD's proposal to revise Specification 3.1, Table 3-2, Item 14a 
to require shiftly verification that SIT level and pressure are 
within limits and remove reference to verifying ``indications are 
between independent high and low alarms for level and pressure,'' is 
consistent with the guidance of GL 93-05, Item 7.4. As stated in GL 
93-05, Item 7.4, the operability of SIT instrumentation is not 
directly related to the capability of a SIT to perform its safety 
function. OPPD proposes to suspend this surveillance on the affected 
SIT while the instrumentation is being repaired, since SIT level and 
pressure are expected to stay within the limits of Specification 
2.3(1)c during the proposed 72 hour LCO. Therefore, these proposed 
changes do not involve a significant reduction in a margin of 
safety.

GL 93-05, Item 8.1, Containment Spray System; Specification 3.6(2)b

    OPPD's proposal to extend the surveillance frequency for 
verifying that the containment spray nozzles are open (Specification 
3.6(2)b) from five to ten years as recommended by GL 93-05, Item 8.1 
is justified by operating experience. OPPD has not experienced 
problems with obstructions in the containment spray nozzles as 
determined by a review of previous surveillance tests and personnel 
interviews.
    The problem that occurred at San Onofre Unit 1 is not a concern 
at FCS since the FCS containment spray system piping and valves are 
constructed of stainless steel (USAR Table 6.3-2). Therefore, OPPD's 
proposal to extend the surveillance frequency of Specification 
3.6(2)b from five to ten years and revise associated basis 
statements 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L 
Street, Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of amendment request: April 10, 1995.
    Description of amendment request: The proposed amendment would 
revise License No. DPR-7, to permit the provisions of 10 CFR 50.59 to 
be applied with respect to changes to the facility or procedures 
described in the Decommissioning Plan or changes to the Decommissioning 
Plan, and the conduct of tests or experiments not described in the 
Decommissioning Plan.
    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 probability or consequences of an accident previously 
evaluated will not be effected by the ability to perform safety 
analyses. As outlined in 10 CFR 50.59, the impact of performing 
special tests, experiments, and modifications would be evaluated to 
verify there would be no impact on previously evaluated accidents or 
increase the probability or consequences of an accident occurring.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because there is no physical alteration to any plant system, nor is 
there a change in the method in which any quality-related activities 
are performed or any direct change in equipment or system function 
or operation. The proposed change is administrative in nature.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change to the HBPP License does not affect the 
margin of safety of any accident analysis since it does not affect 
the parameters for any accident analysis, and has no effect on the 
current operating methodologies or actions that govern plant 
performance.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the analysis of the licensee 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: Humboldt County Library, 636 F 
Street, Eureka, California 95501.
    Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas 
& Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: Seymour H. Weiss.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Dockets 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: January 17, 1995 as 
supplemented by letter dated March 30, 1995.
    Description of amendment request: The proposed change revises the 
Peach Bottom Atomic Power Station, Units 2 and 3 technical 
specifications to reflect the replacement of the source range monitor 
(SRM) and intermediate range monitor (IRM) systems with a new system 
referred to as the wide range neutron monitoring system (WRNMS).
    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 use of the WRNMS as discussed herein will not increase the 
probability or consequences of an accident previously evaluated. 
[[Page 29886]] 
    The probability (frequency of occurrence) of design basis 
accidents (DBAs) occurring is not affected by the WRNMS. The only 
plant safety analysis affected by WRNMS is the Rod Withdrawal Error 
(RWE) at low power, and a reanalysis assuming use of WRNMS shows 
that the criteria of 170 cal/gm for fuel enthalpy increase under RWE 
is satisfied; thus, RWE is not a limiting event. Scram setpoints 
(equipment settings that initiate automatic plant shutdowns) will be 
established such that there is no increase in scram frequency due to 
the WRNMS. No new challenges to safety-related equipment will result 
from WRNMS.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    As summarized below, this change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The components of the WRNMS will be supplied to equivalent or 
better design and qualification criteria than is currently required 
for the plant. Equipment that could be affected by WRNMS has been 
evaluated. No new operating mode, safety-related equipment lineup, 
accident scenario, system interaction, or equipment failure mode was 
identified. Therefore, the WRNMS will not adversely affect plant 
equipment.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    All the SRM/IRM functions required in the Technical 
Specifications are replaced with equivalent (more reliable) WRNMS 
functions. The accuracy and response times of the WRNMS are superior 
to those of the SRM/IRM subsystems. Implementation of the WRNMS does 
not affect any fuel or safety limit. The applicable Bases of the 
Technical Specifications have been rewritten, and the new Bases 
maintain the equivalent margin of safety as was provided by the SRM/
IRM Bases.
    The WRNMS (a) does not decrease a channel trip occurrence beyond 
its acceptable limit, (b) does not increase a channel response time 
beyond its acceptable limit, (c) increases indicated accuracies, and 
(d) does not cause any plant parameter for any analyzed event to 
fall outside of its acceptable limit(s).
    The surveillance test frequency change of 7 to 31 days is based 
on the WRNMS having (1) fixed in-core detectors, (2) greater 
reliability than the SRMs and IRMs, and (3) self test features. The 
13 second allowable value for the WRNM Period-Short surveillance, 
and the surveillance test frequency change of 184 days to 24 months 
is based on trip setpoint calculations using GE's standard (NRC 
approved) setpoint methodology.
    The WRNMS will not involve a reduction in a margin of safety, as 
loads on plant equipment will not increase, and reactions to or 
results of transients and postulated accidents will not increase 
from those presently approved by the NRC.

    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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna 
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

    Date of amendment request: April 11, 1995.
    Description of amendment request: This amendment would extend on a 
one time basis the allowed outage time in the Susquehanna Steam 
Electric Station Technical Specification 3.8.1.1 from 3 to 7 days for 
one offsite circuit being out of service. This change will provide 
additional time if needed to complete modifications to an offsite 
circuit.
    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:

    I. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The consequences of losing offsite power have been evaluated in 
the FSAR [Final Safety Analysis Report] and the Station Blackout 
evaluation. Increasing the AOT [allowed outage time] for T-10 [an 
offsite power source] from 3 to 7 days does not increase the 
consequences of a LOOP [loss of offsite power] event nor change the 
evaluation of LOOP events as stated in the FSAR or Station Blackout 
evaluation.
    Allowing T-10 to be removed from service for an additional 4 
days does increase slightly the possibility of a LOOP event as shown 
in PP&L's [Pennsylvania Power & Light Company's] engineering study. 
However, implementing the following compensatory actions reduces the 
probability of a LOOP event:
    1. prohibiting high risk activities within the confines of the 
plant or the grid system that may result in a loss of T-20 [the 
second offsite power transformer] during the T-10 outage,
    2. performing the modification during the Fall when the 
frequency of grid and weather related LOOPs are reduced,
    3. requiring a unit shutdown if the HPCI [high pressure core 
injection] system becomes inoperable during the T-10 outage,
    4. requiring a unit shutdown if the SLCS [standby liquid control 
system] becomes inoperable during the T-10 outage,
    5. requiring that within 24 hours prior to taking T-10 out of 
service, Surveillance 4.8.1.1.2.a.4 be successfully completed on the 
aligned diesel generators, and
    6. maintaining the following equipment operable during the T-10 
work window and restoring any failed system/component to operable 
status as soon as possible (The failed system/component shall be 
worked around the clock):
     Both CRD [control rod drive] pumps,
     Diesel fire pump, yard fire hydrant (1FH122) and 
associated hydrant hose station,
     RHR [residual heat removal system]/RHRSW [residual heat 
removal service water system]/ESW [emergency service water system] 
for suppression pool cooling,
     RHR/RHRSW cross tie valves,
     RCIC [reactor core isolation cooling]
     CIG [containment instrument gas] 150 psig header and 
bottles,
     Turbine Building Closed Cooling Water System (one pump 
and heat exchanger),
     Portable diesel generator,
     HV-141-F019.
    Therefore, this change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    II. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Allowing the AOT for T-10 to increase from 3 to 7 days is a one 
time exemption in order to install the new T-10 tap and 230 kV 
switch yard. The accident analyses affected by this extension are 
the LOOP events. The remaining portions of the station and equipment 
are not altered by this change. The potential for the loss of other 
plant systems or equipment to mitigate the effects of an accident 
are not altered. One offsite source of power will be out of service 
for an additional 4 days and compensatory actions will be initiated 
to lessen the effect of having the offsite power source out of 
service for an additional 4 days. Therefore, this change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    III. Involve a significant reduction in a margin of safety.
    The proposed change allows, on a one time basis, T-10 to be out 
of service for an additional 4 days. This increase in AOT for T-10 
results in a slight decrease in the margin of safety (defined as 
core damage frequency) with respect to having two offsite sources 
available per Specification 3.8.1.1. By implementing the 
compensatory measures as described in Item 1 above, the margin of 
safety is increased to be the equivalent of allowing the offsite 
power source (T-10) to be out of service for 3 days as is allowed by 
the existing Specification. Therefore, this one time exemption will 
not involve a significant reduction in safety margin.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
[[Page 29887]] satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: April 10, 1995.
    Description of amendment request: This amendment would relocate 
response time limit tables from the Susquehanna Steam Electric Station 
Unit 1 and Unit 2 Technical Specifications (TS) to the Final Safety 
Analysis Report. This modification is a line item improvement to the TS 
as described in Generic Letter 93-08, ``Relocation of the Technical 
Specification Tables of Instrument Response Time Limits,'' dated 
December 29, 1993.
    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:

    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The purpose of the proposed Tech. Spec. [Technical 
Specification] change is to delete and subsequently relocate Tech. 
Spec. Table 3.3.1-2, Table 3.3.2-3, and Table 3.3.3-3, to the SSES 
FSAR consistent with the guidance provided in Generic Letter 93-08. 
This is a line item Tech. Spec. improvement change recommended by 
the NRC in Generic Letter 93-08. This change will allow PP&L 
[Pennsylvania Power & Light Company] to administratively control 
subsequent changes to the response time limits in accordance with 
10CFR50.59. The procedures that contain the various response time 
limits are also subject to the change control provisions in the 
Administrative Controls section of the Tech. Specs. The proposed 
change only relocates the existing response time limits; the 
surveillance requirements and associated Actions are not affected 
and remain in the Tech. Specs. Relocating the response time limit 
information does not affect the analysis of any design basis 
accident. The response times of these systems will be maintained 
within the acceptance limits assumed in SSES [Susquehanna Steam 
Electric Station] safety analyses and required for successful 
mitigation of an initiating event. Also, since any subsequent 
changes to the FSAR or procedures will be evaluated in accordance 
with 10 CFR 50.59, no increase in the probability or consequences of 
an accident previously evaluated will occur. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    II. This proposal does not increase the possibility of a new or 
different kind of accident from any accident previously evaluated.
    As discussed above, the proposed Tech. Spec. changes do not 
affect the capability of the associated systems to perform their 
intended function within the acceptance limits assumed in SSES 
safety analyses and required for successful mitigation of an 
initiating event. The proposed change does not involve a physical 
modification of the plant or changes in methods governing normal 
plant operations. The proposed change will not impose any different 
operational or surveillance requirements. This change only proposes 
to relocate these requirements to other plant documents whereby 
adequate control of information will be maintained. No new failure 
modes will be introduced. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumption. The proposed 
change does not alter the scope of equipment currently required to 
be OPERABLE or subject to testing, nor does the proposed change 
affect any instrument setpoints or equipment safety functions. Since 
any future changes to these requirements in the FSAR or procedures 
will be evaluated per the requirements of 10 CFR 50.59, no reduction 
in a margin of safety will occur. 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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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 2, 1995.
    Description of amendment request: The amendments would eliminate 
the manual start for auxiliary feedwater from the Technical 
Specification for Engineered Safety Feature (ESF) Actuation System 
Instrumentation. The manual start will be tested during the quarterly 
pump test. This change is consistent with NUREG-1431, ``Standard 
Technical Specifications- Westinghouse Plants.''
    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 analyzed.
    The change to the ESF Actuation Instrumentation specification to 
eliminate the requirements for manual initiation of the [Auxiliary 
Feedwater] (AFW) Pumps does not change any operating characteristics 
of the plant. The change will eliminate unnecessary AFW Pump starts 
which increase wear on system components. Manual initiation is not 
credited in the Salem safety analyses. Manual initiation is verified 
quarterly on a staggered test basis by performance of specification 
4.7.1.2.b. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of 
accident.
    The proposed technical specification modifications do not change 
system configurations, plant equipment or safety analyses. 
Therefore, the proposed modifications will not increase the 
possibility of a new or different kind of accident from any accident 
previously identified.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the ESF Actuation Instrumentation 
Specification does not affect the ability of the AFW System to 
perform its design function. The manual initiation of the AFW Pump 
is not credited in the Salem safety analyses. Manual initiation is 
verified quarterly by performance of specification 4.7.1.2.b. 
Therefore, these changes do not involve 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 
[[Page 29888]] amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: January 4, 1995 (TS 355).
    Description of amendment request: The proposed amendment changes 
the applicability and surveillance requirements for the intermediate 
range monitor (IRM), average power range monitor (APRM), and APRM 
Inoperative Trip functions. The proposed amendment adopts provisions of 
the Improved Standard Technical Specifications (NUREG-1433).
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change revises the frequency of functional tests 
for the IRM and APRM High Flux (15% Scram) Trip Functions and 
eliminates operability requirements for the IRM, APRM High Flux (15% 
Scram), and APRM Inoperative Trip Functions in certain modes of 
operation. The operation of these trip functions is not a precursor 
to any design basis accident or transient analyzed in the Browns 
Ferry Updated Final Safety Analysis Report. Therefore, this change 
does not increase the probability of any previously evaluated 
accident.
    The proposed change will eliminate the requirement to re-perform 
the functional tests for these trip functions prior to each startup 
if the test is within its periodicity (once per 7 days). It will 
also eliminate the operability requirement for the IRM High Flux 
Trip Function in the Shutdown Mode and IRM, APRM High Flux (15% 
Scram), and APRM Inoperative Trip Functions during the Refuel Mode 
except when any control rod is withdrawn from a core cell containing 
one or more fuel assemblies. The Specifications will still provide 
for operability of the equipment in Modes where credit is taken in 
the safety analysis. Therefore, this change does not increase the 
consequences of any previously evaluated 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 change to the Technical Specification requirements 
for the IRM, APRM High Flux (15% Scram) and APRM Inoperative Trip 
Functions does not involve a modification to plant equipment. No new 
failure modes are introduced. There is no effect on the function of 
any plant system and no new system interactions are introduced by 
this change. Therefore, the proposed amendment 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 a margin of safety.
    The proposed change will eliminate the requirement to re-perform 
the functional test for the IRM and APRM High Flux (15% Scram) Trip 
Functions prior to each startup if the tests are within their 
periodicity (once per 7 days). The proposed change will also 
eliminate operability requirements for modes of operation in which 
the IRM, APRM High Flux (15% Scram) and APRM Inoperative Trip 
Functions provide no useful function. Since the ability of the trip 
functions to perform their safety function will not be degraded, the 
proposed amendment does not involve a 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: Athens Public Library, South 
Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: March 31, 1995 (TS 349).
    Description of amendment request: The proposed amendment changes 
the reactor pressure vessel pressure-temperature (P-T) curves, lowering 
the temperature at which the reactor vessel head bolting studs may be 
tensioned.
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed Units 1, 2, and 3 change deals exclusively with the 
reactor vessel P-T [pressure-temperature] curves, which define the 
permissible regions for operation and testing. Failure of the 
reactor vessel is not a design basis accident. Through the design 
conservatism used to calculate the P-T curves, reactor vessel 
failure has a low probability of occurrence and is not considered in 
the safety analyses. These changes do not alter or prevent the 
operation of equipment required to mitigate any accident analyzed in 
the BFN Browns Ferry Nuclear Plant] Final Safety Analysis Report. 
Therefore, this change does not increase the probability or 
consequences of any previously evaluated 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 change to the Units 1, 2, and 3 reactor vessel P-T 
curves does not involve a modification to plant equipment. No new 
failure modes are introduced. There is no effect on the function of 
any plant system and no new system interactions are introduced by 
this change. The calculation of the proposed P-T curves was in 
accordance with Regulatory Guide 1.99, Revision 2, and the 
requirements of 10 CFR 50, Appendix G. Therefore, the proposed 
amendment 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 a margin of safety.
    The ductile to brittle transition temperature is shifted 
approximately 10 deg.F at higher temperatures and approximately 
30 deg.F at lower temperatures on the proposed P-T curves. While 
this represents a decreased margin against non-ductile fracture 
during heatup, cooldown and hydrotesting, the proposed curves 
conform to the guidance contained in Regulatory Guide 1.99, Revision 
2, and maintain the safety margins specified in 10 CFR 50, Appendix 
G. Therefore, the proposed amendment 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: Athens Public Library, South 
Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon. [[Page 29889]] 

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

    Date of amendment request: May 11, 1995 (TS 359).
    Description of amendment request: The proposed amendment adds a 
scram pilot air header low pressure reactor trip to Browns Ferry Unit 
3. The proposed amendment also clarifies a note regarding reactor 
protection system instrumentation requirements for all three 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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The scram pilot air header low pressure switches perform the 
same function as the high water level switches in the scram charge 
instrument volume. They automatically initiate control rod insertion 
(SCRAM) in the event that degraded conditions are detected in the 
BWR [boiling water reactor] CRD [control rod drive] System. Since 
the scram pilot air header pressure trip function ensures that the 
CRD System is available to mitigate the consequence of an accident 
or transient, and the addition of the scram pilot air header low 
pressure trip scram function does not affect the precursors for any 
accident or transient analyzed in Chapter 14 of the BFN Updated 
Final Safety Analysis Report (UFSAR), there is no increase in the 
probability of any accident previously evaluated.
    The design criteria for the scram system is contained in the 
generic SER [safety evaluation report], which was transmitted by NRC 
letter to All BWR Licensees, dated December 9, 1980, BWR Scram 
Discharge System. The scram pilot air header pressure trip function 
ensures that the CRD System is available to mitigate the consequence 
of an accident or transient, and the overall scram system design, 
with the addition of the scram pilot air header low pressure trip 
function, satisfies the criteria contained in the generic SER. Since 
the scram function would be successfully performed, the addition of 
the scram pilot air header low pressure trip scram function does not 
involve a significant increase in the consequences of any accident 
previously evaluated.
    The clarification of the description of the SDV [scram discharge 
volume] high water level bypass in the RPS [reactor protection 
system] does not, by itself, reflect a modification to plant 
equipment, maintenance activities, or operating instructions. The 
revised description does not effect the precursors for any accident 
or transient analyzed in Chapter 14 of the BFN UFSAR or equipment 
used in the mitigation of these accidents or transients. Therefore, 
there is no increase in the probability of any accident previously 
evaluated nor an increase in the consequences of any accident 
previously evaluated.
    [2]. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The scram pilot air header low pressure trip performs the same 
protective function as the SDV high water level trip. Both trip 
functions ensure that a reactor scram is initiated while sufficient 
volume remains in the SDV to accept discharged water from the CRDs.
    The scram inlet and outlet valves are held closed by the air 
pressure in the scram air header. The scram outlet valves begin to 
unseat as the air pressure drops below 43 psig (which is higher than 
the pressure that scram inlet valves begin to unseat). The scram 
pilot air header low pressure switches detect losses in air pressure 
and initiate an anticipatory scram to ensure the scram is complete 
prior to the possible onset of hydraulic locking in the SDV. The 
proposed trip level setting of 50 psig is conservative and assures a 
trip signal and successful reactor scram is accomplished prior to 
hydraulic locking occurring in the SDV as a result of significant 
flow past the scram outlet valves.
    The overall scram system design, with the addition of the scram 
pilot air header low pressure trip function is in conformance with 
the generic SER. No new system failure modes are created as a result 
of adding the scram pilot air header low pressure trip scram 
function. Therefore, the addition of the scram pilot air header low 
pressure trip scram function does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The clarification of the description of the SDV high water level 
bypass in the RPS does not, by itself, reflect a modification to 
plant equipment, maintenance activities, or operating instructions. 
No new external threats, system interactions, release pathways, or 
equipment failure modes are created. Therefore, the clarification of 
this description 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 a margin of safety.
    The overall scram system design, with the addition of the scram 
pilot air header low pressure trip function is in conformance with 
the generic SER. Since the scram system would successfully operate 
to mitigate the consequences of accidents and transients previously 
analyzed, the proposed amendment does not involve a significant 
reduction in the margin of safety.
    The clarification of the description of the SDV high water level 
bypass in the RPS does not, by itself, reflect a modification to 
plant equipment, maintenance activities, or operating instructions. 
There is no change to the licensing or design basis of the RPS. 
Therefore, the revised description 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: Athens Public Library, South 
Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.
The Cleveland Electric Illuminating Company, Centerior Service Company, 
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power 
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power 
Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: April 28, 1995.
    Description of amendment request: The proposed amendment would 
remove the license conditions for the Transamerica Delaval, Inc. 
emergency diesel generators specified by paragraph 2.C.(9) and defined 
in Attachment 2 to the Operating License.
    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 removal of license conditions 
associated with teardowns and certain inspections on the 
Transamerica Delaval, Inc. (TDI) Emergency Diesel Generators (EDG). 
A failure of an EDG is not an initiating event for any Updated 
Safety Analysis Report (USAR) Chapter 15 accident scenario. 
Accordingly, there could be no increase in the probability of any 
accident previously evaluated. The availability and reliability of 
the EDGs will remain within the limits previously assumed in the 
safety analyses. Eliminating the disassembly and specified 
inspections would actually tend to decrease the consequences of an 
accident because, as indicated in Topical Report TDI-EDG-001-A, 
``Basis for Modification to Inspection Requirements for Transamerica 
Delaval, Inc., Emergency Diesel Generators,'' this action will 
improve the availability of the engines for service, especially 
during outages, while maintaining current reliability levels. 
Therefore, removal of the existing conditions from the operating 
license will not result in an increase in the consequences of an 
accident previously evaluated. [[Page 29890]] 
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed removal of the license conditions associated with 
the TDI diesel generators does not affect the design or function of 
any plant system, structure, or component, nor does it change the 
way plant systems are operated. No modifications or additions to 
plant equipment are involved. Therefore, removal of the existing 
conditions from the operating license will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not result in a significant 
reduction in the margin of safety.
    The proposed removal of the EDG license conditions from the 
Operating License does not affect any parameters which would result 
in a significant reduction in margin of safety because the results 
of the operational data and inspections have demonstrated that the 
additional license conditions are not required to ensure that the 
EDGs will be maintained with a reliability consistent with that 
assumed for the safety analyses. Therefore, the proposed changes do 
not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of amendment request: May 2, 1995.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Table 15.4.1, ``Minimum Frequencies 
for Checks, Calibrations, and Tests of Instrument Channels.'' The 
radiation monitoring system channel requirements would be deleted, the 
main steam line radiation channel requirements would be added, and the 
containment high range radiation channel requirements would be changed. 
Administrative changes, consistent with the proposed modifications, 
would also be made.
    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 radiation monitors being removed from Table 15.4.1-1 
are not directly involved with mitigating an offsite release in the 
case of an accident. The surveillance requirements for monitors 
which would measure and mitigate such a release are listed in 
Technical Specifications Section 15.7.4, ``Radioactive Effluent 
Monitoring Instrumentation Surveillance Requirements.'' Post-
accident radiation monitors will still be included in Table 15.4.1-
1. Monitors to be removed include area and non-RETS [radiological 
effluent technical specification] required process monitors. These 
are necessary to monitor plant conditions and will still be subject 
to surveillance requirements. The removed monitors do not have any 
safety function with regard to radioactive releases. Therefore, the 
consequences of an accident will not be increased. The radiation 
monitors are not initiators for any accident analyses in the FSAR, 
therefore, the probability of an accident previously evaluated is 
not increased.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. There is no physical change to the facility, its systems, 
or its operation, therefore, a new or different kind of accident 
cannot occur.
    3. The proposed change will not involve a significant reduction 
in the margin of safety. The removal of much of the RMS equipment 
from the Technical Specifications will not affect the surveillance 
program already in place. The change in test frequency for the post-
accident monitoring instrumentation will not have a significant 
impact on the margin of safety. Test frequencies continue to meet 
acceptable standards. RETS required effluent monitors, which are of 
prime importance due to their release mitigation function, are 
checked quarterly in accordance with Technical Specifications 
Section 15.7.4, ``Radioactive Effluent Monitoring Instrumentation 
Surveillance Requirements.'' Therefore, the margin of safety is not 
reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that 
the amendment request involves no significant hazards consideration.

    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

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.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: May 15, 1995.
    Description of amendment request: The proposed amendment would 
authorize a reconfiguration of the cooling water flow to the reactor 
building emergency cooling system.
    Date of individual notice in the Federal Register: May 22, 1995 (60 
FR 27144)
    Expiration date of individual notice: June 21, 1995.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated. 
[[Page 29891]] 
    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.

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

    Date of application for amendments: November 2, 1994.
    Brief description of amendments: These amendments delete the 
condenser vacuum exhaust release point reference on Figure 5.1-3 and 
combine it with the plant vent exhaust release point on the revised 
Figure 5.1-3. In addition to the figure change, Bases Section 3/4.3.3.6 
is changed to reflect the removal of radiation monitor RU-142 and the 
relocation of RU-144 and RU-146 from Table 3.3-13 (deleted by 
amendments 62, 48, and 34, for Units 1, 2, and 3, respectively) to the 
Offsite Dose Calculation Manual.
    Date of issuance: May 25, 1995.
    Effective date: May 25, 1995, to be implemented within 45 days of 
issuance.
    Amendment Nos.: Unit 1--Amendment No. 91; Unit 2--Amendment No. 79; 
Unit 3--Amendment No. 62.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65810). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 25, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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: April 13, 1994, as supplemented 
December 20, 1994, January 12, January 31, March 17, and April 5, 1995. 
Brief description of amendment: The amendment revises TS Sections 3.1.F 
and 4.13 to allow the repair of steam generator tubes by sleeving using 
laser welded sleeves.
    Date of issuance: May 19, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 183.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27051). The December 20, 1994, January 12, January 31, March 17, and 
April 5, 1995, submittals provided clarifying information that did not 
change the initial no significant hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 19, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: December 29, 1994, as 
supplemented February 2 and May 4, 1995.
    Brief description of amendment: This amendment revises the iodine 
removal system Technical Specification (TS) to reflect replacement of 
the sodium hydroxide requirements with trisodium phosphate 
requirements. The revised TS defines operability, applicability, and 
associated action statements for the new system. Associated 
surveillance requirements and bases have also been revised.
    Date of issuance: May 19, 1995.
    Effective date: May 19, 1995.
    Amendment No.: 165.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6299). The February 2 and May 4, 1995, submittals provided clarifying 
information which was within the scope of the initial application and 
did not affect the staff's initial proposed no significant hazards 
consideration findings. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated May 19, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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: April 19, 1994, as supplemented 
March 31, 1995.
    Brief description of amendments: These amendments revise the 
Appendix A Technical Specifications (TSs) 3.4.9.3 and 3.4.11 to 
incorporate changes to the power operated relief valve TSs in 
accordance with the guidance in Generic Letter 90-06, ``Resolution of 
Generic Issue 70, ``Power-Operated Relief Valve and Block Valve 
Reliability,'' and Generic Issue 94, ``Additional Low-Temperature 
Overpressure Protection for Light-Water Reactors,'' Pursuant to 10 CFR 
50.54(f),'' as implemented in the NRC's Improved Standard Technical 
Specifications (NUREG-1431) with some exceptions and modifications to 
reflect plant-specific design features. The amendment includes several 
administrative changes (e.g., renumbering sections, spelling out 
mathematical symbols, changes in nomenclature for consistency, and 
relocation of sentences and paragraphs).
    Date of issuance: May 15, 1995.
    Effective date: May 15, 1995.
    Amendment Nos.: 187 and 69.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34661). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 15, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001. [[Page 29892]] 

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: July 22, 1993.
    Brief description of amendment: The amendment revised the value of 
the Unit 1 reactor building volume as listed in the technical 
specifications. The amendment was submitted after a more precise 
calculation of the reactor building volume was completed.
    Date of issuance: May 22, 1995.
    Effective date: May 22, 1995.
    Amendment No.: 181.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 76843). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 22, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & Light 
Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, 
Claiborne County, Mississippi

    Date of application for amendment: June 25, 1993, as supplemented 
by letter dated April 13, 1995.
    Brief description of amendment: This amendment deleted portions of 
the current Technical Specifications (TSs) Surveillance Requirements 
(SRs) for the inboard Main Steamline Isolation Valve Leakage Control 
System (MSIV-LCS) heaters and blowers. The deleted MSIV-LCS SRs will be 
relocated to documents that are included by reference in the Updated 
Final Safety Analysis Report (UFSAR) and are controlled by the licensee 
under the provisions of 10 CFR 50.59. The change is consistent with the 
format and content of the Improved Standard Technical Specifications 
(NUREG-1434, Revision O).
    Date of issuance: May 22, 1995.
    Effective date: May 22, 1995.
    Amendment No. 122.
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39050). The additional information contained in the supplemental letter 
dated April 13, 1995, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 22, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric 
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: March 18, 1994, as supplemented 
by letters dated February 28 and March 17, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.3.3.6, Accident Monitoring Instrumentation, TS 
3/4.6.4.1, Hydrogen Monitors, and their associated Bases to incorporate 
the technical substance of Specification 3.3.3 from NUREG-1431, 
Revision O (Standard Technical Specifications) for the Westinghouse 
Owners Group.
    Date of issuance: May 15, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 85 and 63.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22008). The February 28 and March 17, 1995, letters provided clarifying 
information that did not change the scope of the March 18, 1994, 
application and initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 15, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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: March 16, 1995.
    Brief description of amendments: The amendments revised Technical 
Specification 4.6.1.2, regarding the test frequency requirements for 
the overall integrated containment leakage rate tests, so that it 
references 10 CFR part 50, appendix J and approved exemptions, rather 
than paraphrase the regulation.
    Date of issuance: May 19, 1995.
    Effective date: May 19, 1995.
    Amendment Nos.: Unit 1--Amendment No. 75; Unit 2--Amendment No. 64.
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20517). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 19, 1995. 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, Texas 77488.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of application for amendment: April 14, 1995.
    Brief description of amendment: The amendment allows the use of the 
Westinghouse Electric Corporation sleeving process for repairing steam 
generator tubes.
    Date of issuance: May 22, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 150.
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1995 (60 FR 
19969). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 22, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire
    Date of amendment request: January 25, 1995, and oral request of 
May 16, 1995.
    Description of amendment request: This amendment revises the 
Appendix [[Page 29893]] A Technical Specifications (TS) relating to the 
schedule for performing Type A containment Integrated Leak Rate Tests 
(ILRTs). Specifically, the amendment replaces the prescribed number of 
ILRTs to be performed and the associated schedule with the requirement 
to conduct ILRTs at intervals as specified in Appendix J to 10 CFR Part 
50.
    Date of issuance: May 17, 1995.
    Effective date: May 17, 1995.
    Amendment No.: 37.
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8754). The licensee's oral request of May 16, 1995, provided a minor 
clarifying addition, but does 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 17, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: October 4, 1993.
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications (TS) relating to A.C. power sources 
during operation in Modes 1 through 4. Specifically, the amendment 
deletes the diesel engine speed specification from Surveillance 
Requirement (SR) 4.8.1.1.2a.5 and replaces the diesel engine speed 
requirement with an electrical frequency requirement in SR 4.8.1.1.2g.
    Date of issuance: May 19, 1995.
    Effective date: As of the date of its issuance, to be implemented 
within 60 days of issuance.
    Amendment No.: 38.
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4941). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 19, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 47 
Front Street, Exeter, NH 03833.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: October 7, 1994.
    Description of amendment request: The amendment modifies Technical 
Specification (TS) 6.4.1.6, 6.4.3.8, and 6.7.1 relating to 
Administrative Controls. Specifically, the amendment removes certain 
audit responsibilities of the Nuclear Safety Audit Review Committee and 
certain review responsibilities of the Station Operation Review 
Committee relating to the Emergency Plan and the Security Plan and 
their implementing procedures, and deletes the requirements for written 
procedures relating to the Emergency Plan and Security Plan.
    Date of issuance: May 19, 1995.
    Effective date: May 19, 1995.
    Amendment No.: 39.
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63125). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 19, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

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

    Date of application for amendment: May 6, 1994, supplemented March 
27, 1995.
    Brief description of amendment: The amendment incorporates 
additional sections and their associated surveillance requirements and 
bases into the Millstone Unit 2 TS that impose additional requirements 
on components that are credited to provide feedwater isolation in the 
event of a main steam line break inside containment. In addition, the 
amendment makes modifications to the TS Bases Sections \3/4\.3.1 and 
\3/4\.3.2 by denoting that the feedwater pumps are assumed to trip 
immediately upon receipt of a main steam line isolation signal; and 
makes several miscellaneous editorial changes.
    Date of issuance: May 17, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 188.
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32232). The March 27, 1995, letter 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 17, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut Date 
of application for amendment: April 21, 1994.

    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.1.2.4, ``Charging Pumps-Operating,'' by adding a 
note that indicates that the provisions of TS 3.0.4 and 4.0.4 are not 
applicable for entry into MODE 4 from MODE 5.
    Date of issuance: May 18, 1995.
    Effective date: As of the date of issuance.
    Amendment No.: 189.
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (60 FR 21558, May 2, 1995). That notice provided an 
opportunity to submit comments on the Commission's proposed no 
significant hazards consideration determination. No comments have been 
received. The notice also provided for an opportunity to request a 
hearing by June 1, 1995, but indicated that if the Commission makes a 
final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated May 
18, 1995.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360. [[Page 29894]] 

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: December 9, 1994, as 
supplemented March 28, 1995.
    Brief description of amendment: The amendment eliminates certain 
surveillance requirements for the emergency diesel generators, in 
accordance with staff guidance contained in Generic Letter 93-05, 
``Line Item Technical Specification Improvements to Reduce Surveillance 
Requirements for Testing during Power Operation,'' dated September 27, 
1993.
    Date of issuance: May 12, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 112.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8749). The March 28, 1995, letter 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 12, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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: January 18, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications to increase the minimum required boron concentration in 
the boric acid tank (BAT) from 6300 to 6600 ppm. The increase is 
required to meet the latest analysis for Cycle 6 which includes 
additional conservatisms which are meant to ensure the new required 
boron concentration will bound future cycle variations.
    Date of issuance: May 17, 1995.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 113.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8753). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 17, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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: January 24, 1995, as 
supplemented
    March 22 and 29, 1995, and April 25, 1995.
    Brief description of amendment: The amendment revises Technical 
Specification 3.2.3.1.a and Table 2.2-1 to reduce the minimum reactor 
coolant system (RCS) flow rate by 4%, with corresponding changes in 
loop flow. The current minimum RCS flow rate of 387,480 gallons per 
minute (gpm) is reduced to 371,920 gpm for four-loop operation.
    Date of issuance: May 23, 1995.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 114.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11136) and April 12, 1995 (60 FR 18626). The April 25, 1995, letter 
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 23, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: January 9, 1995, as 
supplemented February 7, March 15, March 27, April 3, and April 20, 
1995.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) for the Prairie Island Nuclear Plant to 
allow using an alternate steam generator tube plugging criteria (F*) 
for the part of the tubes within the tubesheet. The amendments 
incorporate revised acceptance criteria (F*) for tubes with degradation 
in the tubesheet roll expansion and enable the licensee to avoid 
unnecessary plugging of steam generator tubes. NRC will issue a 
separate safety evaluation for the L* criteria at a later date.
    Date of issuance: May 15, 1995.
    Effective date: May 15, 1995, with full implementation within 30 
days.
    Amendment Nos.: 118/111.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14023). The March 15, March 22, April 3, and April 20, 1995, letters 
provided updated TS pages and clarifying information in response to 
NRC's requests for additional information. This information was within 
the scope of the original application and did not change the staff's 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated May 15, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    PECO Energy Company, Public Service Electric and Gas Company 
Delmarva Power and Light Company, and Atlantic City Electric Company, 
Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit 
Nos. 2 and 3, York County, Pennsylvania
    Date of application for amendments: August 3, 1994.
    Brief description of amendments: The amendments implement a snubber 
functional test surveillance interval of 24 months. The amendments 
change the current one-time snubber functional test interval to a 
permanent interval of 24 months.
    Date of issuance: May 16, 1995.
    Effective date: May 16, 1995.
    Amendments Nos.: 201 and 204.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 18, 1995 (60 FR 
3676).The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 16, 1995. No significant 
hazards consideration comments received: No. [[Page 29895]] 
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    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 application for amendments: September 29, 1994.
    Brief description of amendments: The amendments remove from the 
Technical Specifications the sections entitled ``Seismic 
Instrumentation'' and ``Meteorological Instrumentation'' and relocate 
the information and testing requirements to the Salem Updated Final 
Safety Analysis Report.
    Date of issuance: May 22, 1995.
    Effective date: May 22, 1995.
    Amendment Nos. 167 and 149.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60385). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 22, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.
    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: November 3, 1993.
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 3/4.6.3, ``Containment Isolation Valves,'' to 
require valves listed in Section D of existing Table 3.6-1, 
``Containment Isolation Valves,'' to be in an action statement when 
secured in their engineered safety feature actuation system (ESFAS) 
actuated position. Bases 3/4.6.3 is also revised to reflect these 
changes.
    Date of issuance: May 17, 1995.
    Effective date: May 17, 1995, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 2--Amendment No. 119; Unit 3--Amendment No. 
108.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7699). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 17, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.
    Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia
    Date of application for amendments: October 25, 1994.
    Brief description of amendments: The amendments revise the NA-1&2 
Hydrogen Recombiner System surveillance requirements in accordance with 
Generic Letter 93-05, ``Line-Item Technical Specification Improvements 
to Reduce Surveillance Requirements for Testing During Power 
Operation.'' Also, the amendments delete the surveillance requirement 
to operate the containment purge blower and clarifies that the 
surveillance requirement applies only to the hydrogen recombiner purge 
blowers.
    Date of issuance: May 12, 1995.
    Effective date: May 12, 1995.
    Amendment Nos.: 192 and 173.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60388). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 12, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia
    Date of application for amendments: March 2, 1995.
    Brief description of amendments: The amendments revise the NA-1&2 
Technical Specification 4.6.1.2.a to permit approved exemptions to the 
containment integrated leak rate test frequency requirements.
    Date of issuance: May 15, 1995.
    Effective date: May 15, 1995.
    Amendment Nos.: 193 and 174.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18629). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 15, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington
    Date of application for amendments: September 2, 1992.
    Brief description of amendment: The amendment revises Figure 3.1.5-
2, ``Sodium Pentaborate Tank, Volume Vs. Concentration Requirements,'' 
to reflect the actual low-volume-alarm and low-limit values for the 
standby liquid control tank.
    Date of issuance: May 17, 1995.
    Effective date: May 17, 1995, to be implemented within 30 days of 
issuance.
    Amendment No.: 138.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60388). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 17, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

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

    Date of amendment request: March 21, 1995.
    Brief description of amendment: This amendment revises Technical 
Specification Surveillance Requirement 4.6.2.1.d, ``Containment Spray 
System,'' to change the surveillance interval specified for the 
performance of an air or smoke flow test through the containment spray 
header from at least 5 years to at least once per 10 years.
    Date of issuance: May 17, 1995.
    Effective date: May 17, 1995, to be implemented within 30 days of 
issuance.
    Amendment No.: 86.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18631). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 17, 1995. No 
[[Page 29896]] significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

    Dated at Rockville, Maryland, this 31st day of May, 1995.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 95-13759 Filed 6-5-95; 8:45 am]
BILLING CODE 7590-01-P