[Federal Register Volume 62, Number 58 (Wednesday, March 26, 1997)]
[Notices]
[Pages 14457-14476]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10326]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 3, 1997, through March 14, 1997. The 
last biweekly notice was published on March 12, 1997.

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 Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be

[[Page 14458]]

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 April 25, 1997, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. 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 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of amendments request: January 15, 1997
    Description of amendments request: The proposed change would revise 
the values of the minimum and maximum suppression pool water volumes 
corresponding to the upper and lower limits of the suppression water 
levels specified in TS 3.6.2.1.a.1 such that the implementation of the 
administrative controls will no longer be necessary to ensure 
compliance with the Technical Specifications.
    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:

[[Page 14459]]

    1. The proposed amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change revises the values of the minimum and 
maximum suppression pool water volume limits. The water inventory of 
the suppression chamber is not a precursor of an accident and, 
therefore, cannot increase the probability of an accident previously 
evaluated. The pressure suppression chamber water pool mitigates the 
consequences of loss-of-coolant accidents (LOCAs), transients, and 
other events by providing a heat sink for reactor primary system 
energy releases. The proposed minimum and maximum pool water volume 
values will be consistent with the current suppression pool water 
level limits. No changes to setpoints will be made as a result of 
the proposed change. The impact of the proposed change to the 
minimum and maximum suppression pool volume limits on the 
suppression pool temperatures and pressures following a design basis 
LOCA, an SRV [Safety Relief Valve] blowdown event, an Anticipated 
Transient Without Scram (ATWS) event, an Appendix R fire event, and 
a station blackout event has been evaluated and does not cause 
accident parameters to exceed acceptable values. In addition, the 
impact the proposed change has on the time to reach cold shutdown 
when using the alternate RHR [Residual Heat Removal] shutdown 
cooling function is negligible.
    The potential impact the proposed change to the suppression pool 
water volume limits has on SRV line loads, SRV discharge line 
reflood height, wetwell pressurization, suppression pool swell 
loads, vent thrust loads, and condensation oscillation and chugging 
loads was also reviewed. The proposed change to the suppression pool 
water volume limits has no adverse impact on any of these 
parameters.
    The capability of the suppression chamber water pool to perform 
its mitigative functions is not affected by the proposed
    change. Therefore, the proposed change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. The proposed amendment[s] would not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the values of the minimum and 
maximum volume of the suppression chamber water pool. The proposed 
change will not alter any physical mechanism by which the 
suppression chamber water pool volume is maintained between the 
minimum and maximum values. The suppression pool water level will 
continue to be maintained between -27 and -31 inches. As a result of 
the proposed change there are no physical changes to suppression 
chamber components or instrumentation. No new mode of operation is 
introduced as a result of the proposed change. Analyses have been 
performed which conclude that the proposed change would not affect 
the operability of equipment designed to mitigate the consequences 
of an accident. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed license amendment[s do] not involve a 
significant reduction in a margin of safety.
    The proposed change revises the values of the minimum and 
maximum suppression chamber water pool volumes. The pressure 
suppression chamber water pool mitigates the consequences of several 
postulated accidents and transients by providing a heat sink for the 
primary coolant system. These accidents and events are the 
postulated design basis LOCA, Safety Relief Valve blowdown, ATWS, 
Appendix R fire and station blackout events. The consequences of the 
proposed change in the suppression pool water volume limits have 
been evaluated for these events.
    The results of the analyses for the postulated accidents and 
events indicate the temperature of the suppression pool water could 
increase slightly as a consequence of the decrease in the minimum 
suppression pool water volume limit. However, the containment 
temperatures remain within acceptable values. The impact of the 
calculated increase in containment temperature on the available Net 
Positive Suction head (NPSH) for the Residual Heat Removal (RHR) and 
Core Spray pumps has been evaluated for the postulated design basis 
LOCA and indicate adequate NPSH is maintained throughout the event.
    The potential impact of the proposed change to the suppression 
pool water volume limits on SRV line loads, SRV discharge line 
reflood height, wetwell pressurization, suppression pool swell 
loads, vent thrust loads, and condensation oscillation and chugging 
loads was evaluated with the conclusion that there are no adverse 
impacts on these parameters.
    In addition, a small suppression pool water temperature increase 
could result due to the reduction in the minimum suppression pool 
volume limit in the event reactor shutdown is conducted through a 
path utilizing the suppression pool. Such a shutdown path is an 
alternative to the normal RHR shutdown cooling function, and the 
small potential increase in temperature results in a negligible 
increase in the time required to reach cold shutdown conditions. 
Cold shutdown conditions could still be reached well within the 
Technical Specification requirements.
    The proposed increase in the suppression pool water volume limit 
does not adversely impact containment parameters as a result of 
postulated accidents and events. The potential increase in 
temperature of the pressure suppression pool water does not 
significantly decrease the ability to maintain containment 
parameters within acceptable limits. The potential increase in time 
to reach cold shutdown conditions utilizing the suppression pool as 
an alternative to the normal RHR shutdown cooling function is 
negligible. Therefore, the proposed change to revise the minimum and 
maximum suppression water pool volumes 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Mark Reinhart (Acting)

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of amendment request: March 14, 1997
    Description of amendment request: The proposed change revises 
Technical Specification 3/4.5.4, ``Refueling Water Storage Tank,'' and 
its associated Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The non-safety, non-seismic hydrotest pump is normally 
maintained separated from the RWST [Refueling Water Storage Tank] by 
a safety-related, locked closed manual operated boundary isolation 
valve (1CT-22). However, performance of Technical Specification 
required surveillance test OST-1506, ``Reactor Coolant System 
Isolation Valve Leak Test - 18 Month Interval- Mode 3,'' requires 
the short term use of the hydrotest pump during plant operating 
modes. Specifically, this hydrotest pump provides a high pressure 
source for leak testing the RCS [Reactor Coolant System] pressure 
isolation valves in Mode 3. The test is performed prior to entry 
into Mode 2, each refueling outage, whenever flow is established 
through the pressure isolation valves, or whenever the plant has 
been in cold shutdown for greater than 72 hours. Normally, the test 
is completed in less than 8 hours. Due to the piping configuration, 
a break in the non-seismic portion of the piping during these 
planned evolutions could result in draining the RWST below the 
minimum analyzed volume. Therefore to mitigate the consequences of a 
failure in the non-seismic piping, manual actions will be needed to 
isolate the break flow, (i.e., close valve 1CT-22), prior to 
reducing the water volume in the RWST below the minimum analyzed 
volume.

[[Page 14460]]

    Based on the use of a dedicated attendant to close valve 1CT-22, 
the lack of significant accessibility concerns, and the reliability 
of the valve to function, it can be concluded that 30 minutes is 
ample time for a valve attendant stationed at the valve to execute 
the manual action. Since the RWST volume margin provides up to 103 
minutes to respond to the pipe failure, it is reasonable to assume 
that manual actions to isolate the postulated pipe failure can be 
taken before the RWST level decreases below the minimum analyzed 
volume assumed in the safety analysis.
    Therefore, there would be no increase in the probability or 
consequences of an 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.
    Based on the use of a dedicated attendant to close valve 1CT-22, 
the lack of significant accessibility concerns, and the reliability 
of the valve to function, it can be concluded that 30 minutes is 
ample time for a valve attendant stationed at the valve to execute 
the manual action. Since the RWST volume margin provides up to 103 
minutes to respond to the pipe failure, it is reasonable to assume 
that manual actions to isolate the postulated pipe failure can be 
taken before the RWST level decreases below the minimum analyzed 
volume assumed in the safety analysis. As a result, the capability 
of the RWST to perform its safety function is not impacted.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    As described in the Technical Specification Bases, the 
operability of the RWST ensures that a sufficient supply of borated 
water is available for injection into the core by the emergency core 
cooling system. This borated water is used as cooling water for the 
core in the event of a LOCA [loss-of-coolant accident] and provides 
negative reactivty to counteract any positive increase in reactivity 
caused by reactor coolant system (RCS) cooldown. The limits on RWST 
minimum volume and boron concentration assure that: (1) sufficient 
water is available within containment to permit recirculation 
cooling flow to the core, and (2) the reactor will remain 
subcritical in the cold condition following mixing of the RWST and 
the RCS water volumes with all shutdown and control rods inserted 
except for the most reactive control assembly. These limits are 
consistent with the assumptions of the LOCA and steam line break 
analyses.
    Based on the use of a dedicated attendant to close valve 1CT-22, 
the lack of significant accessibility concerns, and the reliability 
of the valve to function, it can be concluded that 30 minutes is 
ample time for a valve attendant stationed at the valve to execute 
the manual action. Since the RWST volume margin provides up to 103 
minutes to respond to the pipe failure, it is reasonable to assume 
that manual actions to isolate the postulated pipe failure can be 
taken before the RWST level decreases below the minimum analyzed 
volume assumed in the safety analysis. As a result, the capability 
of the RWST to perform its safety function is not impacted.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Mark Reinhart, Acting

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

    Date of amendment request: February 17, 1997
    Description of amendment request: The proposed amendment would 
change the required diesel generator load during the initial 2 hours of 
a surveillance run from 2625 kW and 2750 kW to 2730 kW and 2860 kW.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    The proposed changes represent a correction to the emergency 
diesel generator surveillance requirement. The proposed changes are 
administrative in nature and do not significantly increase the 
probability or consequences of any previously evaluated accidents 
for Quad Cities Station. The proposed amendment is consistent with 
the current safety analyses and represents sufficient requirements 
for the assurance and reliability of equipment assumed to operate in 
the safety analysis. As such, these changes will not significantly 
increase the probability or consequences of a previously evaluated 
accident.
    The associated systems related to this proposed amendment are 
not assumed in any safety analysis to initiate any accident sequence 
for Quad Cities Station; therefore, the probability of any accident 
previously evaluated is not increased by the proposed amendment.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The proposed amendment for Quad Cities Station's Technical 
Specification is required to ensure the diesel generator is tested 
in accordance with the design basis requirements. The proposed 
changes do not create the possibility of a new or different kind of 
accident previously evaluated for Quad Cities Station. No new modes 
of operation are introduced by the proposed changes. The proposed 
changes are administrative in nature and 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 related to this proposed amendment are 
not assumed in any safety analysis to initiate any accident sequence 
for Quad Cities Station; therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    The proposed amendment is required to ensure the diesel 
generator is tested in accordance with the design basis 
requirements. The proposed changes are administrative in nature and 
do not adversely affect existing plant safety margins or the 
reliability of the equipment assumed to operate in the safety 
analysis. The proposed changes have been evaluated and found to be 
acceptable for use at Quad Cities based on system design, safety 
analysis requirements and operational performance. Since the 
proposed changes are administrative in nature and maintain necessary 
levels of system or component reliability, the proposed changes do 
not involve a significant reduction in the margin of safety.
    The proposed amendment for Quad Cities Station will not reduce 
the availability of systems 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 24, 1996 and January 31, 1997

[[Page 14461]]

    Description of amendment request: Changes to Administrative 
Controls section of the Technical Specifications needed to implement 
revised management responsibilities and titles that reflect the 
permanently shut down status of plant.
    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:
    In accordance with 10 CFR 50.92, CYAPCO [Connecticut Yankee 
Atomic Power Company] and NNECO [Northeast Nuclear Energy Company] 
have reviewed the attached proposed changes and have concluded that 
they do not involve a Significant Hazard consideration (SHC). The 
basis of this conclusion is that the three criterion of 10 CFR 50.92 
are not compromised. The proposed changes do not involve an SHC 
because the proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No design basis accidents are affected by these proposed 
changes. The proposed changes are administrative in nature and are 
being proposed to reflect the organizational changes which became 
effective December 9, 1996.
    The Haddam Neck unit changes are replacement of the Executive 
Vice President, Nuclear by the Executive Vice President and Chief 
Nuclear Officer along with the replacement of the Vice President, 
Haddam Neck by the Unit Director.
    No safety systems are adversely affected by the proposed 
changes, and no failure modes are associated with the changes.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no changes in any way that the plants are operated due 
to this administrative change. The potential for an unanalyzed 
accident is not created. There is no impact on plant response, and 
no new failure modes are introduced. The proposed administrative and 
editorial changes have no impact on safety limits or design basis 
accidents, and have no potential to create a new or unanalyzed 
event.
    3. Involve a significant reduction in a margin of safety.
    These changes do not directly affect any protective boundaries 
nor do they impact the safety limits for the protective boundaries. 
These proposed changes are administrative and editorial in nature. 
Therefore there can be no reduction in the margin of safety.
    The Commission has provided guidance concerning the application 
of the standards in 10 CFR 50.92 by providing certain examples (51 
FR 7751, March 4, 1986) of amendments that are considered not likely 
to involve an SHC. The changes proposed herein are enveloped by 
example (1), since they are purely administrative changes to the 
technical specifications to reflect organizational title changes and 
to achieve consistence throughout the technical specifications at 
Haddam Neck.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457
    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: Seymour H. Weiss

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

    Date of amendment request: July 17, 1996
    Description of amendment request: The proposed change request 
modifies Waterford Steam Electric Station, Unit 3, Technical 
Specifications 3/4.7.1.3,'' CONDENSATE STORAGE POOL,'' by increasing 
the minimum required contained water volume from 82 percent to 91 
percent indicated level. This proposed change is required to ensure 
that the minimum useable water volume in the Condensate Storage Pool 
(CSP) is maintained greater than or equal to 170,000 gallons. The new 
minimum level accounts for the minimum level required to prevent 
Emergency Feedwater pump suction line vortexing and instrument 
measurement uncertainties.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    Increasing the minimum required CSP level will insure that the 
minimum required 170,000 gallons of water is available for supply to 
the Emergency Feedwater System. Maintaining the minimum required 
water volume will not increase the probability of any accident 
previously evaluated. Additionally, it will not affect the 
consequences of any accident. Maintaining at least 170,000 gallons 
of water available in the CSP will ensure that the system remains 
within the bounds of the accident analysis. Therefore, the proposed 
change will not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    Increasing the minimum water volume of the CSP from 82 percent 
to 91 percent does not create a possibility for a new or different 
kind of accident. The CSP will be operated in the same manner as 
previously evaluated. Therefore, the proposed change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    Operation in accordance with this proposed change will ensure 
that the minimum contained water volume of the CSP will remain at 
least 170,000 gallons under all conditions. This will maintain the 
present margin of safety. Therefore, the proposed change will not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: February 5, 1997
    Description of amendment request: The proposed amendment will 
change Waterford Steam Electric Station, Unit 3, Technical 
Specifications 3.1.2.7, 3.1.2.8, 3.5.1, 3.5.4, 3.9.1, and Bases 3/
4.1.2. The proposed change will increase the minimum boron 
concentration in the Safety Injection Tanks (SITs) and the Refueling 
Water Storage Pool (RWSP) to 2050 ppm to reflect the safety analysis 
for fuel Cycle 9.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

[[Page 14462]]

    The Safety Injection System (SIS) is designed to provide core 
cooling in the unlikely event of a loss of coolant accident (LOCA). 
The cooling must be sufficient to prevent significant alteration of 
core geometry, preclude fuel melting, limit the cladding metal-water 
reaction, and remove the energy generated in the core for an 
extended period of time following a LOCA. The SIS fluid must contain 
the necessary boron concentration to maintain the core subcritical 
for the duration of a LOCA.
    The proposed change increases the minimum boron concentration in 
the SITs and RWSP from 1720 ppm to 2050 ppm. Thus, the SIT/RWSP will 
at all times contain sufficient borated water to provide adequate 
shutdown margin. Sampling of the system and RWSP required by the 
Technical Specifications assures that the required dissolved boron 
concentration is present. In addition to its emergency core cooling 
function, the SIS functions to inject borated water into the RCS to 
increase shutdown margin following a rapid cooldown of the RCS as a 
result of a steam line rupture.
    Operation of the safety injection system is credited in the 
steam line break analysis for causing a decrease in core reactivity. 
The current minimum RWSP/SIT concentration to be injected is 1720 
ppm. Thus an increase to 2050 ppm will have no adverse affect on 
this analysis.
    The Mode 5 boron dilution event identifies that with an initial 
boron concentration of 1240 ppm, a Keff of 0.98, RCS partially 
drained, and one charging pump operational, the minimum possible 
time to criticality is greater than 90 minutes. For all other 
combinations of Keff, RCS conditions, and number of charging pumps, 
the time to loss of shutdown margin is greater than 55 minutes. 
Thus, the proposed increase in boron concentration will not affect 
the results of the Mode 5 boron dilution event.
    The change to the action statement of TS 3.9.1 assures that the 
more limiting reactivity condition of a Keff less than 0.95 or a 
boron concentration of 2050 ppm specified in the COLR [Core 
Operating Limit Report] will be adhered to during refueling 
operations.
    The upper limit on boron concentration has not changed; 
therefore, there will be no affect on boric acid precipitation post-
LOCA.
    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 physically alter the configuration 
of the plant and, therefore, does not create the possibility of a 
new or different kind of accident from any previously evaluated 
accident.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change maintains the minimum of 55 minutes to 
criticality for the refueling mode boron dilution event analysis. 
The proposed change continues to ensure that borated water of 
sufficient concentration is injected from both the SITs and the RWSP 
in the event of a LOCA or MSLB [main steam line break] and that 
boric acid does not precipitate in the core during long term cooling 
following a LOCA.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: February 10, 1997
    Description of amendment request: The proposed amendment would 
provide the requirements for avoidance and protection from thermal 
hydraulic instabilities as described in NRC Generic Letter 94-02, 
``Long-Term Solutions and Upgrade of Interim Operating Recommendations 
for Thermal Hydraulic Instabilities in Boiling Water Reactors.''
    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 amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. In fact, it does not result in an increase in the 
probability or consequences of any previously evaluated accidents. 
The implementation of [Boiling Water Reactor Owners' Group] BWROG 
Long-Term Stability Solution Option I-D at [Cooper Nuclear Station] 
CNS does not modify the assumptions contained in the existing 
accident analysis. The use of an exclusion region and the operator 
actions required to avoid and minimize operation inside the region 
do not increase the possibility of an accident.
    Conditions of operation outside of the exclusion region are 
within the analytical envelope of the existing safety analysis. The 
operator action requirement to exit the exclusion region upon entry 
minimizes the possibility of an oscillation occurring. The actions 
to drive control rods and/or to increase recirculation flow to exit 
the region are maneuvers within the envelope of normal plant 
evolutions. The flow-biased scram has been analyzed and will provide 
automatic fuel protection in the event of an instability. Thus, each 
proposed Technical Specification requirement provides defense for 
protection from an instability event within the existing assumptions 
of the accident analysis.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    As stated above, the proposed Technical Specification 
requirements either mandate operation within the envelope of 
existing plant operating conditions or force specific operating 
maneuvers within those carried out in normal operation. Since 
operation of the plant with all of the proposed requirements is 
within the existing operating basis, an unanalyzed accident will not 
be created through implementation of the proposed change.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Each of the proposed requirements for plant thermal-hydraulic 
stability provides a means for fuel protection. The combination of 
avoiding possible unstable conditions and the automatic flow-biased 
reactor scram provides an in-depth means for fuel protection. 
Therefore, the individual or combination of means to avoid and 
suppress an instability supplements 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 Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305
    Attorney for licensee: Mr. John R. McPhail, 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. 50-245, 
Millstone Nuclear Power Station, Unit No. 1, New London, 
Connecticut

    Date of amendment request: March 6, 1997
    Description of amendment request: During a self assessment, the 
licensee identified weaknesses in the current Technical Specifications 
regarding allowed outage times for certain specific protective 
instrumentation and also for reactor building access control. The 
proposed amendment is designed to eliminate these weaknesses by 
adopting guidance from NUREG-0123, ``Standard Technical Specifications 
for General Electric Boiling Water Reactors (BWR/

[[Page 14463]]

 5),'' Revision 3, and NUREG-1433, Standard Technical Specifications 
General Electric Plants BWR/4,'' Revision 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 operation of Millstone Nuclear Power Station, Unit No. 1, 
in accordance with the proposed amendment, will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The inherent redundancy and reliability of the protective 
instrumentation trip systems ensure that the consequences of an 
accident are not significantly increased. In addition, the 
restrictive Allowable Outage Time (AOT) interval limits the 
probability of the protective instrument channel being unavailable 
and an accident requiring its function from occurring 
simultaneously. The requirement that the associated trip function 
maintains trip capability ensures that the protective 
instrumentation response will occur such that the consequences of an 
accident are not different from those previously evaluated.
    Instruments addressed in the proposed TS respond to changes in 
the plant. The proposed (AOTs) provide a two-hour interval where the 
instrument is inoperable, yet the Technical Specification (TS) 
Limiting Condition for Operation (LCO) action statement is not 
immediately entered. The probability of a plant transient being 
initiated by a trip of a coincident channel during surveillance 
testing is reduced since the channel under test will only be tripped 
for a small portion of the test interval. Therefore, AOTs provided 
by the proposed TS have no effect on the probability of occurrence 
of previously evaluated accidents.
    The proposed TS changes provide a two-hour interval where the 
instrument is inoperable, but the TS LCO action statement is not 
immediately entered. If a single failure occurred on the other 
channel of the trip system being tested and the channel being tested 
was not in the trip condition, a valid signal might not provide the 
required protective action. The probability of an event requiring 
initiation of the protective function within the proposed AOT is 
low. Additionally, surveillance testing is not generally performed 
on multiple sensors simultaneously. So, other trip functions and 
sensors remain operable and the probability of extensive 
inoperabilities affecting diverse trip functions is low. A spurious 
trip of a coincident channel could initiate a plant transient (for 
example, a reactor scram or a main steam isolation valve closure); 
however, these transients are bounded by the current analyses. 
Moreover, the original TS bases submitted as part of the application 
for Millstone Unit No. 1's Provisional Operating License (dated 
October 7, 1970) included recognition that instruments would be 
inoperable during required functional tests and calibrations. Thus, 
these conditions were recognized in the original design bases and 
constitute part of the licensing bases of the plant. NUREG-0123 
provided specific time frames[,] ...AOTs addressed in the table 
notes[,] and specific action statements. Millstone Unit No. 1 AOT 
values chosen are consistent with these values and less than those 
approved in NUREG-1433 which had a more detailed study performed to 
lengthen the AOT value.
    The existing TS definition for Instrument Functional Test would 
be difficult to satisfy if the LCO condition of tripping the 
inoperable channel was performed. A similar problem of complying 
with the Instrument Calibration definition also exists. The TS 
requirement to perform functional tests and calibrations is not 
consistent with a requirement to trip the system under test. The 
proposed TS changes permit more complete functional and calibration 
testing. For example, the main scram contactors could be included 
within the surveillance tests. Therefore, these TS clarifications do 
not increase the consequences of any previously analyzed accidents.
    The two-hour instrumentation AOT for the Air Ejector Off-Gas 
System radiation monitors is slightly less restrictive than that 
allowed by the NUREG-0123. Since this requirement was relocated from 
NUREG-1433, there is no corresponding requirement for comparison. 
These radiation monitors are arranged in a two-out-of-two logic; 
therefore, both must trip to initiate the required action (closure 
of the off-gas isolation valve to the main stack). This action, 
however, is automatically delayed by 15 minutes. A high radiation 
condition sensed by the monitor in service would provide sufficient 
time to take corrective actions. Since a two-hour AOT is deemed 
acceptable for instrumentation in system[s] such as the Reactor 
Protection System and Emergency Core Cooling Systems, it is 
appropriate to apply a two-hour AOT to these radiation monitors. 
Additionally, the NUREG-0123 AOT of one hour does not allow 
sufficient time to perform required surveillance testing without 
placing undue stress on the test performer. The probability of a 
plant transient (e.g., loss of condenser vacuum) resulting from a 
trip of the coincident channel during surveillance testing is 
reduced since the channel under test will only be tripped for a 
small portion of the test interval. This transient is bounded by 
existing analyses. Therefore, this proposed AOT will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    Since no physical change is being made to the secondary 
containment, or to any systems or components that interface with the 
secondary containment, there is no change in the probability of any 
accident analyzed in the UFSAR [Updated Final Safety Analysis 
Report].
    The proposed change continues to ensure the secondary 
containment requirements meet the licensing basis. Also, the 
proposed changes are based on Standard Technical Specifications, 
NUREG-1433, ``Standard Technical Specifications General Electric 
Plants, BWR/4,'' Revision 1 guidelines and implement actions to be 
taken when secondary containment integrity is not met. If secondary 
containment integrity is not met, existing TS 3.7.C directs the 
plant to be placed in an operating condition where secondary 
containment is not required, e.g., COLD SHUTDOWN. A four hour 
allowable outage time is proposed which provides a period of time to 
correct the problem that is commensurate with the importance of 
maintaining secondary containment during RUN, STARTUP/HOT STANDBY or 
HOT SHUTDOWN. The secondary containment is not an initiator for any 
accident. Therefore, the proposed change will not increase the 
probability of any previously analyzed accident. This short time 
period ensures that the probability of an accident requiring 
secondary containment integrity operability occurring during periods 
when secondary containment integrity is inoperable is minimal.
    The proposed surveillance requirement is based on the NUREG-1433 
surveillance requiring periodic confirmation that at least one door 
in each of the double-doored accesses to the secondary containment 
is closed, provides additional assurance of secondary containment 
system integrity. While this is a deviation from NUREG-1433 (which 
requires that both doors in each access be closed except for normal 
entry and exit), it is consistent with the current definition of 
Secondary Containment Integrity, which requires that at least one 
door in each access opening be closed. Hence, the deviation is 
justifiable and represents increased passive testing which will 
provide increased awareness of plant conditions. Increased awareness 
of plant conditions should reduce the probability or consequences of 
any accident previously evaluated.
    Since the aspects of secondary containment integrity affected by 
reactor building access control are being revised in this proposed 
amendment to agree with the allowable outage time allowed by NUREG-
1433 upon loss of secondary containment integrity, the change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Since the editorial items do not alter the meaning or intent of 
any requirements, they do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of Millstone Nuclear Power Station, Unit No. 1, 
in accordance with the proposed amendment, will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change to the protective instrumentation trip 
system specifications do not create the possibility of a new or 
different kind of accident because they do not introduce any new 
operational modes or physical modifications to the plant.
    Instruments addressed in the proposed TS respond to changes in 
the plant. The proposed AOTs provide a two-hour interval where the 
instrument is inoperable, yet the TS LCO action statement is not 
immediately entered. Given a single failure, this could impact the 
response of the trip channel but not the initiation of the event. 
The only

[[Page 14464]]

action resulting from the AOTs is to perform testing as required by 
TS. Spurious signals during testing could initiate transients but 
would be bounded by the previous transient analyses. These tests do 
not subject the instruments to any conditions beyond their design 
specifications and are performed in accordance with approved testing 
standards. This testing ensures equipment operability by identifying 
degraded conditions, initiating corrective action and properly 
retesting them. Therefore, the proposed TS changes will not 
introduce a new or different kind of accident than previously 
evaluated.
    The two-hour instrumentation AOT for the Air Ejector Off-Gas 
System radiation monitors is slightly less restrictive than that 
allowed by the NUREG-0123. Since this requirement was relocated from 
NUREG-1433, there is no corresponding requirement for comparison. 
These radiation monitors are arranged in a two-out-of-two logic; 
therefore, both must trip to initiate the required action (closure 
of the off-gas isolation valve to the main stack). This action, 
however, is automatically delayed by 15 minutes. A high radiation 
condition sensed by the monitor in service would provide sufficient 
time to take corrective actions. Since a two-hour AOT is deemed 
acceptable for instrumentation in system[s] such as the Reactor 
Protection System and Emergency Core Cooling Systems, it is 
appropriate to apply a two-hour AOT to these radiation monitors.
    The proposed changes to Millstone Unit No. 1 Technical 
Specifications Section 3.7/4.7 and associated bases were developed 
using the guidance provided in the Standard Technical 
Specifications, NUREG-1433, ``Standard Technical Specifications 
General Electric Plants, BWR/4,'' Revision 1. Augmentation of the 
existing surveillance requirements by incorporation of an additional 
NUREG-1433 based surveillance, provides additional assurance of 
secondary containment system integrity. While this is a deviation 
from NUREG-1433 (which requires that both doors in each access be 
closed except for normal entry and exit), it is consistent with the 
current definition of Secondary Containment Integrity which requires 
that at least one door in each access opening be closed. Hence, the 
deviation is justifiable and represents increased passive testing 
which will provide increased awareness of plant conditions. 
Increased awareness of plant conditions will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. Since the proposed changes do not 
significantly degrade the present level of system operability and 
add provisions from NUREG-1433, the proposed amendment does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    Since the editorial items do not alter plant configurations or 
operating modes, they do not create the possibility of a new or 
different kind of accident.
    3. The operation of Millstone Nuclear Power Station, Unit No. 1, 
in accordance with the proposed amendment, will not involve a 
significant reduction in a margin of safety.
    The protective instrumentation surveillance requirements provide 
verification of the operability of the trip system instrumentation 
channels. In addition, the channel that monitors the identical Trip 
Function within the same trip system maintains trip capability for 
the relatively short duration that the coincidence change is in 
effect. This ensures that protective instrumentation reliability is 
maintained. The proposed change provides for a specific time period 
to perform required surveillances on instrument channels without 
trips present in associated trip systems. This time allotment tends 
to enhance the margin of safety by decreasing the probability of 
unnecessary challenges to safety systems and inadvertent plant 
transients.
    The proposed TS provide a two-hour interval where the instrument 
is inoperable, yet the TS LCO action statement is not immediately 
entered. If a single failure occurred on the other channel of the 
trip system being tested and the channel being tested was not in the 
tripped condition, a valid signal might not provide the required 
protective action. The probability of an event requiring initiation 
of the protective function within the proposed AOT is low. 
Additionally, surveillance testing is not generally performed on 
multiple sensors simultaneously. So, other trip functions and 
sensors remain operable and the probability of extensive 
inoperabilities affecting diverse trip functions is low.
    The existing TS definition for Instrument Functional Test would 
be difficult to satisfy if the LCO condition of tripping the 
inoperable channel was performed. A similar problem of complying 
with the Instrument Calibration definition also exists. Moreover, 
the original TS bases submitted as part of the application for 
Millstone Unit No. 1s Provisional Operating License (dated October 
7, 1970) included recognition that instruments would be inoperable 
during required functional test and calibrations. Thus, these 
conditions were recognized in the original design bases and 
constitute part of the licensing bases of the plant. NUREG-0123 
provided specific time frames[,]...AOTs addressed in the table 
notes[,] and specific action statements. Millstone Unit No. 1 AOT 
values chosen are consistent with these values and less than those 
approved in NUREG-1433 which had a more detailed study performed to 
lengthen the AOT value.
    The only action resulting from the proposed TS is to perform 
testing as required by TS. Spurious signals during testing could 
initiate equipment or plant transients but would be bounded by the 
previous transient analysis. These tests do not subject the 
instruments to any conditions beyond their design specifications and 
are performed in accordance with approved testing standards. This 
testing ensures equipment operability by identifying degraded 
conditions, initiating corrective action and properly retesting 
them. Therefore, the proposed TS do not involve a significant 
reduction in a margin of safety.
    The two-hour instrumentation AOT for the Air Ejector Off-Gas 
System radiation monitors is slightly less restrictive than that 
allowed by the NUREG-0123. Since this requirement was relocated from 
NUREG-1433, there is no corresponding requirement for comparison. 
These radiation monitors are arranged in a two-out-of-two logic; 
therefore, both must trip to initiate the required action (closure 
of the off-gas isolation valve to the main stack). This action, 
however, is automatically delayed by 15 minutes. A high radiation 
condition sensed by the monitor in service would provide sufficient 
time to take corrective actions. Since a two-hour AOT is deemed 
acceptable for instrumentation in system[s] such as the Reactor 
Protection System and Emergency Core Cooling Systems, it is 
appropriate to apply a two-hour AOT to these radiation monitors and 
does not involve a significant reduction in the margin of safety.
    The addition of an allowable outage time of four hours for 
Secondary Containment Integrity has negligible effect on accident 
occurrence or consequences. Since the proposed change does not 
involve the addition or modification of plant equipment, is 
consistent with the intent of the existing Technical Specifications, 
is consistent with the current industry practices as outlined in 
NUREG-1433, (except for the deviation noted above), and is 
consistent with the design basis of the plant and the accident 
analysis, no action will occur that will involve a significant 
reduction in a margin of safety.
    Since the editorial items do not alter the meaning or intent of 
any requirements, they do not affect 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, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270. NRC Deputy Director: Phillip F. McKee

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 amendment requests: July 28, 1995, as revised February 21, 
1997
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TSs) to allow use of credit for 
soluble boron in spent fuel pool criticality analyses. The licensee's 
February 21, 1997, submittal is a revision to its original amendment 
requests dated July 28, 1995. The generic methodology for crediting 
soluble boron in spent fuel rack criticality analyses was approved

[[Page 14465]]

by the NRC on October 25, 1996. However, because of changes made to the 
generic methodology as a result of comments from the NRC staff, it was 
necessary for NSP to revise its original amendment requests. In 
addition, the licensee has revised its request by eliminating the 
proposed relocation of the spent fuel pool operating limits to the Unit 
1 core operating limits report and will retain these limits in the TSs.
    The licensee's original application for amendments was published in 
the Federal Register on September 23, 1996, (61 FR 49800).
    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[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    There is no increase in the probability of a fuel assembly drop 
accident in the spent fuel pool when considering the presence of 
soluble boron in the spent fuel pool water for criticality control. 
The handling of the fuel assemblies in the spent fuel pool has 
always been performed in borated water.
    The criticality analysis showed the consequences of a fuel 
assembly drop accident in the spent fuel pool are not affected when 
considering the presence of soluble boron.
    There is no increase in the probability of the accidental 
misloading of spent fuel assemblies into the spent fuel pool racks 
when considering the presence of soluble boron in the pool water for 
criticality control. Fuel assembly placement will continue to be 
controlled pursuant to approved fuel handling procedures and will be 
in accordance with the Technical Specification spent fuel rack 
storage configuration limitations. The addition of the spent fuel 
pool storage configuration surveillance in proposed Specification 
4.20 will provide increased assurance that a spent fuel pool 
inventory verification will be completed in a timely manner after 
completion of a fuel handling campaign in the spent fuel pool.
    There is no increase in the consequences of the accidental 
misloading of spent fuel assemblies into the spent fuel pool racks 
because criticality analyses demonstrate that the pool will remain 
subcritical following an accidental misloading if the pool contains 
an adequate boron concentration. The proposed Technical 
Specifications limitations will ensure that an adequate spent fuel 
pool boron concentration will be maintained.
    There is no increase in the probability of the loss of normal 
cooling to the spent fuel pool water when considering the presence 
of soluble boron in the pool water for subcriticality control since 
a high concentration of soluble boron has always been maintained in 
the spent fuel pool water.
    A loss of normal cooling to the spent fuel pool water causes an 
increase in the temperature of the water passing through the stored 
fuel assemblies. This causes a decrease in water density which would 
result in a decrease in reactivity when Boraflex neutron absorber 
panels are present in the racks. However, since Boraflex is not 
considered to be present, and the spent fuel pool water has a high 
concentration of boron, a density decrease causes a positive 
reactivity addition. However, the additional negative reactivity 
provided by the proposed 1800 ppm boron concentration limit, above 
that provided by the concentration required to maintain K<INF>eff 
less than or equal to 0.95 (750 ppm), will compensate for the 
increased reactivity which could result from a loss of spent fuel 
pool cooling event. Because adequate soluble boron will be 
maintained in the spent fuel pool water, the consequences of a loss 
of normal cooling to the spent fuel pool will not be increased.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    Spent fuel handling accidents are not new or different types of 
accidents, they have been analyzed in Section 14.5.1 of the Updated 
Safety Analysis Report.
    Criticality accidents in the spent fuel pool are not new or 
different types of accidents, they have been analyzed in the Updated 
Safety Analysis Report and in Criticality Analysis reports 
associated with specific licensing amendments for fuel enrichments 
up to 5.0 weight percent U-235.
    The Prairie Island Technical Specifications currently contain 
limitations on the spent fuel pool boron concentration. Current 
Specification 3.8.E.2, which covers the storage of restricted fuel 
assemblies in an unverified condition, and Specification 3.8.B.1.c 
for the loading of fuel assemblies into a cask in the spent fuel 
pool, contain requirements for spent fuel pool boron concentration. 
The actual boron concentration in the spent fuel pool has always 
been kept at a higher value for refueling purposes. New 
Specification 3.8.E.2 establishes new boron concentration 
requirements for the spent fuel pool water consistent with the 
results of the new criticality analysis (Exhibit E [of the February 
21, 1997, submittal]).
    Since soluble boron has always been maintained in the spent fuel 
pool water, and is currently required by Technical Specifications 
under some circumstances, the implementation of this new requirement 
will have little effect on normal pool operations and maintenance. 
The implementation of the proposed new limitations on the spent fuel 
pool boron concentration will only result in increased sampling to 
verify boron concentration. This increased sampling will not create 
the possibility of a new or different kind of accident.
    Because soluble boron has always been present in the spent fuel 
pool and is required by current Technical Specifications as 
discussed above, a dilution of the spent fuel pool soluble boron has 
always been a possibility. However, it was shown in the spent fuel 
pool dilution evaluation (Exhibit D [of the February 21, 1997, 
submittal]) that a dilution of the Prairie Island spent fuel pool 
which could reduce the rack K<INF>eff to less than 0.95 is not a 
credible event. Therefore, the implementation of new limitations on 
the spent fuel pool boron concentration will not result in the 
possibility of a new kind of accident.
    Revised Specifications 3.8.E.1, 5.6.A.1.d and 5.6.A.1.e continue 
to specify the requirements for the spent fuel rack storage 
configurations, the only significant changes relate to the criteria 
for determining the storage configuration. Since the proposed spent 
fuel pool storage configuration limitations will be similar to those 
currently in the Prairie Island Technical Specifications, the new 
limitations will not have any significant effect on normal spent 
fuel pool operations and maintenance and will not create any 
possibility of a new or different kind of accident. Verifications 
will continue to be performed to ensure that the spent fuel pool 
loading configuration meets specified requirements.
    As discussed above, the proposed changes will not create the 
possibility of a new or different kind of accident. There is no 
significant change in plant configuration, equipment design or 
equipment. The accident analysis in the Updated Safety Analysis 
Report remains bounding.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The Technical Specification changes proposed by this License 
Amendment Request and the resulting spent fuel storage operating 
limits will provide adequate safety margin to ensure that the stored 
fuel assembly array will always remain subcritical. Those limits are 
based on a plant specific criticality analysis (Exhibit E) performed 
in accordance [with] the Westinghouse spent fuel rack criticality 
analysis methodology described in Reference 4 [in Exhibit A of the 
February 21, 1997, submittal].
    While the criticality analysis utilized credit for soluble 
boron, a storage configuration has been defined using a 95/95 
K<INF>eff calculation to ensure that the spent fuel rack K<INF>eff 
will be less than 1.0 with no soluble boron. Soluble boron credit is 
used to offset uncertainties, tolerances and off-normal conditions 
and to provide subcritical margin such that the spent fuel pool 
K<INF>eff is maintained less than or equal to 0.95.
    The loss of substantial amounts of soluble boron from the spent 
fuel pool which could lead to exceeding a K<INF>eff of 0.95 has been 
evaluated (Exhibit D) and shown to be not credible.
    The evaluations in Exhibit D, which show that the dilution of 
the spent fuel pool boron concentration from 1800 ppm to 750 ppm is 
not credible, combined with the 95/95 calculation, which shows that 
the spent fuel rack K<INF>eff will remain less than 1.0 when flooded 
with unborated water, provide a level of safety comparable to the 
conservative criticality analysis methodology required by References 
1, 2 and 3 [in Exhibit A of the February 21, 1997, submittal].

[[Page 14466]]

    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the plant's margin of 
safety.
    Based on the evaluation above, and pursuant to 10 CFR 50, 
Section 50.91, Northern States Power Company has determined that 
operation of the Prairie Island Nuclear Generating Plant in 
accordance with the proposed license amendment request does not 
involve any significant hazards considerations as defined by NRC 
regulations in 10 CFR 50, Section 50.92.
    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:  Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of amendment requests: February 14, 1997
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant (DCPP) Unit Nos. 1 and 2 to revise the surveillance 
frequencies from at least once every 18 months to at least once per 
refueling interval (nominally 24 months) for 8 slave relay tests, 20 
electrical system tests and 1 electrical TS Bases change, and 5 
miscellaneous tests.
    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 TS surveillance interval increase to 24 months do 
not alter the intent or method by which the inspections, tests, or 
verifications are conducted; do not alter the way any structure, 
system, or component functions; and do not change the manner in 
which the plant is operated.
    The surveillance, maintenance, and operating histories indicate 
that the equipment will continue to perform satisfactorily with 
longer surveillance intervals. Few surveillance and maintenance 
problems were identified. No problems have recurred following 
identification of root causes and implementation of corrective 
actions.
    There are no known mechanisms that would significantly degrade 
the performance of the evaluated equipment during normal plant 
operation. All potential time related degradation mechanisms have 
insignificant effects in the timeframe of interest (24 months +25 
percent, or 30 months). Based on the past performance of the 
equipment, the probability or consequences of accidents would not be 
significantly affected by the proposed surveillance interval 
increases.
    Deletion of the phrase ``during shutdown'' for the applicable 
electrical TS will not alter the intent or method by which the 
inspections, tests, or verifications are conducted; nor alter the 
way any structure, system, or component functions. DCPP has 
administrative programs in place which require evaluation of risk 
and suitability of surveillance and maintenance activities to ensure 
that performance during plant operation does not adversely affect 
safety.
    The administrative change for one PORV TS regarding channel 
calibration only maintains the existing surveillance frequency. This 
revision does not alter the intent or method by which the 
inspections, tests, or verifications are conducted; nor alter the 
way any structure, system, or component functions.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    For the proposed TS changes involving surveillance interval 
increases to 24 months, the surveillance and maintenance histories 
indicate that the equipment will continue to effectively perform its 
design function over the longer operating cycles. Additionally, the 
increased surveillance intervals do not result in any physical 
modifications, affect safety function performance or the manner in 
which the plant is operated, or alter the intent or method by which 
surveillance tests are performed. No problems have recurred 
following identification of root causes and implementation of 
corrective actions. All identified potential time related 
degradations have insignificant effects in the timeframe of 
interest. The proposed surveillance interval increases would not 
affect the type of accident possible.
    Deletion of the phrase during shutdown for the applicable 
electrical TS does not result in any physical modifications, affect 
safety function performance or the manner in which the plant is 
operated, or alter the intent or method by which surveillance tests 
are performed. DCPP has administrative programs in place which 
require evaluation of risk and suitability of surveillance and 
maintenance activities to ensure that performance during plant 
operation does not adversely affect safety.
    The administrative change for one PORV TS regarding channel 
calibration only maintains the existing surveillance frequency. This 
revision does not result in any physical modifications, affect 
safety performance or the manner in which the plant is operated, or 
alter the intent or method by which surveillance tests are 
performed.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    For the proposed TS changes involving surveillance interval 
increases to 24 months, evaluation of historical surveillance and 
maintenance data indicates there have been few problems experienced 
with the evaluated equipment. There are no indications that 
potential problems would be cycle ength dependent or that potential 
degradation would be significant for the timeframe of interest; 
therefore, increasing the surveillance interval will have little, if 
any, impact on safety. There is no safety analysis impact since 
these changes will have no effect on any safety limit, protection 
system setpoint, or limiting condition for operation, and there are 
no hardware changes that would impact existing safety analysis 
acceptance criteria. Safety margins would not be significantly 
affected by the proposed surveillance interval increases.
    Deletion of the phrase ``during shutdown'' for the applicable 
electrical TS has no safety analysis impact since these changes will 
have no effect on any safety limit, protection system setpoint, or 
limiting condition for operation, and there are no hardware changes 
that would impact existing safety analysis acceptance criteria. DCPP 
has administrative programs in place which require evaluation of 
risk and suitability of surveillance and maintenance activities to 
ensure that performance during plant operation does not adversely 
affect safety.
    The administrative change for one PORV TS regarding channel 
calibration only maintains the existing surveillance frequency. This 
revision has no safety analysis impact.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120

[[Page 14467]]

    NRC Project Director: William H. Bateman

Portland General Electric Company, et al., Docket No. 50-344, 
Trojan Nuclear Plant, Columbia County, Oregon

    Date of amendment request: January 16, 1997, as supplemented 
February 24, 1997.
    Description of amendment request: The proposed amendment would 
allow pre-operational testing and load handling of spent fuel transfer 
and storage casks in the Trojan Fuel Building.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    The proposed changes would not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
With the permanent cessation of operations at the Trojan Plant, the 
number of potential accidents was reduced to those types of 
accidents associated with the storage of irradiated fuel and 
radioactive waste storage and handling. Additional events were 
postulated for decommissioning activities due to the difference in 
the types of activities that were to be performed. The postulated 
accidents described in the Defueled Safety Analysis Report (DSAR) 
are generally classified as: 1) radioactive release from a subsystem 
or component, 2) fuel handling accident, and 3) loss of spent fuel 
decay heat removal capability. The postulated events described in 
the Decommissioning Plan are grouped as: 1) decontamination, 
dismantlement, and materials handling events, 2) loss of support 
systems (offsite power, cooling water, and compressed air), 3) fire 
and explosions, and 4) external events (earthquake, external 
flooding, tornadoes, extreme winds, volcanoes, lightning, toxic 
chemical release). These types of accidents are discussed below.
    Radioactive release from a subsystem or component involves failure 
of a radioactive waste gas decay tank (WGDT) or failure of a chemical 
and volume control system holdup tank (HUT). For a failure of a WGDT, 
the radioactive contents are assumed to be principally noble gases 
krypton and xenon, the particulate daughters of some of the krypton and 
xenon isotopes, and trace quantities of halogens. For the failure of a 
HUT, the assumptions were full power operation with 1-percent failed 
fuel, 40 weeks elapsed since power operation, and 60,000 gallons of 
120 deg.F liquid released over a 2-hour period. However, the WGDTs and 
HUTs are no longer active and have been drained. Therefore, pre-
operational testing and load handling activities cannot increase the 
probability of occurrence of a failure of a WGDT or HUT. Since the 
failure of a WGDT or HUT is no longer credible, the consequences of 
failure of a WGDT or HUT cannot significantly increase as a result of 
pre-operational testing and load handling.
    The fuel handling accident involves a stuck or dropped fuel 
assembly that results in damage of the cladding of the fuel rods in one 
assembly and the release of gaseous fission products. Pre-operational 
testing and load handling do not involve the movement of irradiated 
fuel. A dummy assembly will be used for fit-up testing. The fuel 
handling equipment will be the same as previously analyzed with the 
exception of special tools that may be used to manipulate the dummy 
fuel assembly. These special tools will be similar in size and weight 
to other tools used for underwater manipulation, and therefore, would 
not present a new hazard. In addition, the same administrative controls 
and physical limitations imposed on any fuel handling operation will be 
used for pre-operational testing and load handling. Thus, there is no 
increase in the probability of occurrence of a fuel handling accident 
over what would be expected for any routine fuel handling operation. If 
a dummy fuel assembly were dropped in the spent fuel pool, then only 
one fuel assembly could be damaged. Therefore, the consequences of a 
dummy fuel assembly drop would be the same as the consequences of the 
analysis described in the DSAR. Therefore, the consequences of a dummy 
fuel assembly drop are not significantly increased as a result of pre-
operational testing and load handling.
    The loss of spent fuel decay heat removal capability involves the 
loss of forced spent fuel cooling with and without concurrent spent 
fuel pool (SFP) inventory loss. The only requirement to assume adequate 
decay heat removal capability for the spent fuel is to maintain the 
water level in the SFP so that the spent fuel assemblies remain covered 
(i.e., the capability to makeup water to the SFP must be available when 
required). The potential events that could result in a loss of spent 
fuel decay heat removal capability include external events (explosions, 
toxic chemicals, fires, ship collision with the intake structure, oil 
or corrosive liquid spills in the river, cooling tower collapse, 
seismic events, severe meteorological events), and internal events, 
including SFP makeup water system malfunctions. Pre-operational testing 
and load handling will not require the use of explosive materials, 
toxic chemicals, or flammable materials. The probability of other 
external events (e.g., cooling tower collapse) would be unaffected by 
the pre-operational testing and load handling activities inside the 
fuel building. Pre-operational testing and load handling activities 
will not directly interface with the SFP makeup water systems, and 
therefore could not affect their probability of failure. The safe load 
path and handling height limitations will ensure that a load drop does 
not adversely affect the SFP or makeup water systems. Therefore, there 
is no significant increase in the probability of a loss of spent fuel 
decay heat removal capability. There are no credible adverse 
consequences of the loss of spent fuel decay heat removal as the DSAR 
demonstrates that adequate time is available to establish a source of 
makeup water to the SFP such that uncovering the fuel and an actual 
loss of spent fuel cooling is not credible. The postulated events that 
could affect the SFP (liner tear/breach and heavy load drop) do not 
have a significant adverse effect. In addition, establishment of the 
makeup water path and recovery of spent fuel cooling would not be 
affected because postulated off-normal events and accidents could not 
affect the capability to provide makeup water to the SFP by various 
water sources. Therefore, pre-operational testing and load handling 
cannot significantly increase the consequences of the loss of spent 
fuel decay heat removal.
    The events postulated in the Decommissioning Plan are similar to 
the DSAR with the exception of decontamination, dismantlement, and 
materials handling events. Decontamination events involve gross liquid 
leakage from in-situ decontamination equipment or accidental spraying 
of liquids containing concentrated contamination. Dismantlement events 
include segmentation of components and structures, or removal of 
concrete by rock splitting, explosives, or electric and/or pneumatic 
hammers. Dismantlement events potentially result in airborne 
contamination. Materials handling events involve dropping contaminated 
components, concrete rubble, or filters or packages of particulate 
materials. Pre-operational testing and load handling activities are 
material handling activities and are therefore, within the bounds of 
the existing analysis. Therefore, the probability and consequences of 
decontamination, dismantlement, and materials handling events would not 
be significantly increased.

[[Page 14468]]

    Based on the above, the pre-operational testing and load handling 
activities do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes would not create the possibility of a new or 
different kind of accident from any accident previously evaluated. As 
described in the licensee's safety evaluation of the proposed pre-
operational testing and load handling activities, no types of off-
normal events/accidents were determined to have radiological 
consequences greater than currently evaluated in the DSAR and 
Decommissioning Plan.
    The postulated dummy fuel assembly drop is considered the same type 
or kind of event as the previously analyzed fuel handling accident, 
mainly because the initiator for this postulated event is the same 
(i.e., a (non-specified) failure of the fuel handling equipment or the 
fuel handling bridge crane. During pre-operational testing and load 
handling, a dummy fuel assembly could be dropped in the SFP or the cask 
loading pit. As the cask loading pit is similar in construction to the 
SFP and the cask loading pit will be flooded with borated water of the 
same concentration as the SFP, the differences between the two events 
are negligible and the two events may be considered the same type or 
kind of accident. Therefore the dummy fuel assembly drop is not a new 
or different type or kind of accident.
    The postulated transfer cask drop or mishandling event is similar 
to a materials handling event. Therefore, the consequences of a 
transfer cask drop or mishandling event would not represent a new or 
different type or kind of accident.
    Based on the above, the pre-operational testing and load handling 
activities do not create the possibility of a new or different kind of 
accident.
    The proposed changes do not involve a significant reduction in the 
margin of safety. The Trojan Permanently Defueled Technical 
Specifications (PDTS) contain four limiting conditions of operation 
that address SFP water level, SFP boron concentration, SFP temperature, 
and SFP load restrictions. These PDTS will remain in effect as long as 
spent fuel is stored in the SFP, which is in accordance with their 
applicability statements. The pre-operational testing and load handling 
activities will not affect these PDTS or their bases.
    The cask loading pit (CLP) is immediately adjacent to the SFP. The 
gate between the CLP and the SFP may be opened to allow a dummy fuel 
assembly to moved from the spent fuel storage racks in the SFP to the 
basket in the CLP. Opening the gate will allow free exchange of water 
between the CLP and the SFP. The water in the CLP must be at 
essentially the same level, boron concentration, and temperature as the 
SFP prior to the first opening of the gate to ensure that the limited 
conditions of operation are continuously satisfied for the SFP. 
Therefore, the CLP will be initially filled to about the same level as 
the SFP with water that is about the same boron concentration and 
temperature as the SFP. With these precautions, the limiting conditions 
of operation for SFP level, boron concentration, and temperature will 
be continuously maintained and the margin of safety will be unaffected.
    Pre-operational testing and load handling activities will involve 
lifting and moving heavy loads (e.g., transfer casks). Loads that will 
be carried over fuel in the SFP racks and the heights at which they may 
be carried will be limited in accordance with LCO 3.1.4, ``Spent Fuel 
Pool Load Restrictions,'' in such a way as to preclude impact energies 
over 240,000 in-lbs. With this precaution, the limiting condition of 
operation pertaining to load restrictions over the SFP will be 
satisfied for fuel stored in the SFP racks and the margin of safety 
will be unaffected. The safe load path for heavy loads being lifted and 
moved outside the SFP will be located sufficiently far from the SFP as 
to not have an adverse effect on the SFP in the unlikely event of a 
load drop. In addition, the mechanical stops and electrical interlocks 
on the fuel building overhead crane will provide additional assurance 
that heavy loads are not carried over the fuel in the SFP racks.
    Based on the above, the pre-operational testing and load handling 
activities will not reduce the margin of safety.
    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:  Branford Price Millar 
Library, Portland State University, 934 S.W. Harrison Street, P.O. Box 
1151, Portland, Oregon 97207
    Basis for proposed no significant hazards consideration 
determination:
    Attorney for licensees: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204
    NRR Project Director: Seymour H. Weiss

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: January 2, 1997
    Description of amendment request: The proposed amendment would 
allow a change to the current functional testing frequency for 
Inservice Inspection of American Society of Mechanical Engineers Code 
Class 1, 2, and 3 pumps and valves from the current monthly to a 
quarterly testing frequency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response: Operation of Indian Point 3 in accordance with the 
proposed license does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes involve no hardware changes, no changes to 
the operation of any systems or components, and no changes to 
existing structures. 10 CFR 50.55a(g) requires that safety related 
components (e.g. - pumps and valves) be tested according to the 
requirements of Section XI of the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code (Code) and 
applicable addenda. The revision of functional test frequencies for 
pumps and valves, which are categorized as Code Class 1, 2, or 3, 
from a monthly to a quarterly test interval is consistent with NRC 
guidance provided in NUREG-1366 and in accordance with recommended 
test intervals in the ASME Code. These changes will reduce component 
degradation resulting from unnecessary tests and provide better 
system availability from not having to remove a system/component 
from operability while performing a surveillance. Such changes will 
not alter the probability or consequences of any previously analyzed 
accidents.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: The proposed change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are procedural in nature concerning the 
functional testing frequencies of pumps and valves that have 
historically shown a high percentage of successfully meeting 
surveillance requirements. The methodology of testing these pumps 
and valves will remain unchanged. The proposed changes, while 
slightly increasing the possibility of an

[[Page 14469]]

undetected pump or valve defect, will not create a new or 
unevaluated accident or operating condition.
    (3) Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    Response: The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed changes are in accordance with recommendations 
provided by the NRC regarding the improvement of Technical 
Specifications. These changes will result in the perpetuation of 
current safety margins while reducing the testing burden and 
decreasing equipment degradation.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Acting Director

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

    Date of amendment request: February 11, 1997
    Description of amendment request: The proposed change to Hope Creek 
Technical Specification (TS) Sections 3/4.8.1 ``A.C. Sources,'' 6.8 
``Procedures and Programs,'' and the Bases for Section 3/4.8, 
``Electrical Power Systems,'' would include: 1) the relocation of 
existing surveillance requirements related to diesel fuel oil 
chemistry; 2) the introduction of a new program under TS 6.8.4.e, 
``Diesel Fuel Oil Testing Program;'' 3) revisions to the TS Bases for 
Section 3/4.8 to incorporate information associated with the TS 
changes; and 4) editorial changes to implement required corrections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes involve: 1) no hardware changes; 2) no 
significant changes to the operation of any systems or components in 
normal or accident operating conditions; and 3) no changes to 
existing structures, systems or components. Therefore these changes 
will not increase the probability of an accident previously 
evaluated.
    Establishment of [Emergency Diesel Generator] EDG fuel oil 
testing requirements in TS 6.8.4.e is a change that is consistent 
with changes made in the improved STS [Standard Technical 
Specifications] as contained in Specification 5.5.10 of that 
document. These changes establish a new requirement to test for 
particulates in the EDG fuel oil, but establish a 92 day test 
frequency (as opposed to 31 days in the improved STS) and a 3.0 
micron acceptance criteria (as opposed to 0.8 micron in the improved 
STS) for particulate testing. [Public Service Electric and Gas 
Company] PSE&G concludes that these changes are acceptable based 
upon past EDG fuel oil tests for particulates and acceptable 
performance of the EDG with 5.0 micron filters. In addition, PSE&G 
will utilize more objective test criteria for water and sediment in 
the EDG fuel oil than established by the ``clear and bright'' 
acceptance criteria contained in the improved STS.
    Since the EDG fuel oil will still: 1) meet all of the 
requirements established for fuel oil specified in the improved STS; 
and 2) retain the capability to mitigate the consequences of 
accidents described in the [Hope Creek Generating Station] HC Safety 
Analysis Report, the proposed changes were determined to be 
justified. Based on established fuel oil quality history, the 
proposed testing methods and frequencies will not significantly 
decrease confidence in fuel oil quality and EDG operability, nor 
will they have any negative effect on established plant practices in 
regards to the testing of EDG fuel oil. Therefore, these changes 
will not involve a significant increase in the consequences of an 
accident previously evaluated.
    The revisions proposed to the TS Bases are being made to provide 
additional information supporting the proposed EDG TS. With the 
approval of the proposed TS changes, the associated Bases changes 
would be editorial in nature. Therefore, these changes will not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    In addition, the proposed change to [Limiting Condition for 
Operation] LCO 3.8.1.1, ACTION c., is considered to be editorial in 
nature and will not result in a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The HC EDGs are designed to mitigate the consequences of 
accidents by providing electrical power to safety-related equipment. 
Failure of the EDGs are not considered to initiate any of the 
accidents described in the HC Safety Analysis Report. The proposed 
changes concern fuel oil system surveillances and testing frequency. 
The proposed changes will not adversely impact the operation of any 
safety related component or equipment. Since the proposed changes 
involve: 1) no hardware changes; 2) no significant changes to the 
operation of any systems or components; and 3) no changes to 
existing structures, systems or components, there can be no impact 
on the occurrence of any accident. Furthermore, there is no change 
in plant testing proposed in this change request which could 
initiate an event. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    In addition, the proposed change to LCO 3.8.1.1, ACTION c., is 
considered to be editorial in nature and will not result in 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.
    Establishment of EDG fuel oil testing requirements in TS 6.8.4.e 
is a change that is consistent with changes made in the improved 
STS. The proposed changes address: 1) how EDG fuel oil quality is to 
be determined; 2) how frequently this determination is to be 
performed; and 3) how to control the process for determining fuel 
oil acceptability and resultant EDG operability. With the exception 
of particulate testing (which is being added) all acceptance 
criteria for fuel oil testing remain unchanged. Based on historical 
data, EDG fuel oil quality will not be adversely affected or 
impacted by the proposed changes. Therefore, the proposed amendment 
does not involve any significant reduction in a safety margin.
    The revisions proposed to the TS Bases are being made to provide 
additional information supporting the proposed EDG TS. With the 
approval of the proposed TS changes, the associated Bases changes 
would be editorial in nature. Therefore, these changes will not 
involve a significant reduction in a safety margin.
    In addition, the proposed change to LCO 3.8.1.1, ACTION c., is 
considered to be editorial in nature and will not involve a 
significant reduction in a safety margin.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of amendment request: October 23, 1996, January 31, February 
10 and 24 and March 11, 1997.
    Description of amendment request: The proposed amendment would 
revise

[[Page 14470]]

the Watts Bar Nuclear Plant (WBN) Unit 1 Technical Specifications to 
increase the enrichment and storage capacity of the spent fuel pool 
racks. The proposed modification increases the (Watts Bar Nuclear 
Plant) WBN spent fuel storage capacity from 484 fuel assemblies to 1835 
fuel assemblies. The initial enrichment of the fuel to be stored in the 
spent fuel storage racks will be increased from 3.5 weight percent 
(wt%) to 5.0 wt%. This modification would also change the spacing of 
stored fuel assembly center-to-center spacing from a nominal 10.72 
inches to 10.375 inches in 24 PaR flux trap rack modules and 8.972 
inches in ten smaller burnup credit rack modules to be installed 
peripherally along the south and west pool walls and in a single 15 x 
15 burnup credit rack to be installed in the cask pit.
    In addition to the above proposed revisions, two limiting 
conditions for operation will be added to require that the combination 
of initial enrichment and burnup of each spent fuel assembly to be 
stored is in the acceptable region and to require boron concentration 
of the cask pit to be greater than or equal to 2000 parts per million 
(ppm) during fuel movement in the flooded cask pit. As an added 
protection to the fuel stored in the cask pit area, the Technical 
Requirements Manual (TRM) is being revised to require that an impact 
shield be in place over the fuel when heavy loads are moved near or 
across the cask pit area.
    The WBN Unit 1 Technical Specification Bases and the TRM would be 
revised to support these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Nuclear Regulatory Commission has provided standards for 
determining whether a significant hazards consideration exists (10 
CFR 50.92(c)). A proposed amendment to an operating license for a 
facility involves no significant hazards consideration if 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. Each standard is discussed below for the proposed 
amendment.
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The following potential scenarios were considered:
    1. A spent fuel assembly drop.
    2. Drop of the transfer canal gate or the cask pit divider gate.
    3. A seismic event.
    4. Loss-of-cooling flow in the spent fuel pool.
    5. Installation activities.
    The effect of additional spent fuel pool storage cells fully 
loaded with fuel on the first four potential accident scenarios 
listed above has been considered. It was concluded that after 
installation activities have been completed, the presence of 
additional fuel in the pool does not increase the probability of 
occurrence of these four events. Also, based on evaluations of bulk 
pool temperature, rack seismic responses, and refueling accidents, 
it is reasonable to conclude that there is no significant increase 
in the consequences of these events after installation is complete 
(See Reference 1). During the installation activities, the following 
considerations support a conclusion that neither the probability or 
consequences of these four scenarios would be significantly 
increased.
    A spent fuel assembly cannot be dropped during installation of 
the 24 Programmed and Remote System Corporation (PaR) flux trap rack 
modules because this activity will take place before the end of 
operating cycle one and there will be no spent fuel in the WBN pool 
to be moved or shuffled. Before installing the ten smaller burnup 
credit racks in the pool, some fuel will be moved to create a three 
foot lateral free zone clearance from stored fuel. This would 
involve a one-time movement of an estimated maximum of 225 fuel 
assemblies, which is less that half the fuel movements during one 
refueling outage. This does not significantly increase the 
probability of dropping a fuel assembly, particularly when the many 
administrative controls and physical limitations imposed on fuel 
handling operations are considered. The fuel handling system 
consists of equipment and structures utilized for safely 
implementing refueling operations in accordance with requirements of 
General Design Criteria 61 and 62 of 10 CFR 50, Appendix A. The 
radiological dose consequences of dropping a 5.0 wt% fuel assembly 
are different from the previous FSAR [Final Safety Analysis Report] 
evaluation for the 3.5 wt% fuel assembly. The Beta and Gamma doses 
decrease and the maximum thyroid dose increase is less than 9%. 
Therefore, the change in calculated dose values is insignificant and 
remains well within regulatory guidelines.
    It may be necessary to move the transfer canal gate and the cask 
pit divider gate between their gated and stored positions during 
installation of the burnup credit ``baby'' rack modules along the 
south and west walls. During rack installation, the previously 
mentioned three foot lateral free zone clearance to stored fuel 
would exist. Therefore, no heavy load would be carried directly over 
irradiated fuel during installation of the racks. There are numerous 
design features which comply with NUREG-0612 to preclude these gates 
from dropping on spent fuel. These features include design of the 
lifting devices, design of the crane, and use of written procedures. 
Also, the evaluation results for a gate drop on the racks indicates 
that permanent damage to a fuel storage cell is limited to a maximum 
depth of less than six inches below the top of the rack with no 
effect on the subcriticality of fuel stored in adjacent cells. Based 
on the foregoing, it is reasonable to conclude that gate handling 
during the installation of the ``baby'' racks would not involve a 
significant increase in the probability or consequences of an 
accident.
    The probability of a seismic event is not related to 
installation activities. The worst consequence resulting from a 
seismic event during installation activities would occur during 
handling of a rack. The consequences would be insignificant because 
the Auxiliary Building crane is seismically qualified and both 
handling equipment and operations meet the criteria of NUREG-0612. 
Nevertheless, if the seismic event resulted in a rack drop, the 
consequences are insignificant, i.e., localized damage to the pool 
liner and a minor leak rate which would be small in comparison to 
available installed makeup capacity. The cooling and shielding of 
the spent fuel would remain unaffected. Also the racks being moved 
are empty during installation and therefore, the criticality 
consequences of seismic events are bounded by evaluations for loaded 
racks.
    Rack installation activities cannot cause an accidental loss-of-
cooling flow in the spent fuel pool. The vital components of the 
spent fuel pool cooling and cleanup system (SFPCCS) are not located 
proximate to the pool installation activities. Coolant flow may be 
deliberately curtailed to facilitate installation of the ``baby'' 
racks directly beneath the discharge piping in the southwest corner 
of the pool. The effects of such an action would be readily 
minimized and made inconsequential during the detailed installation 
planning phase by selecting a time when decay heat input from stored 
fuel is relatively constant. Also careful preplanning of the work 
would minimize out-of-service time and provide for intermittent 
coolant flow restart, if necessary, to maintain acceptable bulk 
coolant temperatures. Similarly, the effect of an independently 
initiated loss-of-coolant flow incident on reracking activities can 
be easily accommodated by stopping work, as necessary, to mitigate 
any adverse effects on the installation process. The consequences of 
loss-of-cooling flow in the spent fuel pool during installation are 
bounded by the analysis in Chapter 5 of the report which includes 
the situation in which ``baby'' racks and the 15 x 15 cask pit rack 
are installed, and the pool is filled to capacity with spent fuel.
    With regard to the actual installation activities, the existing 
WBN TRM prohibits loads in excess of 2059 pounds from travel over 
fuel assemblies in the storage pool and requires the associated 
crane interlocks and physical stops be periodically demonstrated 
operable. During installation, racks and associated handling tools 
will be moved over the spent fuel pool, however there will be no 
fuel in the pool when the 24 flux trap rack modules are installed. A 
three foot lateral free zone clearance from stored spent fuel

[[Page 14471]]

will be maintained during installation of the ten smaller burnup 
credit rack modules. Installation work in the spent fuel pit area 
will be controlled and performed in strict accordance with specific 
written instructions.
    NUREG-0612 states that in lieu of providing a single failure-
proof crane system, the control-of-heavy-loads guidelines can be 
satisfied by establishing that the potential for a heavy load drop 
is extremely small. Storage rack movements to be accomplished with 
the WBN Auxiliary Building crane will conform with NUREG-0612 
guidelines in that the probability of a drop of a storage rack is 
extremely small. The crane has a tested capacity of 125 tons. The 
maximum weight of any existing, replacement, or new storage rack and 
its associated handling tool is less than 20 tons. Therefore, there 
is ample safety factor margin for movements of the storage racks by 
the Auxiliary Building crane. Special lifting devices, which have 
redundancy or a rated capacity sufficient to maintain adequate 
safety factors, will also be utilized in the movements of the 
storage racks. In accordance with NUREG-0612, Appendix B, the safety 
margin ensures that the probability of a load drop is extremely low.
    Future load travel over fuel stored in a rack specifically 
designed for the cask loading area of the cask pit will be 
prohibited unless an impact shield, which has been specifically 
designed for this purpose, is covering the area. Loads that are 
permitted when the shield is in place must meet analytically 
determined weight, travel height, and cross-sectional area criteria 
that preclude penetration of the shield. A Technical Requirement 
(TR) has been proposed that incorporates the previously mentioned 
load criteria.
    Also a rack change-out sequence is being developed that 
addresses removal of the existing racks, movement of the new racks 
into the Auxiliary Building, initial staging on the refueling floor, 
and final installation in the pool. The change-out sequence 
objectives include establishing lift heights, travel distances, and 
number of lifts to be as low as reasonably achievable. Accordingly, 
it is concluded that the proposed installation activities will not 
significantly increase the probability of a load-handling accident. 
The consequences of a load-handling accident are unaffected by the 
proposed installation activities.
    The consequences of a spent fuel assembly drop were evaluated, 
and it was determined that the racks will not be distorted such that 
the racks would not perform their safety function. The criticality 
acceptance criterion, K<INF>eff less than or equal to 0.95, is not 
violated, and the calculated doses are well within 10 CFR Part 100 
guidelines. The radiological consequences of the fuel assembly drop 
accident evaluated for WBN, have changed, however, the changes do 
not involve a significant increase in consequences and are well 
within the 10 CFR 100 requirements.
    A TRM change has been proposed that would permit the transfer-
canal gate and the divider gate for the cask pit to travel over fuel 
assemblies in the spent fuel pool during movement between their 
gated and stored position. Rack damage is restricted to an area 
above the active fuel region, therefore, neither criticality nor 
radiological concerns exist.
    The consequences of a seismic event have been evaluated. The 
replacement racks are designed and fabricated and the new racks will 
be fabricated to meet the requirements of applicable portions of the 
NRC regulatory guides and published standards. Design margins have 
been provided for rack tilting, deflection, and movement such that 
the racks do not impact each other or the spent fuel pool walls in 
the active fuel region during the postulated seismic events. The 
free-standing racks will maintain their integrity during and after a 
seismic event. The fuel assemblies also remain intact and therefore 
no criticality concerns exist.
    The spent fuel pool system is a passive system with the 
exception of the fuel pool cooling train and heating, ventilating, 
and air-conditioning (HVAC) equipment. Redundancies in the cooling 
train and HVAC hardware are not reduced by the planned fuel storage 
modification. The potential increased heat load resulting from any 
additional storage of spent fuel is well within the existing system 
cooling capacity. Therefore, the probability of occurrence or 
malfunction of safety equipment leading to the loss-of-cooling flow 
in the spent fuel pool is not significantly affected. Furthermore, 
the consequences of this type incident are not significantly 
increased from previously evaluated cooling system loss of flow 
malfunctions. Thermal-hydraulic scenarios assume the reracked pool 
is approximately 90% full with spent fuel assemblies. From this 
starting point, the remaining storage capacity is utilized by 
analyzing both normal and unplanned full core off loads using 
conservative assumptions and previously established methods. 
Calculated values include maximum pool water bulk temperature, 
coincident maximum pool water local temperature, the maximum fuel 
cladding temperature, time-to-boil after loss-of-cooling paths, and 
the effect of flow blockage in a storage cell.
    Although the proposed modification increases the pool heat load, 
results from the above analyses yield a maximum bulk temperature 
less than 160 degrees Fahrenheit which is below the bulk boiling 
temperature. Also the maximum local water temperature is below 
nucleate boiling condition values. Associated results from 
corresponding loss-of-cooling evaluations give minimums of 5.3 hours 
before boiling begins and 45 hours before the pool water level drops 
to the minimum required for shielding spent fuel. This is sufficient 
time to begin utilization of available alternate sources of makeup 
cooling water. Also, the effect of the increased thermal loading on 
the pool structure, associated cooling system, and components was 
evaluated and determined to establish an acceptable design basis 
with the new storage configuration. No modifications were necessary 
because of the increased temperature.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously analyzed.
    The proposed modification has been evaluated in accordance with 
the guidance of the NRC position paper entitled, ``OT Position for 
Review and Acceptance of Spent-Fuel Storage and Handling 
Applications'', appropriate NRC regulatory guidelines; appropriate 
NRC standard review plans; and appropriate industry codes and 
standards. Proven analytical technology was used in designing the 
planned fuel storage expansion and will be utilized in the 
installation process. Basic reracking technology has been developed 
and demonstrated in applications for fuel pool capacity increases 
that have already received NRC staff approval.
    Proposed TSs for the spent fuel storage racks use burnup credit 
and fuel assembly administrative placement restrictions for 
criticality control. These restrictions are described in the 
proposed change to the design features section of the TSs by 
reference to the Spent Fuel Pool Modifications report. Additional 
evaluations were required to ensure that the criticality criterion, 
k<INF>eff less than 0.95, is maintained. These include evaluation 
for the abnormal placement of unirradiated (fresh) fuel assemblies 
of 5.0 wt% enrichment into a storage cell location designed for 
lower enrichment or irradiated fuel. Soluble boron, for which credit 
is permitted under these abnormal conditions, ensures that 
reactivity is maintained substantially less than the design 
requirement. For example, if the PaR flux trap racks are 
inadvertently all loaded with fresh assemblies of the maximum 5.0 
wt% fuel instead of observing the 3.8 wt% and 6.75 MWD/KgU controls, 
the worth of the 2000 ppm borated water is sufficient to lower the 
k<INF>eff of the storage racks to 0.83. The existing and proposed 
TSs require boron concentration in the pool and cask pit to be more 
than or equal to 2000 ppm during fuel movement. An analytical 
determination of the reactivity worth of 2000 ppm borated water in 
the spent fuel storage pool predicted the change in k<INF>eff to be 
approximately 17 percent k<INF>eff. Although no credit for soluble 
boron was proposed in the TSs, it was also determined by an 
independent calculation that a minimum concentration of 520 ppm 
soluble boron allows the unrestricted storage of 5.0 wt% enriched 
fuel in the PaR flux trap racks.
    The Holtec-designed peripheral ``baby'' racks and the 15 x 15 
racks in the cask loading area can safely and conservatively store 
fuel of 5 wt% initial enrichment burned to 41 MWD/kgU or lower 
enriched fuel with lower burnup, i.e., fuel of equivalent 
reactivity. Evaluations have confirmed that, for the abnormal 
placement of a fresh fuel assembly of 5.0 wt% in these racks, the 
criticality criterion is maintained with the existing and proposed 
TS requirements of 2000 ppm soluble boron.
    Although these changes required addressing additional aspects of 
a previously analyzed accident, the possibility of a previously 
unanalyzed accident is not created.
    The impact shield design together with its attendant 
administrative controls and NUREG-0612 heavy load lift compliance, 
renders the possibility of a heavy load drop

[[Page 14472]]

on fuel as not credible in accordance with the NUREG-0612 single-
failure-proof criteria. Accordingly, since this particular part of 
the proposed reracking modification is not a change that could 
malfunction by a new single failure, the movement of heavy loads 
over the cask pit does not create the possibility of a new or 
different kind of accident.
    It is therefore concluded that the proposed reracking does not 
create the possibility of a new or different kind of accident from 
any previously analyzed.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The design and technical review process applied to the reracking 
modification included addressing the following areas:
    1.
    Nuclear criticality considerations.
    2. Thermal-hydraulic considerations.
    3. Mechanical, material, and structural considerations.
    The established acceptance criterion for criticality is that the 
neutron multiplication factor shall be less than or equal to 0.95, 
including all uncertainties. The results of the criticality analyses 
for the rack designs demonstrate that this criterion is satisfied. 
The methods used in the criticality analysis conform to the 
applicable portions of NRC guidance and industry codes, standards, 
and specifications. In meeting the acceptance criteria for 
criticality in the spent fuel pool and the cask loading area, such 
that k<INF>eff is always less than 0.95 at a 95/95 percent 
probability tolerance level, the proposed amendment does not involve 
a significant reduction in the margin of safety for nuclear 
criticality.
    Conservative methods and assumptions were used to calculate the 
maximum fuel temperature and the increase in temperature of the 
water in the spent fuel pit area. The thermal-hydraulic evaluation 
used methods previously employed. The proposed storage modification 
will increase the heat load in the spent fuel pool, but the 
evaluation shows that the existing spent fuel cooling system will 
maintain the bulk pool water temperature at or below 160 degrees 
Fahrenheit. Thus it is demonstrated that the worst-case peak value 
of the pool bulk temperature is considerably lower than the bulk 
boiling temperature. Evaluation also shows that maximum local water 
temperatures along the hottest fuel assembly are below the nucleate 
boiling condition value. Thus, there is no significant reduction in 
the margin of safety for thermal hydraulic or spent fuel cooling 
considerations.
    The mechanical, material, and structural design of the spent 
fuel racks is in accordance with applicable portions of NRCs 
position in ``OT Position for Review and Acceptance of Spent-Fuel 
Storage and Handling applications,'' dated April 14, 1978 (as 
modified January 18, 1979), as well as other applicable NRC guidance 
and industry codes. The primary safety function of the spent fuel 
racks is to maintain the fuel assemblies in a safe configuration 
through normal and abnormal loading conditions. Abnormal loadings 
that have been evaluated with acceptable results and discussed 
previously include the effect of an earthquake and the impact 
because of the drop of a fuel assembly. The rack materials used are 
compatible with the fuel assemblies and the environment in the spent 
fuel pool. The structural design for the new racks provides tilting, 
deflection, and movement margins such that the racks do not impact 
each other or the spent fuel pit walls in the active fuel region 
during the postulated seismic events. Also the spent fuel assemblies 
themselves remain intact and no criticality concerns exist. In 
addition, finite element analysis methods were used to evaluate the 
continued structural acceptability of the spent fuel pit. The 
analysis was performed in accordance with ``Building Code 
Requirements for Reinforced Concrete,'' (ACI 318-63,77). Therefore, 
with respect to mechanical, material, and structural considerations, 
there is no significant reduction in a margin of safety.
    Summary
    Based on the above analysis, TVA has determined that operation 
of WBN, in accordance with the proposed amendment, would not: (1) 
involve a significant increase in the probability of 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. Therefore, operations of WBN in accordance with the proposed 
amendments as described do not involve significant hazard 
considerations as defined in 10 CFR 50.92 and that the criteria of 
10 CFR 50.91 have accordingly been met.
    TVA has also reviewed the NRC examples of licensing amendments 
considered not likely to involve significant hazards considerations 
as provided in the final adoption of 10 CFR 50.92 published on page 
7751 of the Federal Register, Volume 51, No. 44, March 6, 1986. 
Example (X) provides four criteria that, if satisfied by a reracking 
request, indicate that it is likely no significant hazards 
considerations are involved. The criteria and how TVAs amendment 
request for WBN complies are indicated below.
    Criterion (1):
    The storage expansion method consists of either replacing 
existing racks with a design that allows closer spacing between 
stored spent fuel assemblies or replacing additional racks of the 
original design on the pool floor if space permits.
    Proposed Amendment:
    The WBN reracking involves replacing the existing racks with a 
design that allows slightly closer spacing between stored fuel 
assemblies and also provides additional rack storage on the pool 
floor where space permits.
    Criterion (2):
    The storage expansion method does not involve rod consolidation 
or double tiering.
    Proposed Amendment:
    The WBN racks are not double tiered, and the racks will sit on 
the floor of the spent fuel pool. Additionally, the amendment 
application does not involve consolidation of spent fuel.
    Criterion (3):
    The k<INF>eff of the pool is maintained less than or equal to 
0.95.
    Proposed Amendment
    The design of the spent fuel racks contains a neutron absorber, 
Boral, to allow close storage of spent fuel assemblies while 
ensuring that the k<INF>eff remains less than 0.95 under normal 
operating conditions with unborated water in the pool and less than 
0.95 under abnormal conditions with soluble boron in the pool.
    Criterion (4):
    No new technology or unproven technology is utilized in either 
the construction process or the analytical techniques necessary to 
justify the expansion.
    Proposed Amendment:
    The construction processes and analytical techniques used in the 
fabrication and design are substantially the same as those of 
numerous other rack installations, Thus, no new or unproven 
technology is utilized in the construction or analysis of the high 
density, spent fuel racks at WBN. TVA's contractor, Holtec 
International, has previously supplied licensable racks of several 
similar design for about 10 other reracking projects
    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: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: November 26, 1996
    Description of amendment request: The proposed changes would 
eliminate the records retention requirements from the administrative 
section of the Technical Specifications (TS) in accordance with NRC 
Administrative Letter 95-06, ``Relocation of Technical Specifications 
Administrative Controls Related to Quality Assurance.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Specifically, operation of the Surry... Power [Station] in 
accordance with the

[[Page 14473]]

proposed Technical Specifications changes will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed 
administrative changes do not affect equipment or its operation. 
Therefore, the likelihood that an accident will occur is neither 
increase nor decreased by relocating record retention requirements 
from the Technical Specifications to the Operational Quality 
Assurance Program. This TS change will not impact the function or 
method of operation of plant equipment. Thus, a significant increase 
in the probability of a previously analyzed accident does not result 
due to this change. No systems, equipment, or components are 
affected by the proposed changes. Thus, the consequences of any 
accident previously evaluated in the UFSAR [Updated Final Safety 
Analysis Report] are not increased by this change.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed change 
does not alter the design or operations of the physical plant. Since 
record retention requirements are administrative in nature, a change 
to these requirements does not contribute to accident initiation, an 
administrative change related to this activity does not produce a 
new accident scenario or produce a new type of equipment 
malfunction. [These] changes do not alter any existing accident 
scenarios. The proposed administrative change does not affect 
equipment or its operation, and, thus, does not create the 
possibility of a new or different kind of accident. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident.
    (3) Involve a significant reduction in a margin of safety. 
Section 6.0 of the...Surry Technical Specifications does not have a 
basis description. The proposed administrative change does not 
affect equipment or its operation, and, thus, does not involve any 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Mark Reinhart, Acting

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

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

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of application for amendment: July 31, 1997, as supplemented 
February 13, 1997.
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications to reduce the minimum Reactor Coolant 
System total flow rate from 370,000 gpm to 340,000 gpm. The proposed 
changes are necessary to support a larger number of plugged steam 
generator tubes for future operating cycles.
    Date of publication of individual notice in Federal Register: 
February 26, 1997 (62 FR 8780)
    Expiration date of individual notice: March 28, 1997
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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: February 14, 1997
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications to permit a one-time extension of the 
current steam generator tube inservice inspection cycle. Date of 
publication of individual notice in Federal Register: March 4, 1997 (62 
FR 9816)
    Expiration date of individual notice: March 28, 1997
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: February 17, 1997
    Brief description of amendment: Changes to Technical Specification 
to implement 10 CFR 50, Appendix J Option B relating to containment 
leakage tests.
    Date of publication of individual notice in the Federal Register: 
February 28, 1997 (62 FR 9214).
    Expiration date of individual notice: March 31, 1997
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629

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

    Date of amendment request: February 14, 1997
    Brief description of amendment request: The proposed amendment 
would revise Technical Specification (TS) Section 3/4.5.2, ``Emergency 
Core Cooling Systems, ECCS Subsystems - T<INF>avg more than or equal to 
280 deg.F.'' Surveillance requirement 4.5.2.f would be modified to 
state that opening and closing of the inspection port on the watertight 
enclosure for the decay heat valve pit would not require this 
surveillance procedure to be performed. The applicable TS bases would 
also be changed. Date of publication of individual notice in Federal 
Register: February 26, 1997 (62 FR 8783) Expiration date of individual 
notice: March 28, 1997
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606

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

[[Page 14474]]

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

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: August 1, 1996
    Brief description of amendments: The amendments modify the 
Technical Specifications requirements to allow use of blind flanges 
during Modes 1-4 in the Calvert Cliffs 1 and 2 Containment Purge system 
instead of the two outboard 48-inch isolation valves. Date of issuance: 
March 7, 1997
    Effective date: As of the date of issuance to be implemented by the 
end of the 1998 refueling outage for Unit 1; by the end of the 1997 
refueling outage for Unit 2.
    Amendment Nos.: 221 and 197
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47975) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 7, 1997 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: December 30, 1996
    Brief description of amendment: The amendment revises chemistry 
data for TS Figures 3.4-2 and 3.4-3 and the associated Bases.
    Date of issuance: March 7, 1997
    Effective date: March 7, 1997
    Amendment No.: 68
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4342) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 1997. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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 amendments: September 20, 1996, as 
supplemented January 21, 1997.
    Brief description of amendments: The amendments would update the 
pressure- temperature cures contained in the Dresden and Quad Cities 
Technical Specifications to 22 Effective Full Power Years. Date of 
issuance: February 28, 1997 Effective date: Immediately, to be 
implemented within 30 days.
    Amendment Nos.:  153, 148, 172 and 168
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66703). The January 21, 1997, submittal provided additional 
clarifying information that did not change the original proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated February 28, 1997 No significant hazards consideration 
comments received: No
    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.

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 amendments: December 6, 1996
    Brief description of amendments: The amendments would change the 
Technical Specification (TS) by allowing a single control rod to be 
moved when the plant is in the Hot Shutdown or Cold Shutdown condition 
provided that the one-rod-out interlock is Operable and the reactor 
mode switch is in the refuel position.
    Date of issuance: March 4, 1997
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 154, 149, 173, 169
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 15, 1997 (62 FR 
2187). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 4, 1997. No significant 
hazards consideration comments received: No
    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.

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 amendments: January 6, 1997
    Brief description of amendments: The amendments would change the 
technical specifications to clarify and maintain consistency between 
the operability requirements for protective instrumentation and 
associated automatic bypass features.
    Date of issuance: March 14, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 155, 150, 174, 170

[[Page 14475]]

    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6573). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 14, 1997. No significant 
hazards consideration comments received: No
    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.

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

    Date of amendment request: November 26, 1996 as supplemented by 
letters dated December 17, 1996, March 4, 1997, and March 10, 1997
    Brief description of amendment: The amendment changes reactor 
coolant systems pressure/temperature limits to incorporate updated 
parameters and requirements.
    Date of issuance: March 14, 1997
    Effective date: March 14, 1997
    Amendment No.: 188
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4346) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 14, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: April 11, 1996 as supplemented 
by letters dated June 18, and September 5, 1996.
    Brief description of amendment: The amendment adds low-temperature 
overpressure protection requirements to the Technical Specifications as 
proposed by Generic Letter 90-06.
    Date of issuance: March 7, 1997
    Effective date: March 7, 1997, to be implemented within 30 days.
    Amendment No.: 180
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20846) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 7, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: August 23, 1996, as supplemented 
January 8, 1997 (TSCR 245)
    Brief description of amendment: The amendment updates the pressure-
temperature limits up to 22, 27, and 32 effective full power years.
    Date of Issuance: March 6, 1997
    Effective date: March 6, 1997, to be implemented within 30 days of 
issuance
    Amendment No.: 188
    Facility Operating License No. DPR-16. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47977). The January 8, 1997, letter provided clarifying information 
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 this amendment is 
contained in a Safety Evaluation dated March 6, 1997 No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
Illinois

    Date of application for amendment: October 17, 1996, as 
supplemented and modified on December 13, 1996
    Brief description of amendment: The amendment revises the Operating 
License to reflect the transfer of Soyland Power Cooperative's 13.21-
percent minority ownership of Clinton Power Station to Illinois Power 
Company. The Operating License has been revised to delete Soyland Power 
Cooperative as an owner.
    Date of issuance: March 13, 1997
    Effective date: March 13, 1997
    Amendment No.: 114
    Facility Operating License No. NPF-62: The amendment revised the 
Operating License.
    Date of initial notice in Federal Register: November 19, 1996 (61 
FR 58897) and January 29, 1997 (62 FR 4337) The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
March 13, 1997. No significant hazards consideration comments received: 
No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

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

    Date of application for amendment: June 19, 1996, and supplemented 
September 19, 1996, and December 20, 1996.
    Brief description of amendment: The amendment revises the TS to 
allow a permanent extension of the interim steam generator tube 
voltage-based repair criteria for steam generator tubes used in Cycles 
13, 14 and 15 at the Donald C. Cook Nuclear Power Plant, Unit 1.
    Date of issuance: March 13, 1997
    Effective date: March 13, 1997, with full implementation within 45 
days
    Amendment No.: 215
    Facility Operating License No. DPR-58. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40022) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 13, 1997. No significant 
hazards consideration comments received: No. The September 19, 1996, 
and December 20, 1996, letters provided additional information within 
the scope of the original application and did not change the initial 
proposed no significant hazards consideration determination.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

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

    Date of application for amendments: May 26, 1995, and supplemented 
September 26, 1995, August 2, 1996 and February 6, 1997
    Brief description of amendments: The amendments revise the TS to 
allow operation of Cook Unit 1 at steam generator tube plugging levels 
up to 30%. Additional changes to increase operating margins for both 
Unit 1 and Unit 2 are also included.
    Date of issuance: March 13, 1997
    Effective date: March 13, 1997, with full implementation within 45 
days

[[Page 14476]]

    Amendment Nos.: 214 and 199
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37095) The September 26, 1995, August 2, 1996, and February 6, 1997, 
supplements provided clarifying information that did not expand the 
scope of the initial application or change the staff's proposed no 
significant hazards determination. The Commission's related evaluation 
of the amendments is contained in a Safety Evaluation dated March 13, 
1997. No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

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

    Date of amendment request: October 17, 1996
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications relating to in-core detector 
system, seismic instrumentation, meteorological instrumentation, and 
turbine overspeed protection. The amendment deletes Limiting Conditions 
for Operation and Surveillance Requirements related to these 
instruments. The deleted requirements are to be incorporated into the 
Seabrook Station Technical Requirements Manual (SSTR). The associated 
Bases Sections are also deleted. In addition, Technical Specification 
5.5 is deleted but will not be relocated to the SSTR. The amendment 
also redesignates Paragraph 2.J of the Seabrook Operating License as 
Paragraph 3, and has added new Paragraph 2.J to document the North 
Atlantic commitment to relocate the above mentioned Technical 
Specification requirements to the SSTR.
    Date of issuance: March 12, 1997
    Effective date: March 12, 1997
    Amendment No.: 50
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications and Operating License.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66713). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 12, 1997. 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. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: May 23, 1996, as supplemented 
July 17 and December 4, 1996
    Brief description of amendment: The amendment modifies the 
description of the time constants associated with the Overtemperature 
Delta-T and Overpower Delta-T calculations used to establish the trip 
setpoints and the time constant used in the rate-lag controller for 
Steam Line Isolation, Steam Line Pressure Negative Rate-High.
    Date of issuance: March 11, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 134
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 17, 1996 (61 FR 
30639) The July 17 and December 4, 1996, letters provided additional, 
clarifying information that did not change the scope of the May 23, 
1996, application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated March 11, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Southern California Edison Company, et al., Docket No. 50-362, San 
Onofre Nuclear Generating Station, Unit No. 3, San Diego County, 
California

    Date of application for amendment: February 7, 1997
    Brief description of amendment: The amendment defers implementation 
of Surveillance Requirement 3.1.5.4 of Technical Specification 3.1.5, 
``Control Element Assembly (CEA) Alignment,'' until the next SONGS Unit 
3 shutdown, which will be no later than the upcoming Cycle 9 refueling 
outage (currently scheduled for April 12, 1997).
    Date of issuance: March 5, 1997
    Effective date: March 5, 1997
    Amendment No.: 126
    Facility Operating License No. NPF-15: The amendments revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (62 FR 7477 dated February 19, 
1997). The 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 March 21, 1997, 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 is contained in a Safety Evaluation dated March 5, 1997.
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713
    Dated at Rockville, Maryland, this 19th day of March 1997.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 97-7508 Filed 3-25-97; 8:45 am]
BILLING CODE 7590-01-F