[Federal Register Volume 61, Number 70 (Wednesday, April 10, 1996)]
[Notices]
[Pages 15985-16005]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-8786]



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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 16, 1996, through March 29, 1996. The 
last biweekly notice was published on March 27, 1996 (61 FR 13521).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S.

[[Page 15986]]

Nuclear Regulatory Commission, Washington, DC 20555, and should cite 
the publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By May 10, 1996, 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, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: September 30, 1994, as supplemented 
September 18, 1995, January 19 and March 15, 1996
    Description of amendment request: Currently, the steam generators 
(SGs) in place in the Catawba units are Westinghouse Model ``D'' type 
preheat SGs. The tube degradation levels in the SGs at Catawba Unit 1 
have affected the reliability of the unit. Therefore, these generators 
are scheduled to be replaced with feedring SGs designed by Babcock

[[Page 15987]]
& Wilcox International. The design differences and analysis changes to 
support the feedring SGs result in the need to change the Technical 
Specifications (TS) in the following areas: (a) revise low-low SG water 
level for the reactor trip setpoint in TS Table 2.2-1 and for auxiliary 
feedwater actuation in TS Table 3.3-4, (b) revise high-high SG water 
level setpoint for turbine trip and feedwater isolation in TS Table 
3.3-4, (c) delete reference to SG tube repair methods which will no 
longer be applicable after the replacement of the SGs and clarify 
initial surveillances, (d) revise reactor coolant system volume, (e) 
update Topical Report revision numbers in the Administrative Controls 
Section 6.9 of the TS, and (f) change the nominal average temperature 
in TS Table 2.2-1 for the reactor trip system setpoints to reflect the 
value incorporated into the safety analyses for the replacement SGs. 
The change made in the September 30, 1994, submittal, to reduce the 
steam line safety valve lift settings in TS Table 3.7-2, was withdrawn 
in the September 18, 1995, submittal. The January 19, 1996, submittal 
proposed changes to reflect the NRC's approved revisions to Topical 
Reports DPC-NE-3000 and DPC-NE-3002. The March 15, 1996, submittal 
provided additional information in response to NRC staff requests and 
also updated and clarified the involved TS pages including changes made 
to these TS pages by license amendments issued on other topics since 
the original application dated September 30, 1994.
    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:

    Operation of Catawba Unit 1 in accordance with the proposed 
changes to the Technical Specifications will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The low-low steam generator water 
level reactor trip setpoint, the high-high steam generator water 
level setpoint for turbine trip and feedwater isolation, and the 
low-low steam generator water level setpoint for auxiliary feedwater 
initiation are changing to support operation with the replacement 
steam generators. These setpoints were chosen both to optimize plant 
operation, and ensure that all applicable acceptance criteria are 
met for licensing basis safety analysis. These setpoints do not 
contribute to the initiation of any accident evaluated in the 
Catawba FSAR [Final Safety Analysis Report] and have no adverse 
impact on system operation, therefore it can be concluded that these 
changes will not significantly increase the probability or 
consequences of an accident evaluated in the FSAR.
    The increase in Reactor Coolant System volume due to the 
replacement steam generators will not increase the probability or 
consequences of an accident previously evaluated. The increase in 
volume has no effect on the probability of occurrence of any 
accident evaluated in the FSAR. The mass and energy release inside 
containment due to postulated loss of coolant accidents inside 
containment has been analyzed to ensure that the peak containment 
pressure limit is not exceeded. All Chapter 15 reanalysis which was 
required due to the replacement steam generators assumed the new 
Reactor Coolant System volume. Since the results of these analyses 
show the applicable acceptance criteria continue to be met, it can 
be concluded that the consequences of an accident previously 
evaluated are not significantly increased due to this change.
    Operation of Catawba Unit 1 in accordance with the proposed 
changes to the Technical Specification will not create the 
possibility of a new or different accident from any accident 
previously evaluated. The proposed changes to revise the low-low 
steam generator water level reactor trip setpoint, high-high steam 
generator water level setpoint for turbine trip and feedwater 
isolation, and low-low steam generator water level setpoint for 
auxiliary feedwater initiation ensure that the appropriate 
acceptance criteria for FSAR Chapter 15 transients which rely on 
these functions are met for operation with the replacement steam 
generators. ... The increase in Reactor Coolant System volume is 
taken into account in the analysis of the mass and energy release 
due to a postulated loss of coolant inside containment, and Chapter 
15 events which have been reanalyzed due to replacement of the steam 
generators. As discussed above, the proposed changes will not 
introduce the possibility of a new or different accident from any 
previously evaluated, they will ensure that transients that take 
credit for these functions and dose analyses meet applicable 
acceptance criteria for operation with the replacement steam 
generators.
    Operation of Catawba Unit 1 in accordance with the proposed 
changes to the Technical Specifications will not involve a 
significant reduction in a margin of safety. The proposed changes 
were made to ensure that transients that rely on low-low steam 
generator water level reactor trip setpoint, high-high steam 
generator water level setpoint for turbine trip and feedwater 
isolation, and low-low steam generator water level setpoint for 
auxiliary feedwater actuation meet applicable acceptance criteria. 
... The proposed change in the Reactor Coolant System volume will 
not involve a significant reduction in a margin of safety. The 
increased volume affects the mass and energy release due to a 
postulated loss of coolant accident inside containment and the other 
Chapter 15 events which were reanalyzed due to replacement of the 
steam generators. This event has been analyzed and the results are 
within current acceptable limits. As discussed above, the acceptance 
criteria for FSAR transients which are affected by these proposed 
changes continue to be met, therefore there is no significant 
reduction in the margin of safety.
    Changes to the steam generator surveillance requirements will 
simply delete inspection requirements which are no longer applicable 
after installation of the replacement steam generators. References 
to F* criteria, interim plugging criteria, and sleeving are deleted 
since these repair criteria were approved for use on the current 
steam generators. Since these changes only delete criteria which 
will no longer be applicable and cannot be used, no significant 
hazards considerations are involved.
    The changes to Technical Specification 6.9.1.9 are 
administrative in nature. These changes are being made to reflect 
the most recent revisions of DPC-NE-3002 and DPC-NE-3000, which 
includes changes associated with the replacement steam generators. 
These topical report revisions [have been] reviewed and approved for 
use regarding McGuire and Catawba Nuclear Stations. Since these 
changes are administrative in nature, no significant hazards 
considerations are involved.
    The proposed change to Technical Specifications [average coolant 
temperature in Table 2.2-1] does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Changing the value for [the average coolant temperature] 
in Notes 1 and 3 of Table 2.2-1 will update the value to agree with 
[the average coolant temperature] assumed in the applicable safety 
analyses for replacement of the steam generators. Acceptable results 
were obtained for all required reanalyses. The probability of an 
accident will not be significantly affected by operation with the 
new [average coolant temperature] value, because all equipment will 
be operated within acceptable design limits. The consequences of 
previously evaluated accidents which are affected by this change 
have been evaluated, and have been determined to be within 
acceptable limits.
    This proposed change [to TS Table 2.2-1] will not create the 
possibility of a new or different kind of accident from any 
previously evaluated. This change does not change the physical 
configuration of the plant, and all analyses which are affected by 
replacement of the steam generators have been determined to have 
acceptable results assuming this value for [average coolant 
temperature].
    This proposed change to the Technical Specifications [Table 2.2-
1] will not involve a significant reduction in the margin of safety. 
All safety analyses which were affected by replacement of the steam 
generators assumed this value for [average coolant temperature] and 
the results were determined to be within previously acceptable 
limits.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

[[Page 15988]]

    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: January 12, 1996, as supplemented March 
4, 1996
    Description of amendment request: This request was previously 
published in the Federal Register on January 31, 1996 (61 FR 3498). It 
is being renoticed to provide clarification to the scope of the 
original request. Compliance with 10 CFR Part 50, Appendix J, provides 
assurance that the primary containment, including those systems and 
components that penetrate the primary containment, do not exceed the 
allowable leakage rate values specified in the Technical Specifications 
(TS) and Bases. The allowable leakage rate is determined so that the 
leakage assumed in the safety analyses is not exceeded.
    On September 12, 1995, the NRC approved issuance of a revision to 
10 CFR Part 50, Appendix J, which was subsequently published in the 
Federal Register on September 26, 1995, and became effective on October 
26, 1995. The revision added Option B ``Performance-Based 
Requirements'' to Appendix J to allow licensees to voluntarily replace 
the prescriptive testing requirements of Appendix J with testing 
requirements based on both overall and individual component leakage 
rate performance.
    Regulatory Guide 1.163, ``Performance-Based Containment Leak Test 
Program,'' was developed as a method acceptable to the staff for 
implementing Option B. Accordingly, the licensee has submitted, in its 
application dated January 12, 1996, proposed changes to the TS to 
implement 10 CFR Part 50, Appendix J, Option B, by referring to 
Regulatory Guide (RG) 1.163, ``Performance-Based Containment Leakage-
Test Program.'' Although the licensee's proposal indicated that it was 
consistent with RG 1.163, it did not include the clarifying changes to 
the TS that would require the visual examination of containment systems 
to be consistent with the guidance of RG 1.163. The licensee submitted 
a supplement, dated March 4, 1996, to its January 12, 1996 proposal, 
which proposes such changes to TS Surveillance Requirements 4.6.1.6 and 
4.6.1.7 and associated Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Containment leak rate testing is not an initiator of any 
accident; the proposed change does not affect reactor operations or 
accident analysis, and has no significant radiological consequences. 
Therefore, this proposed change will not involve an increase in the 
probability or consequences of any previously-evaluated accident.
    2. The proposed change will not create the possibility of any 
new not previously evaluated.
    The proposed change does not affect normal plant operations or 
configuration, nor does it affect leak rate test methods. The test 
history at Catawba (no ILRT [integral leak rate test] failures) 
provides continued assurance of the leak tightness of the 
containment structure.
    3. There is no significant reduction in a margin of safety.
    The proposed changes are based on NRC-accepted provisions, and 
maintain necessary levels of reliability of containment integrity. 
The performanced-based approach to leakage rate testing recognizes 
that historically good results of containment testing provide 
appropriate assurance of future containment integrity; this supports 
the conclusion that the impact on the health and safety of the 
public as a result of extended test intervals is negligible.
    Based on the above, no significant hazards consideration is 
created by the proposed change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: March 11, 1996
    Description of amendment request: The proposed amendment would 
increase the alarm setpoints of the in-containment high range area and 
containment purge radiation monitors. These alarm setpoints are 
specified in Table 3.3-6 of Technical Specification 3.3.3.1. The 
proposed amendment would also include several editorial changes.
    The proposed change to the in-containment high range area radiation 
monitor alarm setpoint would make the setpoint consistent with the 
Beaver Valley Power Station Emergency Action Levels (EALs) approved by 
the NRC in August 1994. These EALs use the in-containment high 
radiation area monitors as indication of fission product barrier 
challenges or failures.
    The containment purge radiation monitors are provided to: (1) 
analyze the ventilation effluent from the reactor containment building, 
(2) detect abnormal releases and isolate the release if the setpoint is 
reached or exceeded, and (3) alert refueling personnel of the need to 
evacuate affected areas so as to maintain occupational exposures as low 
as reasonably achievable. The proposed increase in this setpoint value 
provides alarm and isolation based on offsite dose considerations and 
will provide greater operational flexibility since inadvertent 
engineered safety feature actuations due to evacuation alarms caused by 
minor (greater than three times background) increases in radiation 
levels will be minimized.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed monitor alarm setpoint changes and editorial 
changes are administrative in nature. Should the in-containment high 
range area monitors fail to annunciate or give a false alarm, there 
would be no effect on any other plant equipment or systems. These 
monitors are safety related; however, they do not initiate any 
safety function, nor do they interface with any other safety related 
system. The monitors' alarm as a visual (lighted icon) and audible 
alarm in the control room. The operator is then responsible for 
taking any corrective actions necessary, based on the alarm and 
Emergency Action Level (EAL) guidelines. The in-containment high 
range area monitors do not provide for any automatic actions of 
other equipment or systems when an alarm condition occurs.
    The containment purge monitors are also safety related with the 
ability for an operator to input a radiation level value for high 
alarm levels during Mode 6, which upon actuation, create both a 
visual (lighted icon) and

[[Page 15989]]
audible alarm in the control room. At the high alarm level, each 
monitor automatically sends a signal to close the purge supply and 
exhaust isolation dampers in the containment building. A change in 
the value of the alarm setpoint has no effect on the performance of 
the containment purge and exhaust system. The high alarm and 
subsequent automatic termination of a radioactive release will now 
be based on offsite dose considerations. There is no credible 
failure of the monitors associated with a change of the alarm 
setpoint value.
    The operating and design parameters of the subject radiation 
monitors will not change. The proposed change affects only 
theradiation level at which an alarm condition is created and does 
not affect any accident assumptions. The in-containment high range 
area monitors' alarm setpoint change will not affect the 
radiological consequences of an accident. However, since the 
containment purge monitors revised setpoint is based on offsite 
doses consequences and is a higher value than the current setpoint 
of three times the background radiation level, the postulated 
offsite radiological consequences of a fuel handling accident inside 
containment would be increased. An analysis of a fuel handling 
accident inside containment with the purge and exhaust system 
discharging through the Supplementary Leak Collection and Release 
System (SLCRS) filter trains was performed and a summary of this 
analysis is to be added to Chapter 15 of the Updated Final Safety 
Analysis Report (UFSAR). The analysis which determined the 
containment purge monitors' setpoint postulated offsite doses that 
are less than a small fraction (less than twenty-five percent) of 
the 10 CFR Part 100 guidelines. The fuel handling accident inside 
containment calculation demonstrated control room operator doses 
that comply with General Design Criteria (GDC) 19. Therefore, the 
increased radiological consequences of the change in the alarm 
setpoint are acceptable. The analysis assumed no isolation, so 
isolation actuated by the monitor alarm will reduce doses further.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed radiation monitor alarm revisions cannot initiate a 
new type of accident. The referenced radiation monitors' alarms 
cannot initiate a new type of accident, since even a failure of the 
monitor itself cannot serve as the initiating event of an accident. 
Operator action is not made solely on a radiation monitor alarm; 
other plant condition indicators are also evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The in-containment high range area monitors have no capability 
to mitigate the consequences of an accident and do not interface 
with any safety related system. These monitors are safety related 
channels which provide indication to the operator of the integrity 
of the fission product barriers incontainment. This indication, 
combined with other indications of plant conditions may direct an 
operator to take action to mitigate the consequences of an accident. 
The alarm setpoint itself does not perform any specific safety 
related function and the trip value is not referenced in the UFSAR, 
nor does any site design basis document take credit for this 
setpoint. Safety limits and limiting safety system settings are not 
affected by this proposed change. The site will continue to meet the 
requirements of 10 CFR Part 100 which limits offsite dose following 
a postulated fission product release.
    The containment purge monitors' revised setpoint is based on 
offsite dose consequences and is a higher value than the current 
setpoint of three times the background radiation level. Thus the 
postulated offsite radiological consequences of a fuel handling 
accident inside containment are increased which reduces the current 
margin of safety. An analysis of a fuel handling accident inside 
containment with the purge and exhaust system discharging through 
the SLCRS filter trains was performed and a summary of this analysis 
will be added to Chapter 15 of the UFSAR. The analysis postulated 
offsite doses to be less than twenty-five percent of the 10 CFR Part 
100 guidelines and control room operator doses that comply with GDC 
19. The analysis shows that the increased radiological consequences 
of the change in the alarm setpoint are acceptable. Further, the 
analysis assumed that no isolation would occur; therefore, isolation 
actuated by the monitors' alarm will reduce the postulated doses.
    Therefore, use of the proposed technical specification would not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 1500l.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of amendment request: March 5, 1996
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) as follows:
    (1) TS Surveillance Requirement (SR) 4.4.3.3: Delete the 
requirement for testing the switching capability for pressurizer heater 
power supplies on an 18-month interval.
    (2) TS SR 4.5.2.d: Change the containment sump inspection 
requirements from each containment entry to once daily if a containment 
entry has been made and upon the final entry prior to establishing 
CONTAINMENT INTEGRITY.
    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) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed amendments conform to the uidance given in 
Enclosure 1 of the NRC Generic Letter 93-05. The overall functional 
capabilities of the pressurizer heater system and the Emergency Core 
Cooling System (ECCS) will not be modified by the proposed changes. 
These amendments will not involve a significant increase in the 
probability or consequences of an accident previously evaluated for 
the following reasons:
    (1) Deleting the requirement to test the switching capabilities 
of the pressurizer heater emergency power supplies will reduce an 
unnecessary testing requirement since the pressurizer heaters are 
already connected to the emergency bus.
    (2) Increasing the interval of containment sump inspections to 
once daily if containment has been entered and upon final entry will 
reduce unnecessary personnel exposure from performance of 
containment sump inspections for each containment entry.
    [The staff notes that although statement (2) is correct, it does 
not provide a reason why the amendments will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The staff finds that once daily 
inspection of the containment adequately ensures that the 
containment sump remains free of debris.]
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The use of the proposed changes to the TS can not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined

[[Page 15990]]
in the facility operating license. No new failure mode is introduced 
due to the surveillance changes and inspection requirements, since 
the proposed changes do not involve the addition or modification of 
equipment nor do they alter the design or operation of affected 
plant systems.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems are unchanged by the proposed amendments. The proposed 
changes to the TS which establish new or clarify old surveillance 
and inspection requirements [are] consistent with the NRC Generic 
Letter 93-05 line-item improvement guidance [and] do not 
significantly reduce any of the margins of safety even though the 
number of surveillances is decreased. These requested amendments are 
justified by the following reasoning from NUREG-1366:
    (1) The surveillance or inspection results in radiation exposure 
to plant personnel which is not justified by the safety significance 
of the surveillances as in the case of the containment sump 
inspection requirements.
    (2) The surveillance places an unnecessary burden on plant 
personnel because the time required is not justified by the safety 
significance of the surveillance as in the emergency power switching 
requirements for the pressurizer heater system.

    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: Florida International 
University, University Park, Miami, Florida 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: Eugene V. Imbro

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

    Date of amendment requests: February 22, 1996 (AEP:NRC:1243)
    Description of amendment requests: The proposed amendments would 
revise the technical specifications to reference NRC Regulatory Guide 
1.9, Revision 3 rather than NRC Regulatory Guide 1.108, Revision 1 
criteria for the determination of a valid diesel generator test.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Per 10 CFR 50.92, proposed changes do not involve a significant 
hazards consideration if the changes do not:
    1. involve a significant increase in the probability [or] 
consequences of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety
    Criterion 1
    This amendment request does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the proposed change to the T/S [technical 
specifications] does not affect the assumptions, parameters, or 
results of any UFSAR [updated final safety analysis report] accident 
analysis.
    The proposed amendment does not modify any existing equipment, 
and the proposed acceptance criteria for diesel generator testing 
will conform to NRC guidance. Based on these considerations, it is 
concluded that the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Criterion 2
    The proposed changes do not involve physical changes to the 
plant or changes in plant operating configuration. The proposed 
changes update guidance for diesel generator testing. Thus, it is 
concluded that the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Criterion 3
    The proposed changes update guidance for the testing of diesel 
generators. The guidance is endorsed by the NRC in Regulatory Guide 
1.9, and compliance with this guidance will ensure the operability 
of the diesel generators. Thus, there is no significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Mark Reinhart, Acting

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

    Date of amendment request: December 7, 1995
    Description of amendment request: The proposed change will remove 
the requirement that primary containment always be purged or vented 
through the standby gas treatment (SBGT) system and adds requirements 
that would limit the use of SBGT for purging and venting. The proposed 
amendment also makes editorial changes and revises the associated Bases 
section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    NNECO has reviewed the proposed change in accordance with 
10CFR50.92 and concluded that the change does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed change will allow primary containment to be purged 
or vented without the use of the SBGT system. This change only 
modifies the alignment of the atmospheric control system for purging 
or venting containment. The change does not affect any primary 
system, nor does it affect the ability of the containment isolation 
valves to close. As such, the proposed change can not affect the 
probability of occurrence of an accident previously analyzed. This 
change increases the possibility that some initial post-accident 
containment atmosphere could be released directly to the atmosphere 
at the top of the 375 foot stack prior to the closure of the 
containment isolation valves. However, this condition is bounded by 
the original radiological release analysis. This is balanced by the 
increased likelihood that post-accident reactor building atmosphere 
(from the time that the containment isolation valves close) is 
processed by the SBGT system.
    The proposed technical specification also establishes strict 
controls for the use of the SBGT system for purging and venting 
containment atmosphere. This includes disabling the automatic 
initiation of the train not in use and relying on a dedicated 
operator to initiate the remaining train, should a DBA [design basis 
accident] occur. Since SBGT system operation does not affect the 
initiation of any postulated accident, disabling the automatic 
initiation and relying upon operator action to start the remaining 
train can not affect the probability of an accident previously 
evaluated. The failure of the train to start within one minute 
following the DBA could increase the consequences of

[[Page 15991]]
an analyzed accident. To ensure timely initiation, NNECO has 
implemented a procedure for purging or venting through the SBGT 
system which establishes a dedicated operator whose function at the 
onset of a DBA is to isolate the train in use (the train expected to 
be damaged by the pressure spike), verify the open AC [atmospheric 
control] valves go closed, and then start the second train. This 
procedure has been validated to ensure that these actions can be 
completed within one minute.
    Although not expected, a delay in operator action to initiate 
the SBGT has been evaluated for impact upon the radiological 
consequences. The evaluation shows that the offsite doses remain 
well within the 10CFR100 limit even if the operator actions are not 
completed until three minutes after the DBA occurs.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change allows removal of the SBGT system from the 
release path for normal containment purge and venting. The change 
does not affect the frequency or requirement for venting. Nor does 
the proposed LCO [limiting condition for operation] affect the 
processes of venting or purging primary containment; the same 
penetrations and containment isolation valves will continue to be 
used. All purging and venting functions can still be performed when 
required by existing specifications and plant procedures. The 
proposed change does not diminish the capability of any isolation 
valve for performing its isolation function.
    Therefore, the proposed change can not create a new or different 
kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The affect of this change has been analyzed against the criteria 
of 10CFR100 and 10CFR20. The potential release which may occur as a 
result of a postulated DBA while purging or venting directly to the 
stack will not exceed the limits of 10CFR100. Likewise, the 
technical specifications and administrative controls established for 
purging or venting through the SBGT minimize the potential for an 
unfiltered release should a DBA occur during that evolution. 
Further, the amount of time that a SBGT train is aligned to primary 
containment is expected to be substantially reduced from that 
required by the existing Technical Specification. Decreasing the 
amount of time that SBGT is aligned to primary containment decreases 
the possibility that a DBA would occur while in such an alignment.
    Finally, the potential increase in dose which could occur as a 
result of normal purge and vent activities will be controlled such 
that it remains below acceptable limits.
    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: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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: January 17, 1996
    Description of amendment requests: The proposed amendment would 
revise selected technical specifications (TS) in accordance with the 
NRC's Final Policy Statement on TS Improvements for Nuclear Power 
Reactors and relocate the TS to the Diablo Canyon Power Plant Equipment 
Control Guidelines. The proposed change would also create TS 6.8.4.j, 
``Explosive Gas and Storage Tank Radioactivity Monitoring Program.'' 
Some of the TS would be relocated and maintained in accordance with 
this program. Specifically, the following TS would be relocated: TS 
3.1.2.1, ``Boration Systems Flow Path - Shutdown,'' TS 3.1.2.3, 
``Charging Pumps - Shutdown,'' TS 3.1.2.4, ``Charging Pumps - 
Operating,'' TS 3.1.2.5, ``Borated Water Sources - Shutdown,'' TS 
3.1.2.6, ``Borated Water Sources - Operating,'' TS 3.3.3.2, ``Movable 
Incore Detectors,'' TS 3.3.3.4, ``Meteorological Instrumentation,'' TS 
3.3.3.10, ``Explosive Gas Effluent Monitoring Instrumentation,'' TS 
3.9.3, ``Decay Time,'' TS 3.9.5, ``Communications,'' TS 3.9.6, 
``Manipulator Crane,'' TS 3.9.7, ``Crane Travel - Fuel Handling 
Building,'' TS 3.9.10.2, ``Water Level - Reactor Vessel - Control 
Rods,'' TS 3.9.13, ``Spent Fuel Shipping Cask Movement,'' TS 3.10.1, 
``Special Test Exceptions - Shutdown Margin,'' TS 3.10.4, ``Position 
Indication System - Shutdown,'' TS 3.11.1.4, ``Liquid Holdup Tanks,'' 
TS 3.11.2.5, ``Explosive Gas Mixture,'' and TS 3.11.2.6, ``Gas Storage 
Tanks.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes simplify the Technical Specifications (TS), 
meet regulatory requirements for relocated TS, and implement the 
recommendations of the Commission's Final Policy Statement on TS 
Improvements and revised 10 CFR 50.36. Future changes to these 
requirements will be controlled by 10 CFR 50.59. The proposed 
changes are administrative in nature and do not involve any 
modifications to any plant equipment or affect plant operation.
    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.
    The proposed changes are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by which any safety-related system performs 
its function. Also, no changes to the operation of the plant or 
equipment are involved.
    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.
    The proposed changes involve relocating TS requirements to a 
licensee-controlled document. The requirements to be relocated were 
identified by applying the criteria endorsed in the Commission's 
Final Policy Statement, which is included in the new revision of 10 
CFR 50.36, and are consistent with NUREG-1431, Rev. 1. Thus, the 
proposed changes do not alter the basic regulatory requirements and 
do not affect any safety analysis.
    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
    NRC Project Director: William H. Bateman
    
[[Page 15992]]


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

    Date of amendment request: February 12, 1996
    Description of amendment request: The amendments would revise the 
Susquehanna Units 1 and 2 Technical Specifications establish and 
reference a Primary Containment Leakage Rate Testing Program in order 
to implement 10 CFR 50, Appendix J, Option B in accordance with the 
guidelines contained in Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Test Program'', dated September 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The proposed license amendments revise the 
Technical Specifications to reflect the adoption of a performance-
based containment leakage-testing program. The Nuclear Regulatory 
Commission has approved the use of a performance-based option for 
containment leakage testing programs when it amended 10 CFR Part 50, 
Appendix J (60 FR 49495).
    To adopt of (sic) the revised regulations, licensees are 
required to incorporate into their Technical Specifications, by 
general reference, the NRC regulatory guide or other plant specific 
implementing document. A new Administrative Controls Specification 
is being added to the Susquehanna SES Technical Specifications that 
requires the establishment and maintenance of a Primary Containment 
Leakage Rate Testing Program. As stated in the Technical 
Specification, this Primary Containment Leakage Rate Testing Program 
will conform with NRC Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Rate Testing Program'', dated September 1995. The 
Primary Containment Leakage Rate Testing Program establishes 
requirements intended to ensure on-going containment integrity, 
including the performance of a periodic general visual inspection of 
the containment to detect early indications of structural 
deterioration.
    The effect of increasing containment leakage rate testing 
intervals has been evaluated by the Nuclear Energy Institute using 
the methodology described in NUREG-1493 and historical 
representative industry leakage rate testing data. The results of 
this evaluation, as published in NEI 94-01, Revision 0, are that the 
increased risk corresponding to the extended test interval is small 
(less than 0.1 percent of total risk) and compares well to the 
guidance of the NRC's safety goal. The primary containment leak rate 
data and component performance history at Susquehanna SES are 
consistent with the conclusions reached in NUREG-1493 and NEI 94-01. 
Therefore, adoption of performance-based verification of leakage 
rates for isolation valves, containment penetrations, and the 
overall containment boundary will provide an equivalent level of 
safety and does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    No safety-related equipment, safety function, or plant 
operations will be altered as a result of the proposed license 
amendment.
    The safety objective for the primary containment is stated in 10 
CFR 50, Appendix A, ``General Design Criteria for Nuclear Power 
Plants.'' The safety function of the primary containment will be met 
since the containment will continue to provide ``an essentially leak 
tight barrier against the uncontrolled release of radioactivity to 
the environment...'' for postulated accidents. Therefore, the 
proposed license amendments will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    As stated above, the Nuclear Regulatory Commission has approved 
the use of a performance-based option for containment leakage 
testing programs when it amended 10 CFR Part 50, Appendix J (60 FR 
49495). The new Primary Containment Leakage Rate Testing Program 
will conform with NRC Regulatory Guide 1.163, Revision 0, dated 
September 1995, ``Performance-Based Containment Leak-Rate Testing 
Program'' by requiring that leakage testing intervals be established 
based on the criteria in Section 11.0 of NEI 94-01, Revision 0.
    As discussed in Part 1 above, the effect of increasing 
containment leakage rate testing intervals has been evaluated by the 
Nuclear Energy Institute using the methodology described in NUREG-
1493 and historical representative industry leakage rate testing 
data. The results of this evaluation, as published in NEI 94-01, 
Revision 0, are that the increased safety risk corresponding to the 
extended test intervals is small (less than 0.1 percent of total 
risk) and compares well to the guidance of the NRC's safety goal. In 
addition, as demonstrated by risk analyses contained in NUREG-1482, 
relaxation of the integrated leak rate test frequency does not 
significantly increase the probability or consequences of a 
previously evaluated accident. Integrated leakage rate tests have 
been demonstrated to be of limited value in detecting significant 
leakages from penetrations and isolation valves. The primary 
containment leak rate data and component performance history at 
Susquehanna SES are consistent with the conclusions reached in 
NUREG-1493 and NEI 94-01. Therefore, the proposed license amendments 
adopting a performance-based approach for verification of leakage 
rates for isolation valves, containment penetrations, and the 
containment overall will continue to meet the regulatory goal of 
providing an essentially leak-tight containment boundary, will 
provide an equivalent level of safety, and do not involve a 
significant reduction in a margin of safety.
    The revised Technical Specifications will continue to maintain 
the allowable leak rate (La) as the Type A test performance 
criterion. In addition, a requirement to perform a periodic general 
visual inspection of the containment is part of the performance-
based leakage testing program.
    The revised Technical Specifications will continue to maintain 
the allowable leak rate (La) as the Type B and C tests' performance 
criterion. As supported by the findings of NUREG-1493, the 
percentage of leakages detected only by integrated leak rate tests 
is small (only a few percent) and Type B and C leakage tests are 
capable of detecting more than 97 percent of containment leakages 
and virtually all such leakages are identified by local leak rate 
tests (LLRTs) of containment isolation valves.
    Thus, the proposed license amendments do not involve a 
significant reduction in a margin of safety and will continue to 
support the regulatory goal of ensuring an essentially leak-tight 
containment boundary.

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

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

    Date of amendment request: February 23, 1996
    Description of amendment request: The proposed amendment would 
change the Technical Specification (TS) Surveillance Requirement 
4.6.2.1d concerning drywell-to suppression chamber bypass testing. 
Currently, Susquehanna TSs require the performance of a bypass test at 
40 plus or minus 10-month intervals. The proposed TS change would 
request that the bypass test interval be revised to correspond with the 
interval for Primary Containment Integrated Leak Rate

[[Page 15993]]
Testing (ILRT) under 10 CFR Part 50, Appendix J, Option B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to allow bypass testing at the [Integrated 
Leak Rate Testing] interval involves no physical or operational 
changes to the Susquehanna SES. Reviews of bypass leakage test 
results at Susquehanna and other similarly designed plants confirm 
that minimal suppression pool bypass leakage has occurred. Based on 
this data, the risk of suppression pool bypass leakage from non 
vacuum breaker sources is no greater than that of other primary 
containment passive structures which are tested at the ILRT 
frequency. Leak testing of the drywell-to-suppression chamber vacuum 
breakers will continue to be performed on a refueling and inspection 
outage frequency to ensure that their contribution to the leakage 
area is acceptable. In addition, inspection of the diaphragm slab 
within the testing interval provides additional assurance that any 
degradation to the structure will be detected and resolved. 
Therefore, the pressure suppression capability of the containment is 
not reduced from the existing design, and there will be no 
significant increase in the probability or consequences of an 
accident previously evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change to allow bypass testing at the ILRT interval 
involves no physical or operational changes to the Susquehanna SES. 
The surveillance change does not impact the LOCA response of the 
units, or impact the design basis of the units in any way. 
Therefore, the possibility of a new or different kind of accident 
will not be created.
    III. This change does not result in a significant reduction in a 
margin of safety.
    The drywell-to-suppression chamber bypass leak test data 
obtained during previous testing at Susquehanna SES and other 
similarly designed plants demonstrates conformance, by a large 
margin, to the Technical Specification and design leakage 
requirements. The test data and safety analysis provided here 
indicate that there is negligible risk that the bypass leakage will 
change adversely in future years. Furthermore, the proposed 
performance based test methodology is judged to be acceptable based 
on the small risk of bypass leakage through paths other than those 
containing the suppression pool vacuum breakers. Testing of the 
bypass leak pathway containing the vacuum breakers will be used to 
verify acceptable bypass leakage during those outages when the 
bypass leak test is not performed. In addition, periodic visual 
inspection of the diaphragm slab within the bypass test interval 
provides additional assurance that any degradation to the structure 
will be detected and resolved.
    Testing of the bypass leakage pathways containing vacuum 
breakers, with stringent acceptance criteria, combined with the 
other negligible potential leakage areas, and periodic inspection of 
the diaphragm slab, provide an acceptable level of assurance that 
the bypass leakage will be minimized. The proposed performance based 
approach to bypass testing and inspection ensures that adverse 
conditions can be detected and corrected such that the existing 
level of confidence that the primary containment will function as 
required during a LOCA is maintained. Therefore, the proposed 
Technical Specification changes do not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: February 29, 1996
    Description of amendment request: The proposed amendment relocates 
Technical Specification 3/4.9.6, ``Refueling Platform,'' to the 
Technical Requirements Manual, which is controlled under the 
requirements of 10 CFR 50.59.
    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. Involves a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change relocates the provisions of the Refueling 
Platform that are contained in the Technical Specifications and 
places them in the Technical Requirements Manual. Review and 
approval of those portions of the Refueling Platform requirements 
contained in the Technical Requirements Manual and revisions thereto 
will be the responsibility of the Plant Operations Review Committee 
just as it was their responsibility to review changes to the 
refueling platform Limiting Condition for Operation and Surveillance 
Requirements when they were part of the Technical Specifications. 
Requiring review by the Plant Operations Review Committee reinforces 
the importance of the Technical Requirements manual and the 
requirements controlled by it and assures a multidisciplined review. 
Approved Technical Requirements or changes thereto are provided to 
the Susquehanna Review Committee for information. No design basis 
accidents are affected by the change, nor are safety systems 
adversely affected by the change. Therefore, these changes will not 
result in any change to current Technical Specification 
requirements, but will reduce the level of regulatory control 
associated with the identified requirements. The level regulatory 
control has no impact on the probability or the consequences of an 
accident previously evaluated, therefore, the proposed change will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change relocates the provisions of the Refueling 
Platform that are contained in the Technical Specifications and 
places them in the Technical Requirements Manual. This change will 
not involve any physical changes to the Refueling Platform and its 
associated instrumentation nor any changes in the manner in which 
this equipment is operated, maintained, tested or inspected. Future 
changes to these relocated requirements or surveillances will be 
evaluated in accordance with the requirements of 10CFR50.59. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety is not reduced. The relocated requirements 
do not meet any of the four criteria in the NRC Policy Statement 
used for defining the scope of Technical Specifications. In 
addition, the relocated requirements and surveillances for the 
refuel platform and associated instrumentation remain the same as 
stated in the existing Technical Specifications. Future changes to 
these relocated requirements or surveillances will be evaluated in 
accordance with the requirements of 10CFR50.59. Review and approval 
of those portions of the Refueling Platform requirements contained 
in the Technical Requirements Manual and the revisions thereto will 
be the responsibility of the Plant Operations Review Committee just 
as it was their responsibility to review changes to the refueling 
platform Limiting Condition for Operation and Surveillance 
Requirements when they were part of the Technical Specifications. 
Approved Technical Requirements or changes thereto are provided to 
the Susquehanna Review Committee for information. Therefore, no

[[Page 15994]]
significant reduction in a margin to safety is proposed.

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

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

    Date of amendment request: February 29, 1996
    Description of amendment request: The proposed amendment removes 
the Rod Block Monitor (RBM) requirements from the Technical 
Specifications, thereby reducing the number of rod movements during 
power maneuvers.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change removes the Rod Block Monitor requirements 
from Technical Specifications based on no credit being taken for the 
RBM in the reload licensing analysis. The RBM was originally 
designed to prevent fuel damage during the Rod Withdrawal Error 
[RWE] event by automatically stopping control rod motion before any 
fuel design limits are exceeded. However, due to control rod drift 
events in which the RBM can not (sic) stop control rod motion, the 
RWE is analyzed without taking credit for the RBM. The results of 
this analysis are operating limits that prevent fuel damage from a 
RWE in which control rod motion is not stopped by the RBM. 
Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This proposed change of removing the RBM requirements from 
Technical Specifications does not change the currently approved 
approach for performing the reload licensing analysis for either 
Unit. To date all reload analyses have been performed considering 
the rod drift event as a moderate frequency event and no credit 
being taken for the RBM. Since no credit is taken, removal of these 
requirements from Technical Specifications does not impact the 
current approach for performing reload analysis. Therefore, this 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Continued compliance to the governing General Design Criteria 
[GDC] for the RWE analysis assumes an appropriate margin of safety.
    GDC 10 is met when the specified acceptable fuel design limits 
(SAFDLs) are not exceeded for the RWE. The first SAFDL requires that 
a MCPR [Minimum Critical Power Ratio] Operating Limit be determined 
such that the reduction of MCPR margin due to an RWE does not 
violate the MCPR Safety Limit. The second SAFDL requires that the 
uniform cladding strain does not exceed 1% during an RWE. PP&L's 
[Pennsylvania Power and Light Company] licensing analysis of the 
RWE, without taking credit for the RBM, determines a MCPR Operating 
limit such that the reduction of MCPR margin due to an RWE does not 
violate the MCPR Safety Limit and validates that the maximum uniform 
cladding strain is less than 1%. Therefore, the applicable SAFDLs 
for the RWE are satisfied and the GDC requirements met.
    GDC 20 is met when the reactivity control system is 
automatically actuated to prevent exceeding the SAFDLs. PP&L's 
licensing analysis of the RWE, without taking credit for the RBM, 
conservatively determines a MCPR Operating Limit and validates that 
the maximum uniform cladding strain is less than 1%. Therefore, 
actuation of the RBM is not necessary to prevent exceeding the 
applicable SAFDLs for the RWE.
    GDC 25 is met when a single malfunction in the reactivity 
control system will not cause the SAFDLs to be exceeded. The current 
RWE licensing analysis assumes a control rod drift event without any 
credit for the RBM. With respect to the reactivity control system, 
the assumptions of a control rod drift event and no actuation of the 
RBM are more conservative than the assumptions in the original SSES 
Safety Evaluation. Therefore, the requirements from GDC 25 are still 
met. Therefore, no significant reduction in the safety margin 
exists.

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

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: February 9, 1996, as supplemented March 
15, 1996. This notice supersedes the notice published on February 28, 
1996 (61 FR 7557) in its entirety.
    Description of amendment request: The proposed amendment would 
revise the Administrative Controls Section 5.6.6 of the Ginna Technical 
Specifications (TSs) to incorporate a reference to the methodology for 
determining pressure/temperature (P/T) and low-temperature overpressure 
protection (LTOP) limits. The proposed amendment would follow guidance 
given in Generic Letter 96-03 for relocating LTOP and the reactor 
coolant system (RCS) P/T limits to the RCS Pressure and Temperature 
Limits Report (PTLR). The proposed amendment will allow the licensee to 
perform future LTOP and RCS P/T evaluations, using NRC-approved 
methodology, without requiring changes to the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
changes only require that future RCS P/T and LTOP limits be 
developed using NRC approved methodology as specified within the 
Administrative Controls section and do not involve any technical 
changes. As such, these changes are administrative in nature and do 
not impact initiators or analyzed events or assumed mitigation of 
accident or transient events. Therefore, these changes do not 
involve a significant increase in the probability or consequences of 
an accident previously analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. The proposed changes 
will not impose any new or different requirements. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

[[Page 15995]]

    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes will not reduce a margin of plant 
safety because the changes do not impact any safety analysis 
assumptions other than requiring future evaluations of RCS P/T and 
LTOP limits to be performed in accordance with NRC approved 
methodology. These changes are administrative in nature. As such, no 
question of safety is involved, and the change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005 NRC Acting Project Director: Susan 
Frant Shankman

South Carolina Electric & Gas Company (SCE&G), South Carolina 
Public Service Authority, Docket No. 50-395, Virgil C. Summer 
Nuclear Station, Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: March 19, 1996
    Description of amendment request: The licensee is proposing to 
change the Technical Specification (TS) 3/4.2.4, QUADRANT POWER TILT 
RATIO (QPTR), the Bases for QPTR, and TS 3/4.3.1, REACTOR TRIP SYSTEM 
INSTRUMENTATION, Table 3.3-1, ``Table Notation, Action Statement 2.c.'' 
The licensee is requesting the changes in order to use the guidance in 
the improved Westinghouse Standardized Technical Specifications, NUREG 
1431, Rev. 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 probability or consequences of an accident previously 
evaluated in the FSAR is not significantly increased.
    The QPTR limits ensure that FNdelta-H and FQ(z) 
remain below their limiting values by preventing an undetected 
change in the gross radial power distribution. In MODE 1, the 
FNdelta-H and FQ(z) limits must be maintained to 
preclude core power distributions from exceeding design limits 
assumed in the safety analyses. The QPTR satisfies Criterion 2 of 
the NRC Policy Statement.
    The QPTR limit of 1.02, at which corrective action is required, 
provides a margin of protection for both the departure from nucleate 
boiling ratio and linear heat generation rate contributing to 
excessive power peaks resulting from X-Y plane power tilts. A 
limiting QPTR of 1.02 can be tolerated before the margin for 
uncertainty in FQ(z) and FNdelta-H is possibly 
challenged. With the QPTR exceeding its limit, a power level 
reduction of 3% from RATED THERMAL POWER for each 1% by which the 
QPTR exceeds 1.00 is a conservative tradeoff of total core power 
with peak linear power.
    The Power Range Neutron Flux trip setpoint reduction is not 
required since incore flux measurements are not expected to change 
concurrent with the loss of a Power Range Channel. These setpoints, 
which were previously reduced in order to account for uncertainties, 
will now be monitored and corrected, if necessary, per TS 3.2.4.
    Any change in the QPTR would be detected by requiring a check of 
the QPTR once per 12 hours. If the QPTR indicates an increase, 
THERMAL POWER has to be reduced accordingly. A 12 hour completion 
time is sufficient because any additional change in QPTR would be 
relatively slow.
    The improvement of TS 3/4.2.4 to reflect the improved STS in no 
way impacts the accident analysis of the FSAR. Therefore, the 
probability or consequences of a previously evaluated accident has 
not been increased.
    2. The possibility of an accident or a malfunction of a 
different type than any previously evaluated is not created.
    The proposed amendment request does not necessitate physical 
alteration of the plant nor changes in parameters governing normal 
plant operation. Therefore, the change does not create the 
possibility of a new or different kind of accident or malfunction.
    3. The margin of safety has not been significantly reduced.
    This proposed amendment request precludes core power 
distributions that may lead to violation of the following fuel 
design criteria:
    a. During a large break loss of coolant accident, the peak 
cladding temperature must not exceed 2200 deg.
    b. During a loss of forced reactor coolant flow accident, there 
must be at least 95% probability at the 95% confidence level (the 
95/95 departure from nucleate boiling (DNB) criterion) that the hot 
fuel rod in the core does not experience a DNB condition;
    c. During an ejected rod accident, the energy deposition to the 
fuel must not exceed 280 cal/gm; and
    d. The control rods must be capable of shutting down the reactor 
with a minimum required shutdown margin with the highest worth 
control rod stuck fully withdrawn.
    The improvement of TS 3/4.2.4 ensures that the gross radial 
power distribution remains consistent with the design values used in 
the safety analyses.
    The core peaking factors and the quadrant tilt must be evaluated 
because they are the factors that best characterize the core power 
distribution. This reevaluation is required to ensure that the 
reactor core conditions are consistent with the assumptions in the 
safety analyses. Therefore, the margin of safety has not decreased.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment requests: November 2, 1995
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 3.8.1, ``AC Sources - Operating,'' of the 
improved TS, to (1) extend the offsite circuit allowed outage time 
(AOT) from ``72 hours AND 6 days from discovery of failure to meet 
LCO'' to ``72 hours AND 10 days from discovery of failure to meet LCO'' 
and (2) extend the emergency diesel generator (EDG) AOT from ``72 hours 
AND 6 days from discovery of failure to meet LCO'' to ``7 days AND 10 
days from discovery of failure to meet LCO.'' Additionally, the 
licensee proposes to further extend the EDG AOT to ``10 days AND 10 
days from discovery of failure to meet LCO'' on a once-per-refueling 
cycle frequency for maintenance purposes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Emergency Diesel Generators (EDGs) are backup alternating 
current power sources designed to power essential safety systems in 
the event of a loss of offsite power. EDGs are not accident 
initiators in any accident previously evaluated. Therefore, this 
change does not involve an increase in the probability of an 
accident previously evaluated.
    The EDGs provide backup power to components that mitigate the 
consequences of accidents. The proposed changes to the Allowed 
Outage Times (AOTs) do not affect

[[Page 15996]]
any of the assumptions used in the deterministic safety analysis.
    To fully evaluate the effect of the EDG AOT extension, 
Probabilistic Safety Analysis (PSA) methods were utilized. The 
results of these analyses show no significant increase in the core 
damage frequency. As a result, there would be no significant 
increase in the consequences of accidents previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change does not alter the design, configuration, 
or method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the Limiting Conditions for 
Operation or their Bases that are used in the deterministic analyses 
to establish the margin of safety. PSA evaluations were used to 
evaluate these changes, and these evaluations determined that the 
changes are either risk neutral or risk beneficial.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P.O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

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

    Date of amendment requests: November 6, 1995
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 3.5.1, ``Safety Injection Tanks (SITs),'' 
of the improved TS to extend, in general, the allowed outage time (AOT) 
for a single inoperable SIT from 1 hour to 24 hours. Additionally, the 
licensee proposes to extend the SIT AOT from 1 hour to 72 hours if a 
single SIT becomes inoperable due to malfunctioning SIT water level 
and/or nitrogen cover pressure instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Safety Injection Tanks (SITs) are passive components in the 
Emergency Core Cooling System (ECCS). The SITs are not accident 
initiators in any accident previously evaluated.
    Therefore, this change does not involve an increase in the 
probability of an accident previously evaluated.
    The SITs are designed to mitigate the consequences of Loss of 
Coolant Accidents (LOCAs). The proposed changes do not affect any of 
the assumptions used in deterministic LOCA analysis. Therefore, the 
consequences of accidents previously evaluated do not change.
    To fully evaluate the SIT Allowed Outage Time (AOT) extension, 
Probabilistic Safety Analysis (PSA) methods were utilized. The 
results of these analyses show no significant increase in core 
damage frequency. As a result, there would be no significant 
increase in the consequences of an accident previously evaluated.
    The proposed change pertaining to SIT inoperability based solely 
on instrumentation malfunction does not involve a significant 
increase in the consequences of an accident as evaluated and 
endorsed by the Nuclear Regulatory Commission (NRC) in NUREG-1366, 
``Improvements to Technical Specifications Surveillance 
Requirements.''
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change does not change the design, configuration, 
or method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses 
to establish the margin of safety. PSA evaluations were used to 
evaluate these changes. These evaluations demonstrate that the 
changes are either risk neutral or risk beneficial.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

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

    Date of amendment requests: November 8, 1995
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 3.5.2, ``ECCS - Operating,'' in the 
improved TS to extend the allowed outage time from 72 hours to 7 days 
for a single low pressure safety injection (LPSI) train.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Low Pressure Safety Injection (LPSI) system is a part of the 
Emergency Core Cooling System (ECCS). Inoperable LPSI components are 
not considered to be accident initiators. Therefore, this change 
does not involve an increase in the probability of an accident 
previously evaluated.
    The LPSI system is primarily designed to mitigate the 
consequences of a large Loss of Coolant Accident (LOCA). This 
proposed change does not affect any of the assumptions used in the 
deterministic LOCA analysis. Therefore, the consequences of 
accidents previously evaluated do not change.
    To fully evaluate the LPSI Allowed Outage Time (AOT) extension, 
Probabilistic Safety Analysis (PSA) methods were utilized. The 
results of these analyses show no significant increase in core 
damage frequency. As a result, there would be no significant 
increase in the consequences of an accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

[[Page 15997]]

    This proposed change does not change the design, configuration, 
or method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses 
to establish the margin of safety. PSA evaluations were used to 
evaluate these changes.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

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

    Date of amendment requests: December 6, 1995
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 4.3, ``Fuel Storage,'' of the improved TS, 
to allow fuel assemblies having a maximum U-235 enrichment of 4.8 
weight percent to be stored in both the spent fuel racks and the new 
fuel racks.
    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.
    There is no increase in the probability of an accident because 
the physical characteristics of a fuel assembly are not changed when 
fuel enrichment is increased. No changes will be made to any safety 
related equipment or systems. Fuel assembly movement will continue 
to be controlled by approved fuel handling procedures.
    Fuel cycle designs will continue to be analyzed with Nuclear 
Regulatory Commission (NRC)-approved codes and methods to ensure the 
design bases for San Onofre Units 2 and 3 are satisfied.
    The double contingency principle of American National Standards 
Institute/American Nuclear Society (ANSI/ANS) Standard 8.1-1983 can 
be applied to any postulated accident in the Spent Fuel Pool (SFP) 
which could cause reactivity to increase. In conjunction with 
administrative controls for heavy loads and impact zones, a boron 
concentration of 1850 parts per million (PPM) (the current Technical 
Specification (TS) limit) is sufficient to maintain k-eff less than 
or equal to 0.95 for all normal and postulated accident conditions.
    Regarding the new fuel storage racks, there is no postulated 
accident which could cause reactivity to increase above 0.95 for all 
moderator densities from 0.0 to 1.0 grams/cubic centimeter (gms/cc).
    The radiological consequence analyses performed in the Updated 
Final Safety Analysis Report (UFSAR) include the development of 
source terms which bound discharge fuel burnups to 60,000 megawatt 
days per ton (MWD/T). Increasing the San Onofre Units 2 and 3 
enrichment to 4.8 weight percent (w/o) does not result in discharge 
fuel assembly burnups greater than 60,000 MWD/T. Thus, the 
consequences of the fuel handling accident are unchanged from the 
current UFSAR bases.
    Therefore, this proposed change will 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.
    The proposed changes do not involve any physical changes to the 
plant or any changes to the method in which the plant is operated. 
They do not affect the performance or qualification of safety 
related equipment. Fuel handling accidents were previously 
considered. Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    For the SFP, the NRC acceptance criteria is k-eff less than or 
equal to 0.95 under all normal and accident conditions and including 
uncertainties. For the new fuel storage racks, k-eff must remain 
less than 0.95 if completely flooded with unborated water, and must 
remain below 0.98 in an optimum moderation event. Analyses have been 
performed which demonstrate that these acceptance criteria will 
continue to be met when the enrichment is increased to 4.8 w/o.
    The current UFSAR design bases SFP decay heat loads bound the 
proposed enrichment increase due to the reduced fuel batch size.
    Radiological effects of fuel handling accidents are unchanged by 
this enrichment increase.
    The proposed design of the higher enriched fuel will result in a 
slight weight increase. However, the seismic event is bounded by the 
analyses performed for the rerack project.
    Therefore, there will not be 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: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

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

    Date of amendment requests: January 4, 1996
    Description of amendment requests: The licensee proposes to delete 
License Conditions 2.C(26) and 2.C(27). These license conditions 
require the licensee to implement and maintain a plan for scheduling 
all capital modifications based on an NRC approved Integrated 
Implementation Schedule Program Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change deletes an administrative means of tracking 
and scheduling NRC required plant modifications and license 
commitments. It does not affect the plant configuration nor NRC 
mandated schedules for implementation of modifications. Because the 
deletion of the license condition does not affect the plant 
configuration, no accident analyses are affected; therefore, the 
proposed change does not increase the probability or consequences of 
any previously evaluated accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change will not alter the configuration of the 
plant or its operation; therefore, the proposed change does not 
create a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is administrative and does not affect any 
accident analyses or

[[Page 15998]]
involve any modification to the plant configuration; therefore, the 
proposed change does not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of amendment request: February 28, 1996
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to extend the ice weighing and 
flow channel inspection surveillance frequencies from 9 to 18 months. 
Concurrently, the required total ice bed weight would be increased from 
2,360,875 to 2,403,800 lbs. to account for the anticipated additional 
ice sublimation during the longer interval between weighing and 
inspection.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below.
    1. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The ice condenser system is provided to absorb thermal energy 
release following a LOCA or high energy line break (HELB) and to limit 
the peak pressure inside containment. The containment analysis for 
Watts Bar is based on a minimum of 1093 lbs of ice per ice basket 
evenly distributed throughout the ice condenser, and the subcompartment 
analysis is based on 85 percent of the available flow area (flow 
channels) being open uniformly throughout the ice condenser. For the 
predicted sublimation rate of up to 12 percent for 18 months, an 
average ice basket weight of 1093 lbs at the end of the 18 month period 
would still be available. An evaluation of the operating history of the 
other operating ice condenser plants shows that after 18 months 85 
percent of the flow channels will still be available.
    Thus the ice condenser will perform its design functions with the 
revised minimum ice weight and inspection interval. There will be no 
design change or other operational changes. Accordingly, the proposed 
changes to the technical specifications do not affect the probability 
or consequences of an accident.
    2.
    The changes do not create the possibility of a new or different 
kind of accident from any previously analyzed.
    As stated above, the proposed changes do not involve modifications 
to the ice condenser or other plant systems. Hence there is no 
possibility of a new or different kind of accident since no new design 
is involved.
    3. The changes do not involve a significant reduction in a margin 
of safety.
    Plant safety margins are established through limiting conditions of 
operation, limiting safety system settings, and safety limits specified 
in the TS. None of these will be changed.
    Based on this analysis, 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

Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear 
Plant,Unit 1, Rhea County, Tennessee

    Date of amendment request: February 28, 1996
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) surveillance frequency for 
Westinghouse type AR relays, used as solid state protection system 
slave relays or auxiliary relays, from quarterly to a refueling outage 
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) 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.
    This change to the Technical Specifications does not result in a 
condition where the design, material, and construction standards 
that were applicable prior to the change are altered. The same ESFAS 
instrumentation is being used and the same ESFAS system reliability 
is expected. The proposed change will not modify any system 
interface or function and could not increase the likelihood of an 
accident since these events are independent of this change. The 
proposed activity will not change, degrade or prevent the 
performance of any accident mitigation systems or alter any 
assumptions previously made in evaluating the radiological 
consequences of an accident described in the safety analysis report. 
Therefore, the proposed amendment does not result in any increase in 
the probability or consequences of an accident previously evaluated.
    (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 previously evaluated.
    This change does not alter the performance of the ESFAS 
mitigation systems assumed in the plant safety analysis. Changing 
the interval for periodically verifying ESFAS slave relays (assuring 
equipment operability) will not create any new accident initiators 
or scenarios. Implementation of the proposed amendment does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    This change does not affect the total ESFAS system response 
assumed in the safety analysis. The periodic slave relay functional 
verification is relaxed because of the demonstrated high reliability 
of the relay and its insensitivity to any short term wear or aging 
effects. Implementation of the proposed amendment does not result in 
a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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

[[Page 15999]]

    NRC Project Director: Frederick J. Hebdon

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: March 12, 1996
    Brief description of amendments: The proposed amendments would 
revise Technical Specification (TS) 3/4.6.1.1, ``Containment 
Integrity,'' 3/4.6.1.2, ``Containment Leakage,'' 3/4.6.1.3, 
``Containment Air Locks,'' and 3/4.6.1.6, ``Containment Structural 
Integrity,'' and add new TS 6.8.3g, ``Containment Leakage Rate Testing 
Program,'' to implement the new performance-based leakage rate testing 
program as permitted by 10 CFR Part 50, Appendix J.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to the TS and the addition of specification 
6.8.3g to implement the new performance based Containment Leakage 
Rate Testing Program, have no effect on plant operation. The 
proposed changes only provide mechanisms within the TS for 
implementing a performance based methodology for determining the 
frequency of leak rate testing which has been approved by the 
Commission. The test type and test method used for testing would not 
be changed. The test acceptance criteria would not be changed and 
containment leakage will continue to be maintained within the 
required limits.
    Directly referencing the Containment Leakage Rate Testing 
Program for containment [integrated leak rate test] ILRT and [local 
leak rate test] LLRT requirements does not involve any modification 
to plant equipment or affect the operation or design basis of the 
containment. Leakage rate testing is not a precursor to or an 
initiating event for any accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    The proposed changes only allow for the implementation of Option 
B testing frequencies and do not involve any modifications to any 
plant equipment or affect the operation or design basis of the 
containment. The proposed changes do not affect the response of the 
containment during a design basis accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed changes do not adversely affect a Safety Limit, 
Limiting Condition for Operation (LCO) or plant operations. These 
changes only implement the allowed Option B testing frequencies that 
have been determined by the Commission not to involve a safety 
concern. The testing method, acceptance criteria and bases are not 
changed and still provide assurance that the containment will 
provide its intended function.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: William D. Beckner

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: March 21, 1996
    Description of amendment request: The proposed changes clarify the 
requirements for testing the charcoal adsorbent in the auxiliary 
ventilation and control room air filtration systems as outlined in 
Technical Specifications 4.12 and 4.20, respectively.
    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 probability or consequences of an accident previously 
evaluated is not significantly increased.

    The charcoal testing clarifications and explict reference to the 
testing currently conducted do not affect system operation or 
performance, nor do they affect the probability of any event 
initiators. The changes do not affect any Engineered Safety Features 
actuation setpoints or accident mitigation capabilities. Therefore, 
the proposed changes do not significantly increase the consequences 
of an accident or malfunction of equipment important to safety 
previously evaluated in the UFSAR [Updated Final Safety Analysis 
Report].

    2. The possibility of an accident or a malfunction of a different 
type than any previously evaluated is not created.

    The clarification to the charcoal sample testing protocol does 
not affect the method of operation of the system. The proposed 
changes clarify and explicitly identify the testing methodology for 
the charcoal samples. No new or different accident scenarios, 
transient precursors, failure mechanisms, or limiting single 
failures are introduced as a result of these changes. Therefore, the 
possibility of a new or different kind of accident other than those 
already evaluated is not created by this change.

    3. The margin of safety has not been significantly reduced.
    The charcoal adsorber sample laboratory testing accurately 
demonstrates the required performance of the adsorbers following a 
design basis LOCA [loss-of-coolant accident] or Fuel Handling Accident. 
Changing the Technical Specifications to clarify the actual test 
methodology and explicitly [referencing] the charcoal testing actually 
performed does not affect system performance or operation. Therefore, 
these changes do not result in a significant reduction in any margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Eugene V. Imbro

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: February 19, 1996
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) Section 4.2 and its associated basis by allowing the application 
of a voltage-based repair limit for the steam generator tube support 
plate intersections experiencing

[[Page 16000]]
outside diameter stress corrosion cracking. The proposed repair 
criteria are based on guidance provided in Generic Letter 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
affected by Outside Diameter Stress Corrosion Cracking,'' dated August 
14, 1995, and on associated industry guidance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Testing of model boiler specimens for free span tubing (no TSP 
[tube support plate] restraint) at room temperature conditions show 
burst pressures in excess of 5,000 psig for indications of ODSCC 
[outside diameter stress corrosion cracking] with voltage 
measurements as high as 19 volts. Burst testing performed on five 
intersections pulled from the Kewaunee SGs [steam generators] with 
up to a 2 volt indication showed measured tube burst in the range of 
9,537 to 9,756 psig. Burst testing performed on pulled tubes from 
other plants with up to 7.5 volt indications show burst pressures in 
excess of 6,300 psi at room temperatures. Correcting for the effects 
of temperature on material properties and the minimum strength 
levels, tube burst capability significantly exceeds the safety 
factor requirements of RG [Regulatory Guide] 1.121.
    Tube burst criteria are inherently satisfied during normal 
operating conditions due to the presence of the TSP. Test data 
indicates that tube burst cannot occur within the TSP, even for 
tubes with through wall EDM [electro-discharge machining] notches 
0.75 inch long, when the notch is adjacent to the TSP. Since tube 
burst is precluded during normal operating conditions, the criterion 
that must be satisfied to demonstrate adequate tube integrity is a 
safety margin of 1.43 times MSLB [main steam line break] pressure 
differential. The BOC [beginning of cycle] structural limit for 7/8 
inch diameter tubing is 8.82 volts. Applying an allowance of 20.5% 
for NDE [nondestructive examination] uncertainty and 50% for crack 
growth rate over an operating cycle results in a voltage repair 
limit of 5.4 volts. The proposed repair limit of 2 volts is very 
conservative when compared to the 5.4 volts taking into account the 
low average growth rates experienced at Kewaunee and the high tube 
burst pressures.
    Relative to the expected leakage during accident condition 
loadings, a plant specific calculation was performed to determine 
the maximum primary-to-secondary leakage during a postulated MSLB 
event. The evaluation considered both pre-accident and accident 
initiated iodine spikes. The results of the evaluation show that the 
accident spike yielded the limiting leak rate. This case was based 
on a 30 rem thyroid dose at the site boundary and initial primary 
and secondary coolant activity levels of 1.0 uCi/gm and 0.1 uCi/gm 
dose equivalent iodine-131, respectively. A leak rate of 34.0 gpm 
was determined to be the upper limit for allowable primary to 
secondary leakage in the SG in the faulted loop. The SG in the 
intact loop was assumed to leak at a rate of 0.1 gpm (150 gpd).
    Application of the voltage-based repair limit will be 
supplemented with a projected EOC [end of cycle] MSLB leakage 
calculation and conditional burst probability assessment. The 
methodology for performing these calculations will be in accordance 
with the GL [generic letter]. Should the projected MSLB leakage be 
exceeded indications will be repaired or removed from service until 
the projected leakage is less than or equal to 34.0 gpm.
    Application of the voltage-based repair limit will not adversely 
affect SG tube integrity. Therefore, the proposed amendment will not 
increase the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    Implementation of the proposed voltage-based repair limit will 
not reduce the overall safety or functional requirements of the SG 
tube bundles. The tube burst criteria will be satisfied during 
normal operating conditions by the presence of the TSPs. The RG 
1.121 criteria that must be satisfied during accident loading 
conditions is 1.43 times MSLB differential pressure. Conservatively, 
the existing data base of burst testing shows that the tube burst 
margins can be satisfied with bobbin coil signal amplitudes of about 
8.82 volts or less regardless of the depth of tube wall penetration.
    The proposed repair criteria will be supplemented with a reduced 
operating leakage requirement of 150 gpd through either SG to 
preclude the potential for excessive leakage during operating 
conditions. The 150 gpd restriction will provide for timely leakage 
detection and plant shutdown in the event of the occurrence of an 
unexpected single crack resulting in leakage that is associated with 
the longest permissible crack length. The operating leakage limit is 
based on leak-before break considerations, critical crack length and 
predicted leakage.
    The SG tube integrity will continue to be maintained through 
inservice inspections and primary-to-secondary leakage monitoring. 
Therefore, the proposed change will not create the possibility of a 
new or different kind or accident.
    3. Involve a significant reduction in the margin of safety.
    Application of the voltage-based repair criteria has been 
demonstrated to maintain tube integrity commensurate with the RG 
1.121 criteria. RG 1.121 describes a method acceptable to the staff 
for meeting GDCs [general design criteria] 2, 14, 15, 31 and 32. 
This is accomplished by determining the limiting degradation of SG 
tubing as established by inservice inspection, beyond which tubes 
should be removed from service. Upon implementation of the repair 
criteria, even under the worst case conditions, the occurrence of 
ODSCC at the TSPs is not expected to lead to a SG tube rupture event 
during normal or faulted conditions. The most limiting event would 
be a potential increase in leakage during a MSLB event. Excessive 
leakage during a MSLB is precluded by verifying that the expected 
EOC crack distribution of ODSCC indications at TSP locations would 
result in an acceptably low primary-to-secondary leakage. Therefore, 
the radiological consequences from tubes remaining in service is a 
small fraction of the 10 CFR 100 limits.
    The combined effects of a LOCA [loss-of-coolant accident] plus 
SSE [safe shutdown earthquake] on the SGs were assessed as required 
by GDC 2. This issue was addressed for the Kewaunee SGs through the 
application of leak-before-break (LBB) principles to the primary 
loop piping. Based on the results of this analysis, it is concluded 
that the LBB is applicable to the Kewaunee primary loops and, thus, 
the probability of breaks in the primary loop piping is sufficiently 
low that they need not be considered in the structural design basis 
of the plant. Excluding breaks in the primary loops, the LOCA loads 
from the large branch lines were also assessed and found to be of 
insufficient magnitude to result in SG tube collapse. Based on these 
analysis results, no tubes are expected to collapse or deform to the 
degree that the secondary-to-primary in-leakage would be increased 
over currently expected levels. On this basis no tubes need to be 
excluded from the voltage-based repair criteria for reasons of 
deformation resulting from combined LOCA or SSE loadings.
    Addressing the RG 1.83 considerations, implementation of the 
voltage-based repair criteria will include a 100% bobbin coil probe 
inspection of all tube-to-TSP intersections with known ODSCC down to 
the lowest cold leg TSP identified. This will be supplemented by a 
reduced operating leakage limit, enhanced eddy current data analysis 
guidelines, MRPC [motorized rotating pancake coil] inspection 
requirements and a projected EOC voltage distribution. It is 
concluded that the proposed change will not result in 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: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus
    
[[Page 16001]]


Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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: January 16, 1996
    Brief description of amendments: The amendments revise the 
Technical Specifications to reflect approval of the use of 10 CFR Part 
50, Appendix J, Option B, for the Calvert Cliffs Nuclear Power Plant, 
Unit Nos. 1 and 2, containment leakage rate test program for Type A 
tests only.
    Date of issuance: March 13, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 212 and 189
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5810) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 13, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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: November 30, 1995, as 
supplemented by letter dated March 15, 1996.
    Brief description of amendments: The amendments allow the 
installation of tube sleeves as an alternative to plugging for repair 
of steam generator (SG) tubes using repair techniques developed by 
Westinghouse Electric Corporation. The November 30, 1995, letter also 
requested approval of repair techniques developed by ABB Combustion 
Engineering, Inc., for repairing SG tubes. The NRC staff is still 
reviewing that portion of the request and will notice the results of 
its review at a future date.
    Date of issuance: March 22, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 213 and 190
    Facility Operating License No. DPR-53 and DPR-69: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
176) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 22, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: November 22, 1995
    Brief description of amendment: The proposed change will delete the 
qualifying statement, ''... provided the remaining systems are in 
continuous operation,'' from TS Section 3.3.4.2.
    Date of issuance: March 15, 1996
    Effective date: March 15, 1996
    Amendment No. 168
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, l995 (60 FR 
62487) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 15, 1996No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: July 18, 1994, as supplemented 
by letters dated October 9, 1995, February 13 and March 8, 1996
    Brief description of amendments: The amendments revise the current 
combined Technical Specifications (TS) for Units 1 and 2 by separating 
them into individual volumes for Unit 1 and Unit 2. In addition to the 
changes required by the TS split, some administrative and editorial 
changes were made, such as the correction of typographical errors and 
the deletion of unnecessary blank pages.
    Date of issuance: March 21, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: Unit 1 - 166 - Unit 2 - 148
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47166) The October 9, 1995, February 13 and March 8, 1996, letters 
provided additional information that did not change the scope of the 
July 18, 1994, application and the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
March 21, 1996 and Environmental Assessment dated February 7, 1996. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

[[Page 16002]]


Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of application for amendments: December 15, 1995, as 
supplemented March 5, 1996
    Brief description of amendments: These amendments (1) revise 
Technical Specifications (TSs) 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, 3/
4.6.1.6, and associated Bases, (2) delete TS 6.9.2.g, and (3) add a new 
TS 6.17. These changes make the TSs consistent with Option B of 
Appendix J of 10 CFR Part 50 and the implementing guidance of 
Regulatory Guide 1.163, ``Performance-Based Containment Leak Test 
Program,'' dated September 1995. Option B of Appendix J permits 
implementation of a performance-based leak rate test schedule in lieu 
of the prescriptive requirements contained in Option A of Appendix J. 
These amendments remove from the TSs the prescriptive requirements of 
Option A concerning test frequencies and test methodology. These 
amendments also include minor administrative and editorial changes to 
add consistency between the Bases and the TSs and provide additional 
clarification.
    Date of issuance: March 19, 1996
    Effective date: Both units, as of the date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 197 and 80
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
179) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 19, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Florida Power and Light Company, et al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: August 16, 1995
    Brief description of amendment: This amendment modifies Technical 
Specification 3.6.6.1, Shield Building Ventilation System (SBVS), to 
more effectively address the design functions performed by the SBVS for 
both the Shield Building and the Fuel Handling Building.
    Date of issuance: March 20, 1996
    Effective date: March 20, 1996
    Amendment No.: 81
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49937) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 20, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mill 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: August 10, 1995, as supplemented 
on December 21, 1995, and February 22, 1996
    Brief description of amendment: The amendment deletes a Technical 
Specification (TS) reference to the reactor trip input to the reactor 
building isolation system, changes the surveillance frequency for the 
sodium hydroxide storage tank and station battery, and removes an 
inappropriate reference in the TS bases section to testing that is not 
required by the TSs themselves.
    Date of issuance: March 21, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 200
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58401). The December 21, 1995 and February 22, 1996 letters did not 
change the staff's determination hazards consideration exist. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 21, 1996No significant hazards 
consideration comments received: No
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105

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: December 19, 1995 and 
supplemented February 16, 1996 (AEP:NRC:1215B&D)
    Brief description of amendments: The amendments modify the 
technical specifications to replace the existing scheduling 
requirements for overall integrated and local containment leakage rate 
testing with a requirement to perform the testing in accordance with 10 
CFR Part 50, Appendix J, Option B. Option B allows test scheduling to 
be adjusted based on past performance.
    Date of issuance: March 19, 1996
    Effective date: March 19, 1996, with full implementation within 45 
days
    Amendment Nos.: Unit 1 - 209, Unit 2 -193
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1632) The February 16, 1996 supplement made only a minor change to the 
proposed technical specifications that provided consistency between the 
wording for Units 1 and 2. The change did not affect the staff's 
proposed finding that the amendments involve no significant hazards 
consideration.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 19, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Maud Preston Palenske 
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: February 17, 1996
    Brief description of amendment: The amendment allows one main steam 
line's leakage rate to be as high as 35 standard cubic feet per hour 
(scfh) as long as the total leakage through all four main steam lines 
does not exceed 100 scfh until the end of Operating Cycle 6.
    Date of issuance: March 18, 1996
    Effective date: March 18, 1996
    Amendment No.: 83
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications. The Commission's related evaluation of the 
amendment and final no significant hazards consideration determination 
is contained in a Safety Evaluation dated March 18, 1996. Public 
comments requested as to proposed no significant hazards consideration: 
Yes (61 FR 7823 dated February 29, 1996). That notice provided an 
opportunity to submit comments on the Commission's

[[Page 16003]]
proposed no significant hazards consideration determination. No 
comments have been received. The notice also provided for an 
opportunity to request a hearing by April 1, 1996, 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.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: December 19, 1994 (TXX-94274), as 
supplemented by letter dated January 25, 1996 (TXX-96026)
    Brief description of amendments: These changes allowed testing of 
Reactor Protection System and Engineered Safety Features Actuation 
System instrument channels with the channel under test in bypass in 
order to reduce the vulnerability to spurious trips during surveillance 
testing.
    Date of issuance: March 14, 1996
    Effective date: March 14, 1996
    Amendment Nos.: Unit 1 - 47; Unit 2 - 33
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6312) The additional information contained in the supplemental letter 
dated January 25, 1996, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated March 14, 1996. No significant hazards consideration 
comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment requests: November 21, 1995 (TXX-95289), as 
supplemented by letters dated February 22 (TXX-96061 and TXX-96062) and 
28, (TXX-96068), and March 13, 1996 (TXX-96090).
    Brief description of amendments: The amendments allowed both doors 
of the containment personnel airlock to be open during fuel movement 
and core alterations, providing one airlock door is capable of being 
closed and the water level in the refueling pool is maintained.
    Date of issuance: March 18, 1996
    Effective date: March 18, 1996
    Amendment Nos.: Unit 1 - 48; Unit 2 - 34
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
185) The additional information contained in the supplemental letters 
dated February 22 (2 letters) and 28, and March 13, 1996, were 
clarifying in nature and thus, within the scope of the initial notice 
and did not affect the staff's proposed no significant hazards 
consideration determination.The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated March 18, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of application for amendments: May 26, 1994, as supplemented 
January 5, April 25 and October 12, 1995, and February 2 and March 1, 
1996.
    Brief description of amendments: These amendments revise the 
Technical Specifications by extending the operation of both units with 
the current heatup and cooldown limit curves to 23.6 effective full 
power years. The basis for TS Section 15.3.1.B, ``Pressure/Temperature 
Limits,'' is also revised to reflect the methodology for the curve 
compilation.
    Date of issuance: March 20, 1996
    Effective date: March 20, 1995
    Amendment Nos.: 168 and 172
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37093). The supplemental submittals provided additional information 
that did not change the initial proposed no significant hazards 
consideration determination.The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated March 20, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been

[[Page 16004]]
issued without opportunity for comment. If there has been some time for 
public comment but less than 30 days, the Commission may provide an 
opportunity for public comment. If comments have been requested, it is 
so stated. In either event, the State has been consulted by telephone 
whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By May 10, 1996, 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be

[[Page 16005]]
granted based upon a balancing of the factors specified in 10 CFR 
2.714(a)(1)(i)-(v) and 2.714(d).

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

    Date of application for amendment: March 23, 1996
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) 4.8.2.1.c, ``DC Sources - Operating,'' to specify 
that the provisions of TS 4.0.1 and 4.0.4 are not applicable. This 
provision expires upon entry into Mode 4 coming out of the sixth 
refueling outage or upon any deep discharge cycle of the battery.
    Date of issuance: March 23, 1996
    Effective date: March 23, 1996
    Amendment No.: Unit 2 - 94
    Facility Operating License No. NPF-51: The amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No.The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated March 23, 1996.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq. Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

    Date of application for amendment: March 26, 1996
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3/4.9.6 to allow the refueling machine overload 
cutoff limit to be increased to as much as 2000 pounds, from the 
current 1600 pound limit, in an effort to free the stuck fuel assembly 
from core location A-06. The additional 400 pound increase will be 
applied in 50 pound increments. This change will expire when the fuel 
assembly located at core location A-06 is successfully withdrawn.
    Date of issuance: March 26, 1996
    Effective date: March 26, 1996, to be implemented prior to entry 
into Mode 4 from the current refueling outage.
    Amendment No.: Unit 2 - 95
    Facility Operating License No. NPF-51: The amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No.The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated March 26, 1996.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq. Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

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

    Date of application for amendment: March 29, 1996
    Brief description of amendment: The amendment clarifies the testing 
requirements and updates the regulatory and industry guidance 
references for charcoal adsorber units addressed by TS 4.6.4.4, 
Hydrogen Purge System; TS 4.6.5.1, Emergency Ventilation System; and TS 
4.7.6.1, Control Room Emergency Ventilation System.
    Date of issuance: March 29, 1996
    Effective date: March 29, 1996
    Amendment No.: 209
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No.The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated March 29, 1996.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

    Dated at Rockville, Maryland, this 3rd day of April 1996.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation.
[FR Doc. 96-8786 Filed 4-9-96; 8:45 am]
BILLING CODE 7590-01-F