[Federal Register Volume 63, Number 250 (Wednesday, December 30, 1998)]
[Notices]
[Pages 71962-71984]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-34440]


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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 December 7, 1998, through December 17, 1998. 
The last biweekly notice was published on December 16, 1998 (63 FR 
69332).

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

[[Page 71963]]

different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By January 29, 1999, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

[[Page 71964]]

    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: October 27, 1998.
    Description of amendment request: The Carolina Power & Light 
Company, licensee for the Brunswick Steam Electric Plant (BSEP), Unit 
Nos. 1 and 2, proposed amendments to the Operating Licenses for the 
BSEP units. The amendments are administrative in nature and would 
delete various completed license conditions, make editorial changes, 
and provide clarifying information.
    The licensee has concluded that the proposed license amendments do 
not involve a Significant Hazards Consideration. In support of this 
determination, an evaluation of each of the three standards set forth 
in 10 CFR 50.92 is provided below.
    Basis for a 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 license amendments do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    The proposed changes revise the BSEP, Unit Nos. 1 and 2, Facility 
Operating Licenses to delete various license conditions that have been 
completed, make editorial changes, and provide clarifying information. 
The changes are administrative and only provide updated and clarifying 
information. No physical or operational changes to the facility will 
result from the proposed changes. Therefore, the proposed license 
amendments do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed license amendments will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes revise the BSEP, Unit Nos. 1 and 2, Facility 
Operating Licenses to delete various license conditions that have been 
completed, make editorial changes, and provide clarifying information. 
The changes are administrative and only provide updated and clarifying 
information. The proposed license amendments do not alter any plant 
operation and will not result in a physical change to the facility. 
Therefore, the proposed license amendments do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed changes revise the BSEP, Unit Nos. 1 and 2, Facility 
Operating Licenses to delete various license conditions that have been 
completed, make editorial changes, and provide clarifying information. 
The changes are administrative and only provide updated and clarifying 
information. No physical or operational changes to the facility will 
result from the proposed changes. Therefore, the proposed license 
amendments do not involve a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of application for amendment request: November 30, 1998.
    Description of amendment request: This amendment request proposes 
to relocate, to a licensee controlled document, the requirement for 
removal of the Reactor Protection System (RPS) shorting links. Removal 
of the shorting links enables a non-coincident scram on high neutron 
flux as detected by the Source Range Monitors (SRMs).
    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:
    Does the change involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The RPS shorting links are not precursors to any previously 
evaluated accident. The Source Range Monitors (SRMs), and the ability 
of the SRMs to provide a RPS trip, are also not precursors to any 
previously evaluated accident. Therefore, relocating the RPS shorting 
link requirement to administrative controls [the Updated Final Safety 
Analysis Report (UFSAR)] will not increase the probability of an 
accident previously evaluated.
    The RPS shorting links are not assumed to be removed in any 
accident analysis, and the SRMs are not assumed to provide a RPS trip 
in any accident analysis. The refueling interlocks and SHUTDOWN MARGIN 
calculations will continue to provide assurance of reactivity control. 
Therefore, relocating the RPS shorting link requirements to 
administrative controls [the UFSAR] will not increase the consequences 
of an accident previously evaluated.
    The RPS shorting link requirements will be relocated to 
administrative controls that are administered pursuant to the 
requirements of 10 CFR 50.59, thereby reducing the level of regulatory 
control. The level of regulatory control has no impact on the 
probability or consequences of an accident previously evaluated.
    Consequently, this proposed amendment does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Relocating the RPS shorting link requirements to administrative 
controls [the UFSAR] does not create any new failure mechanisms. No new 
equipment will be installed or utilized, and no new operating 
conditions will be initiated as a result of this change. Therefore, the 
proposed change does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The refuel interlocks and SHUTDOWN MARGIN calculations will 
continue to ensure that the reactor stays

[[Page 71965]]

subcritical in the Refuel Mode. The margin to safety as represented by 
the SHUTDOWN MARGIN designed into the core and verified in the SHUTDOWN 
MARGIN calculations will be unaffected by relocation of the RPS 
shorting link requirements to administrative controls [the UFSAR]. The 
margin to safety as represented by the fuel bundle drop assumptions 
protected by the refuel interlocks will be unaffected. In addition, no 
accident analysis assumes that the RPS shorting links are removed. In 
addition, the RPS shorting link requirements will be relocated to 
administrative controls [the UFSAR] for which future change will be 
evaluated pursuant to the requirements of 10 CFR 50.59. Therefore, 
there will be no change in the types or significant increase in the 
amounts of any effluents released offsite, and, thus, these 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 
amendments requested involve no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: October 30, 1998 (LAR-236).
    Description of amendment request: The proposed amendment would 
change the Crystal River Unit 3 (CR-3) Improved Technical 
Specifications (ITS) Section 5.6.2.19, Section 3.4.11, Bases 3.4.11 and 
Bases 3.4.3. The changes reflect the use of fluence methodology 
described in Topical Report BAW-2241P, ``Fluence and Uncertainty 
Methodologies,'' and the use of American Society of Mechanical 
Engineers (ASME) Code Case N-514, ``Low Temperature Overpressure 
Protection,'' for developing Low Temperature Overpressure Protection 
(LTOP) limits. Reference to Topical Report BAW-1543A, ``Integrated 
Reactor Vessel Surveillance Program,'' was also added to ITS Section 
5.6.2.19. ITS Section 3.4.11 (Low Temperature Overpressure Protection 
System), was revised to reflect the new LTOP limits based on revised 
fluence projections through 32 Effective Full Power Years (EFPY). The 
Pressure/Temperature (P/T) Limits Report is being revised to reflect 
the new P/T limits for heatup, cooldown, hydrostatic and leak test, and 
to incorporate the CR-3 LTOP curve.
    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.
    LAR [License Amendment Request] #236 proposes several changes to 
the ITS operational limits. These changes are being proposed to 
maintain the necessary margins of safety through 32 EFPY using analyses 
based on methodologies that have been previously approved for use at 
CR-3, ASME Code Case N-514 and LTOP SER [Safety Evaluation Report], and 
are currently being reviewed by the NRC staff:

--NRC to FPC letter, 3N1293-30, dated December 20, 1993, ``Crystal 
River Unit 3--Issuance of Amendment RE: Improved Technical 
Specifications (TAC No. M74563)''
--NRC to FPC letter, 3N1297-16, dated December 22, 1997, ``Crystal 
River Unit 3--Staff Evaluation and Issuance of Amendment RE: Low-
Temperature Overpressure Protection (TAC No. M99277)''
--NRC to FPC letter, 3N079705, dated July 3, 1997, ``Crystal River 3--
Exemption from Requirements of 10 CFR 50.60, Acceptance Criteria for 
Fracture Prevention for Lightwater Nuclear Power Reactors for Normal 
Operation (TAC No. M98380)''
--BAW-2241P, ``Fluence and Uncertainty Methodologies''

    The limiting transient for LTOP remains a failed-open makeup valve. 
Existing LTOP controls (maximum of one makeup pump capable of injecting 
into the RCS [reactor coolant system], high pressure injection (HPI) 
deactivated, the CFTs [core flood tanks] isolated, pressure relief 
capability and maintaining a gas volume in the RCS) remain unchanged 
from the current ITS 3.4.11 as approved by Reference 3, except the 
setpoints proposed herein. The setpoints are being updated to reflect 
the new 32 EFPY fluence analysis and P/T limits. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not create the possibility of a new or 
different kind of accident from any previously evaluated since they do 
not introduce new systems, failure modes or plant perturbations. 
Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will not involve a significant reduction in 
the margin of safety since the proposed P/T limitations have been 
developed consistent with the requirements of 10 CFR 50.60. The 
operational limits have been developed to maintain the necessary 
margins of safety as defined by ASME through 32 EFPY using 
methodologies previously reviewed and approved by the NRC. The 
objective of these limits is to prevent non-ductile failure during any 
normal operating condition, including anticipated operational 
occurrences and system hydrostatic tests.
    The LTOP safety factors are based on reanalyzed conditions for 32 
EFPY of operation utilizing methodology contained in ASME Code Case N-
514 which has been approved for use at CR-3. The Code Case provides an 
acceptable margin of safety against flaw initiation and reactor vessel 
failure. The application of Code Case N-514 for CR-3 ensures an 
acceptable level of safety. 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 
request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P. O. Box 14024, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

[[Page 71966]]

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

    Date of amendment request: October 30, 1998.
    Description of amendment request: The proposed amendment requests 
approval of a change to the Crystal River Unit 3 (CR-3) Final Safety 
Analysis Report (FSAR) regarding the methodology for performing the 
Spent Fuel Pool (SFP) B criticality analysis. Recent Boraflex samples 
from the SFP B demonstrate a weight loss in excess of the available 
margin within the current licensing basis calculation. The criticality 
analysis calculations proposed in this amendment request demonstrate 
that the burnup/enrichment curves in the current Improved Technical 
Specifications (ITS) have sufficient margin to accommodate up to a 20% 
loss in Boraflex neutron absorption, and still maintain SFP B at less 
than or equal to 0.95 k-effective when fully loaded and flooded with 
unborated water. Florida Power Corporation has concluded that the 
change in the criticality analysis methodology represents an unreviewed 
safety question, and thus requires prior NRC approval.
    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.
    No. The two possible accidents are: (1) criticality during normal 
storage and (2) criticality due to a misloaded fuel assembly during 
handling fuel. Each are discussed below:
    (1) Criticality during normal storage.
    For criticality during normal storage to occur, there must be a 
loss of negative reactivity since an addition of positive reactivity is 
not possible without fuel movement. A loss in negative reactivity could 
result only from reduction in Boraflex inventory below that needed to 
meet the design basis. The proposed criticality analysis for Spent Fuel 
Pool B demonstrates that Spent Fuel Pool B is capable of maintaining 
the design basis requirement of k-effective less than or equal to 0.95 
when flooded with unborated water and with a loss of up to 20% of the 
Boraflex absorber material. Therefore, allowing up to 20% Boraflex loss 
with the new analysis does not significantly increase the probability 
of an accident previously evaluated.
    (2) Criticality during fuel handling.
    Criticality during fuel handling could occur due to loss of 
negative reactivity, or the addition of positive reactivity. Loss of 
negative reactivity could result from loss of Boraflex as discussed 
above.
    Addition of positive reactivity would result from the misloading of 
fuel in a fashion not in accordance with ITS LCO 3.7.15, such as the 
misloading of a fresh 5.05% enriched fuel assembly into Region 2 or 
side-by-side with another fresh fuel assembly in Region 1. The minimum 
required boron concentration of ITS LCO 3.7.14 and CR-3 FSAR 9.3.2.1.2 
are intended to compensate for just such an accident. Consistent with 
the double-contingency principle, a boron dilution is not required to 
be considered concurrent with a misloaded new fuel assembly (bases of 
ITS LCO 3.7.14). The use of a new calculational method will not 
increase the probability of fuel assembly misloading. A boron dilution 
event without an accompanying misloaded fuel assembly is not impacted 
by the new criticality analysis, since the design basis allows for 
unborated water for normal storage conditions.
    Therefore, since the proposed criticality analysis does not 
increase the probability of a misloaded fuel assembly, the probability 
of an occurrence of an accident previously evaluated is not 
significantly increased.
    Boraflex is credited with preventing inadvertent criticality. It is 
not credited with mitigating the effects, or dose consequences, to the 
public or to plant personnel from an inadvertent criticality. The 
criticality analysis does not affect or mitigate the dose consequences 
to the public or plant personnel from an inadvertent criticality.
    There are no other SAR accidents that could be affected. Therefore, 
the use of the proposed criticality analysis, does not significantly 
increase the consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No. The only purpose, or function, of Boraflex is reactivity 
control. Therefore, the use of the proposed criticality analysis can 
only result in reactivity related accidents, such as an inadvertent 
criticality. Though a spent fuel pool criticality accident is not 
discussed in detail, a calculation to ensure such an accident could not 
occur is referenced by both FSAR 9.3 and 9.6. Therefore, this is an 
accident already discussed by the SAR and dependence on a new 
criticality analysis does not create the possibility of an accident of 
a new or different kind than any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    No. The proposed analysis demonstrates that the safety function and 
design basis are met even for a Boraflex loss of up to 20%. Though the 
proposed criticality analysis methodology is more realistic, and has 
been licensed at other sites, it is less conservative than the 
existing, NRC approved analysis that is currently part of the CR-3 
licensing basis. Additionally, it permits operation with a greater loss 
of Boraflex than the existing analysis.
    The current licensing basis, BAW-2209, ``Crystal River Unit 3 Spent 
Fuel Storage Pool Criticality Analysis'', provides the analytical basis 
of both ITS LCO 3.7.14 and LCO 3.7.15. This analysis uses very 
conservative assumptions and methodologies, and results in very little 
margin remaining for identified Boraflex loss. The margin of safety, 
although less than previously evaluated, is not significantly reduced 
with reliance on the current criticality analysis. The margin of safety 
is restored with use of the proposed criticality analysis. Therefore, 
the margin of safety is not significantly reduced with use of the 
proposed criticality analysis.
    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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: November 23, 1998.
    Description of amendment request: The proposed amendment would 
change the CR-3 Improved Technical Specifications (ITS) to raise the 
Engineered Safeguards Actuation System (ESAS) setpoint for reactor 
coolant system (RCS) low pressure from

[[Page 71967]]

1500 psig to 1625 psig. This change is intended to provide for earlier 
actuation of high pressure injection (HPI) following certain small 
break loss of coolant accidents and result in a lower peak center line 
temperature (PCT) during these transients. The applicability 
requirement for ESAS operability would be changed from greater than 
1700 psig to greater than 1800 psig to maintain the previous margin 
above the ESAS setpoint. Similarly, the reactor protection system (RPS) 
setpoint for RCS low pressure and the RPS setpoint for Shutdown Bypass 
(RCS High Pressure) would each be raised by 100 psig to maintain the 
previous pressure margins. In addition, Surveillance Requirement 
3.5.2.5 would be revised such that valves in the HPI flowpath that are 
throttled to balance flow between the four HPI lines would be verified 
in the correct position. The need for these changes resulted from 
planned modifications to the HPI system to improve performance and 
reliability of this system. Changes to ITS Bases necessitated by the 
system modifications and setpoint changes are included in the 
submittal.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The setpoint changes for reactor trip and High Pressure Injection 
(HPI) actuation will result in a very small (approximately one-percent) 
increase in the probability for reactor trips. Review of industry data 
shows that this increase is not significant. The revised accident 
analysis has determined that transients which reduce Reactor Coolant 
System (RCS) pressure below the new setpoints, warrant the associated 
action. Engineered Safeguards Actuation System (ESAS) and Reactor 
Protection System (RPS) actuations are used to mitigate accidents and 
are not the initiator of analyzed accidents. Therefore, the probability 
of previously evaluated accidents is not affected.
    RPS and ESAS functions are assumed to actuate to mitigate 
transients. The revised setpoints will ensure earlier actuation of the 
RPS and ESAS on a low RCS pressure condition. Raising the ESAS Low RCS 
Pressure Setpoint will ensure earlier automatic HPI actuation for a 
portion of the spectrum of pressure decreasing events. For rapid 
depressurization events, such as main steam line break and large break 
Loss of Coolant Accident (LOCA), this will have little impact. For 
slower events, or those that do not reach the current setpoint during 
the initial subcooled blowdown phase, HPI will be automatically 
initiated substantially earlier in the event. This will increase the 
integrated HPI flow to the RCS during the time the core is likely to be 
uncovered, thereby reducing the consequential PCT. This additional flow 
results in a significant peak clad temperature (PCT) decrease for small 
break LOCA scenarios less than 0.07 square feet. Based on the above, 
the consequences of previously evaluated accidents will not be 
increased.
    The HPI system characteristics will not be affected such that the 
probability of any accident is increased. The system flow restriction 
for protection from low temperature overpressure (LTOP) events will be 
maintained. The HPI system is used for accident mitigation and is not 
the initiator of evaluated accidents other than LTOP. The proposed 
surveillance changes will ensure that all valves throttled in the HPI 
flowpath are verified and secured in the correct position. The throttle 
valves and stop check valves will be positioned to ensure HPI flow is 
within analyzed limits. Therefore, the consequences of accidents that 
rely on HPI flow will not be increased.
    Based on the above evaluation, the probability or consequences of 
evaluated accidents are not significantly increased by these changes.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The change to RPS and ESAS setpoints will not change the functions 
of plant equipment, no new system interactions will be created, and no 
new failure modes will be introduced. The setpoint changes will permit 
earlier actuation for the associated actions. However, no new plant 
conditions will be introduced by the setpoint changes.
    The HPI modifications include the installation of throttle valves 
that will change the flow characteristics of the system. The new 
throttle valves are manual valves that will be secured in position. The 
revised surveillance requirements will ensure these valves are 
positioned such that HPI flow is within analyzed limits. Therefore, no 
conditions are created that could cause a new type of accident.
    Based on the above evaluation, these changes cannot create the 
possibility of an accident of a different type than previously 
evaluated in the [Safety Analysis Report] SAR.
    3. Does not involve a significant reduction in the margin of 
safety.
    The safety function of the affected portions of the RPS and ESAS 
systems is to actuate their respective functions if RCS pressure drops 
below the setpoint. The raised RPS and ESAS setpoints will provide 
earlier actuation for these protective features. These changes will 
increase the margin of safety provided by the associated Technical 
Specifications.
    The safety function of the HPI system is to provide cooling to 
limit fuel peak clad temperature. The revised surveillance requirements 
will ensure valves are positioned such that HPI flow is within analyzed 
limits. Therefore, the margin of safety provided by the HPI 
surveillance requirements is maintained.
    Based on the above evaluation, there is no reduction in the margin 
of safety associated with the equipment and systems affected by this 
change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: December 3, 1998.
    Description of amendment request: The proposed change revises the 
TMI-1 Core Protection Safety Limits and Core Protection Safety Bases, 
as specified in Technical Specification Figures 2.1-1 and 2.1-3, to 
provide more restrictive limits which reflect the decrease in reactor 
coolant system flow resulting from the analysis of increased once-
through steam generator (OTSG) tube plugging limits (total allowable 
number of tubes plugged). The licensee is currently restricted to a 
total of 2,000 tubes plugged in both OTSGs which corresponds to 6.4 
percent of the total number of tubes. The licensee's more restrictive 
Core Protection Safety Limits reflect the reduction in reactor coolant

[[Page 71968]]

flow that would exist if an average of 20 percent of the OTSG tubes 
were plugged.
    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 
of occurrence or the consequences of an accident previously evaluated. 
An increase in the average steam generator tube plugging (SGTP) level 
to 20% results in a small reduction of reactor coolant system (RCS) 
flow rates and primary to secondary heat transfer. These changes result 
in small changes to the primary and secondary side operating 
parameters, and do not result in any additional challenges to plant 
equipment. The proposed Technical Specification Changes resulting from 
the increase in allowable tube plugging limits are more restrictive but 
remain bounded by the existing reactor protection system (RPS) trip 
setpoints. The assessment of the NSSS [nuclear steam supply system] 
primary components, including the reactor pressure vessel, reactor 
core, reactor coolant pump, steam generator, pressurizer, control rod 
drive mechanisms, and RCS piping concluded that the integrity of these 
components will be unaffected by the increase in average SGTP level.
    A re-analysis of the bounding Updated Final Safety Analysis Report 
(UFSAR) Chapter 14 accidents, specifically the startup accident, loss 
of coolant flow, loss of feedwater, and large and small break LOCA 
demonstrated compliance with the acceptance criteria. The RCS pressure 
boundary is not challenged, and the DNBR [departure from nucleate 
boiling ratio] and peak clad temperature values remain within the 
specified limits of the licensing basis. An analysis of the loss of 
electric power accident demonstrated the ability of the plant to 
transition smoothly to natural circulation with an average of 20% SGTP 
or with asymmetric plugging. It was also determined that the current 
mass and energy release data used for the containment integrity and 
equipment qualification remain bounding. Since the design requirements 
and safety limits continue to be met, system functions are not 
adversely impacted, and the integrity of the RCS pressure boundary is 
not challenged, the radiological consequences remain unchanged. 
Therefore, this activity does not involve a significant increase in the 
probability of occurrence or the 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. The proposed Technical 
Specification changes are more restrictive core protection safety 
limits but remain bounded by the existing RPS trip setpoints. This 
proposed change assures safe operation commensurate with the effects of 
steam generator tube plugging. This increase in the average level of 
SGTP to 20% will not introduce any new accident initiator mechanisms. 
No new failure modes or limiting single failures have been identified. 
Since the safety and design requirements continue to be met and the 
integrity of the RCS pressure boundary is not challenged, no new 
accident scenarios have been created. This change does not add any new 
equipment, modify any interfaces with existing equipment, or change the 
equipment function or the method of operating the equipment. Reactor 
core, RCS, and steam generator parameters remain within appropriate 
design limits during normal operation. Therefore, this activity does 
not create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The existing RPS trip setpoints bound the proposed Technical 
Specification changes resulting from 20% SGTP. This change assures safe 
operation commensurate with the effects of steam generator tube 
plugging. The TMI-1 DNB design basis, RCS pressure limits, peak clad 
temperature limits and dose criteria are maintained for all UFSAR 
transients. Therefore, this activity does not reduce the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania (REGIONAL DEPOSITORY), Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2 (NMP2), Oswego County, New York

    Date of amendment request: November 16, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) related to the implementation of 
systems for the detection and suppression of coupled neutronic/thermal-
hydraulic instabilities in the reactor. Average Power Range Monitor 
(APRM) flow control trip reference cards will initiate a reactor scram 
to limit the oscillation magnitude at reactor trip so as to limit the 
associated Critical Power Ratio change and, in conjunction with Minimum 
Critical Power Ratio (MCPR) operating limits, assure compliance with 
the MCPR safety limit. In addition, the changes would increase the APRM 
flow biased neutron flux scram and control rod block settings to allow 
plant operation in the Extended Load Line Limit Analysis region. Thus, 
the proposed changes are in regard to setpoints and calculations for 
fuel cladding integrity and the associated TS Bases. In the Bases for 
TS 2.1.1, the proposed change would reference new equations in TS 
2.1.2a. In TS 2.1.2a, the proposed change would be to the equation for 
determining the flow biased APRM scram and rod block trip setpoints. In 
the Bases for TS 2.1.2a, the proposed change would reflect the new 
setpoints. In the Bases for TS 2.2.2, the proposed change would be to 
the description of the setpoint methodology which is based upon General 
Electric Report NEDC-31336, ``GE Instrumentation Setpoint 
Methodology.'' In Note (m) of TS Table 3.6.2/4.6.2, the proposed change 
would be to the calibration range for the APRM channel setpoint. In the 
Bases for TS 3.6.2/4.6.2, the proposed change would be to the equations 
and methodology for determining APRM scram and rod block setpoints. In 
TS 6.9.1.f, which identifies documents approved by NRC for analytical 
methods used to determine core operating limits, the proposed change 
would add ``NEDO-32465-A, Reactor Stability Detect and Suppress 
Solutions Licensing Basis Methodology for Reload Applications, August 
1996.''
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards

[[Page 71969]]

consideration, which is presented below:
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The APRM neutron monitoring system is not an initiator or a 
precursor to an accident. The neutron monitoring system monitors the 
power level of the reactor core and provides automatic core protection 
signals in the event of a power transient. A Restricted Region will be 
maintained such that the probability of a stability event is not 
increased. Therefore, the proposed TS changes cannot affect the 
probability of a previously evaluated accident.
    The proposed TS changes will revise the APRM flow-biased neutron 
flux scram TS setting to provide automatic protection to assure that 
anticipated coupled neutronic/thermal-hydraulic instabilities will not 
compromise established fuel safety limits. The proposed changes will 
result in a more restrictive APRM flow-biased scram trip setting in the 
low flow regions of the power/flow operating map (i.e., operational 
conditions where reactor instabilities are most probable). In other 
words, the new settings will provide a scram sooner (at a lower power 
level) than the existing settings. The associated control rod block 
setting will also be revised. A margin between the control rod block 
and flux scram has been determined by calculation.
    The proposed changes will also revise the APRM flow-biased neutron 
flux scram and control rod block TS settings to provide an increase 
above the current values in operating conditions not susceptible to 
reactor instabilities. Specifically, the proposed changes will 
implement a 2% increase in the analytical limit of the APRM flow-biased 
flux scram and a 7% increase in the analytical limit of the APRM flow-
biased control rod block. Evaluation demonstrates that these proposed 
analytical limit increases have negligible impact on the transient 
events results for NMP1 [Nine Mile Point Unit 1] as documented in 
Chapter XV of the NMP1 UFSAR, [Updated Final Safety Analysis Report], 
including the limiting transient events which are reanalyzed each 
reload. Of the twenty-five (25) transient events analyzed in Section XV 
of the NMP1 UFSAR, only the Inadvertent Startup of Cold Recirculation 
Loop event and the Recirculation Flow Controller Malfunction--Increase 
Flow event have potentially impacted results. The Chapter XV Control 
Rod Drop Accident as well as the Turbine Trip with No Bypass at Partial 
Power event were also evaluated.
    For the Inadvertent Startup of Cold Recirculation Loop event, the 
proposed 2% increase in the high neutron flux scram would result in an 
increase in the fuel average surface heat flux response. However, there 
is significant margin between the surface heat flux value for this 
event and the current limiting MCPR [Minimum Critical Power Ratio] 
event (the Feedwater Controller Failure Maximum Demand event). As such, 
any small change to the fuel surface heat flux response due to the high 
neutron flux scram analytical limit increase would not result in the 
fuel thermal margin requirements for the Inadvertent Startup of Cold 
Recirculation Loop event to exceed the MCPR limits set by the limiting 
reload analysis event.
    The reactor neutron flux for the Recirculation Flow Controller 
Malfunction--Increase Flow event also showed an increasing trend from 
its initial value. However, the peak response for this parameter (104% 
of rated) is significantly below the high neutron flux scram analytical 
limit. Accordingly, the proposed increase to the high neutron flux 
scram analytical limit does not affect the response to this transient 
event.
    The Control Rod Drop Accident is included in Chapter XV of the NMP1 
UFSAR. As noted in NEDE-24011-P-A, ``GESTAR II: General Electric 
Standard Application for Reactor Fuel,'' the initial power burst from 
this event is terminated by the Doppler reactivity feedback while the 
scram provides the final event termination several seconds later. The 
120% APRM scram limit was conservatively chosen. The time delay 
introduced by the small change in analytical limit will be 
inconsequential due to the extremely rapid power rise for this event 
(i.e., the time of scram for a 120% analytical limit vs. a 122% 
analytical limit is essentially the same).
    The proposed Bases changes to TS 3.6.2/4.6.2 and TS 2.2.2 simply 
provide details of the setpoint methodology currently used as well as 
specific allowable values.
    Therefore, the proposed TS changes to implement a more restrictive 
flow-biased scram setting to protect against reactor instabilities and 
the proposed change to increase the high neutron flux scram and rod 
block analytical limits do not result in a significant increase in the 
consequences of an accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes will revise the APRM flow-biased neutron flux 
scram TS settings to assure anticipated coupled neutronic/thermal-
hydraulic instabilities will not compromise established fuel safety 
limits in the low flow regions of the power/flow operating map as well 
as revise the associated control rod block settings. These changes also 
propose a 2% increase in the analytical limit of the APRM flow-biased 
neutron flux scram and a 7% increase in the analytical limit of the 
APRM flow-biased control rod block. These changes do not introduce any 
new accident precursors and do not involve any alterations to plant 
configurations which could initiate a new or different kind of 
accident. The proposed changes do not affect the intended function of 
the APRM system nor do they affect the operation of the system in a way 
which would create a new or different kind of accident.
    Therefore, the proposed changes will not create the possibility of 
a new or different kind of accident from any previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    More conservative APRM flow-biased neutron flux scram and control 
rod block settings will be implemented in the low flow regions of the 
power/flow operating map. The scram setting change will assure that 
anticipated coupled neutronic/thermal-hydraulic instabilities will not 
compromise established fuel safety limits. The proposed changes will 
also implement a 2% increase in the APRM flow-biased neutron flux scram 
and a 7% increase in the APRM flow-biased control rod block in those 
operating regions not susceptible to reactor instabilities. Evaluation 
demonstrates that these proposed increases have negligible impact on 
the transient events or accident results for NMP1. The impacted 
transient events are either not the limiting MCPR event, the peak 
response to the event is significantly below the high neutron flux 
scram analytical limit or in the case of the Control Rod Drop Accident, 
the time delay introduced by the change will be inconsequential due to 
the extremely rapid power rise. No other events are adversely affected. 
Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 71970]]

review, it appears that the three standards of 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2 (NMP2), Oswego County, New York

    Date of amendment request: November 19, 1998.
    Description of amendment request: The proposed amendment would 
change the surveillance frequencies in Technical Specifications (TSs) 
4.8.4.4a, ``Surveillance Requirements--Reactor Protection System 
Electric Power Monitoring (RPS Logic),'' and 4.8.4.5a, ``Surveillance 
Requirements--Reactor Protection System Electric Power Monitoring 
(Scram Solenoids),'' to require channel functional testing of the RPS 
Motor Generator Set (M/G) and RPS Uninterruptible Power Supplies (UPS) 
Electrical Protection Assemblies (EPAs) at least once every 6 months. 
These TSs currently require that channel functional testing be 
performed each time the plant is in cold shutdown for a period of more 
than 24 hours, unless performed within the previous 6 months.
    Basis for proposed no significant hazards consideration 
determination: During the last refueling outage, the licensee modified 
the Nine Mile Point Unit No. 2 (NMP2) design for the RPS M/G and RPS 
UPS EPAs to provide relay actuated protection systems. The relays of 
the new design may be individually isolated from an essential power 
circuit for testing and may be actuated without tripping the associated 
breaker. The relay actuated system will allow the EPA system monitoring 
an essential power supply to be functionally tested with the plant on-
line. The EPA relay actuation setpoints are not affected by the 
modification or the proposed TS changes. The licensee states that the 
design, installation, and testing of the new units meet the criteria of 
the same standards that were applied to the previous units.
    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 operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes affect surveillance testing frequency only. 
The new relay actuated protection system design functions in the same 
fail safe manner as the old units. Also, the new design in conjunction 
with the testing capability has increased EPA reliability, while 
introducing little risk to testing the EPAs with the plant in 
operation. Therefore, the proposed changes to the NMP2 TS do not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes affect surveillance testing frequency of relay 
actuated protection circuits only. The proposed changes do not 
introduce any new or different accident initiators from any that were 
previously evaluated. EPA relay actuation setpoints are not affected. 
The actual fail safe system conditions required for EPA actuation will 
remain the same. Therefore, the operation of NMP2, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The function of the EPA systems is to isolate the loads from supply 
power. That function was not altered by the proposed change. 
Reliability of the EPA systems is improved. Therefore, the operation of 
NMP2, in accordance with the proposed amendment, will not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station, Unit 1 (NMP1), Oswego County, New York

    Date of amendment request: November 30, 1998.
    Description of amendment request: The proposed amendment would 
correct Technical Specification (TS) 3.1.2, ``Liquid Poison System,'' 
and the associated TS Bases. Specifically, in the Bases for TS 3.1.2, 
the boron-10 concentration of 120 ppm (which is incorrectly calculated 
using atomic percent instead of weight percent) would be changed to 
109.8 ppm. In TS 3.1.2, the minimum volume of the sodium pentaborate 
solution contained in the Liquid Poison System storage tank would be 
increased from 1185 gallons to 1325 gallons.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The Liquid Poison System is designed to provide the capability to 
bring the reactor from a full design rating to a shutdown condition 
assuming none of the control rods can be inserted. The system is 
manually initiated in response to a failure of the Control Rod Drive 
System to shutdown the reactor. The proposed changes revise the 
required liquid poison solution volume and concentration. The proposed 
changes to the Technical Specifications and the Bases require no 
changes to the physical facility which could adversely affect any 
accident precursors. Therefore, the proposed changes cannot 
significantly increase the probability of an accident.
    The proposed changes will assure that the Liquid Poison System 
continues to provide the capability to shutdown the reactor during an 
ATWS [Anticipated Transient Without Scram] event. In addition, the 
system will continue to be capable of bringing the reactor to cold 
shutdown, 3 percent delta k subcritical (0.97 keff), from a 
full design rating of

[[Page 71971]]

1850 megawatts thermal assuming none of the control rods can be 
inserted, and considering the combined effects of coolant voids, 
temperature change, fuel doppler, and xenon and samarium. Therefore, 
the change to the Technical Specifications does not significantly 
increase the consequences of a previously evaluated accident.
    2. The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Injection of the sodium pentaborate solution into the reactor 
vessel has been considered in the plant design. The proposed changes 
revise the required liquid poison solution volume and concentration. 
The proposed changes make no physical modification to the plant which 
could create the possibility of a new or different kind of accident. 
The proposed changes will maintain the capability of the Liquid Poison 
System to shutdown the reactor from its full design rating assuming 
none of the control rods are inserted, and considering the combined 
effects of coolant voids, temperature change, fuel doppler, and xenon 
and samarium. Consequently, these changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed changes revise the required liquid poison solution 
volume and concentration. The proposed changes make no physical 
modification to the plant which could reduce the margin of safety. 
These changes will assure compliance with the requirements of 
10CFR50.62, ``Requirements for Reduction of Risk from Anticipated 
Transients without Scram (ATWS) Events for Light-Water-Cooled Nuclear 
Power Plants.'' In addition, these changes will maintain the capability 
of the Liquid Poison System to bring the reactor from a full design 
rating of 1850 megawatts thermal to greater than 3 percent delta k 
subcritical (0.97 keff) assuming none of the control rods 
can be inserted, and considering the combined effects of coolant voids, 
temperature change, fuel doppler, xenon and samarium.
    The required volume of boron-10 solution in the Liquid Poison 
System storage tank includes an additional 25 percent margin beyond the 
amount needed to shutdown the reactor to allow for any unexpected non-
uniform mixing. Also, the total storage tank volume of sodium 
pentaborate solution incorporates 197 gallons of solution which is 
unavailable for injection into the reactor vessel and a 25 gallon 
margin for conservatism. Additionally, using one 30 gpm Liquid Poison 
System pump, the injection time is greater than 17 minutes thereby 
assuring adequate mixing. The proposed changes to the liquid poison 
concentration and volume ensure the NMP1 [Nine Mile Point Unit 1] 
Liquid Poison System is able to meet its safety function requirements. 
Therefore, this change will not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: December 4, 1998.
    Description of amendment request: The proposed amendment would 
eliminate the need to cycle the plant and its components through a 
shutdown-startup cycle by allowing the next snubber surveillance 
interval to be deferred until the end of refueling outage 6 or 
September 10, 1999, whichever date is earlier.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 10 CFR 
50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this conclusion 
is that the three criteria of 10CFR50.92(c) are not satisfied. The 
proposed revision does not involve [an] SHC because the revision would 
not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    The proposed change is for a one time extension to the surveillance 
interval of snubber inspections required by Technical Specification 
4.7.10.e. The change involves revising the calendar time for snubber 
interval inspections to 36 months to coincide with the time frame of 
the current cycle 6 operation.
    Snubber testing experience at Millstone Unit No. 3 has shown that 
historical failure rates of snubbers are low. During the third 
refueling outage, after an operating cycle of approximately 22 months, 
the functional testing program identified multiple Type A failures 
attributed primarily to original plant construction, and resulted in a 
full inspection of all Type A snubbers. The snubber inspection interval 
was extended to approximately 30 months by a one-time extension to the 
Technical Specifications for the fourth refueling outage and only one 
Type A snubber failure was identified. Subsequent outages with 
operating durations of 18 and 17 months also identified only a single 
Type B failure in each outage. The results of piping stress analysis 
which have been performed to assess the impact of snubbers which have 
failed to meet functional test acceptance criteria have shown that 
neither piping system functionality or structural integrity have ever 
been compromised.
    During the recent cycle 6 operation Millstone 3 has experienced an 
extended midcycle shutdown, where temperature, vibration effects and 
normal wear on snubbers have been minimized as compared to a normal 
operating cycle. The last snubber surveillance interval inspections 
were completed during this midcycle shutdown. Although the calendar 
surveillance interval is impacted by this change the primary conditions 
that present challenges to snubbers have not been prevalent during the 
extended shutdown. Given the low failure rates of snubbers over the 
last 3 surveillance intervals, and the fact the operating time of the 
remainder of cycle 6 will be approximately 1 year, snubber failures are 
expected to be similar to previous intervals.
    Accordingly the possibility of a snubber failure leading to a 
Decrease in Reactor Coolant Inventory or a Decrease in Heat Removal by 
the Secondary System is not increased and there is no affect on the 
probability of previously evaluated accidents.
    This change does not include any physical changes to the plant and 
does not affect acceptance criteria or the

[[Page 71972]]

required actions for functional failures of snubbers. Accordingly there 
is no increase in the consequences of previously evaluated accidents 
resulting in a Decrease in Reactor Coolant Inventory or a Decrease in 
Heat Removal by the Secondary System.
    Thus it is concluded that the proposed revision does not involve a 
significant increase in the probability or consequence of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This proposed revision to the surveillance interval does not change 
the operation of any plant system or component during normal or 
accident conditions. The proposed change extends the surveillance 
interval of snubber inspections required by Technical Specification 
4.7.10.e. The change involves revising the calendar time for snubber 
interval inspections to coincide with the time frame of current cycle 6 
operation. This change does not include any physical changes to the 
plant and does not affect acceptance criteria or the required actions 
for functional failures of snubbers.
    Thus, this proposed revision does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change extends the surveillance interval of snubber 
inspections required by Technical Specification 4.7.10.e. The change 
involves revising the calendar time for snubber interval inspections to 
coincide with the time frame of current cycle 6 operation. This change 
does not include any physical changes to the plant and does not affect 
acceptance criteria or the required actions for functional failures of 
snubbers. The service life of the snubbers or parts as required by 
Technical Specification 4.7.10.i will not be impacted by this change 
since the required replacements have already occurred and no additional 
service life dates will expire prior to September 10, 1999.
    Thus, it is concluded that the proposed revision does not involve a 
significant reduction in a margin of safety.
    In conclusion, based on the information provided, it is determined 
that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, Attn: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Project Director: William M. Dean.

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

    Date of amendment request: November 24, 1998.
    Description of amendment request: The proposed amendment would 
revise the Ginna Station Improved Technical Specifications description 
of the fuel cladding material (TS 4.2.1) and to update the list of 
references provided in Specification 5.6.5 for the Core Operating 
Limits Report.
    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:

Evaluation of Administrative Changes

    The administrative changes [related to the update of references 
provided in Specification 5.6.5 for the Core Operating Limits report] 
do not involve a significant hazards consideration as discussed 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 
revise Administrative Controls Section 5.6.5.b to update the references 
to NRC approved documents which support the analysis for the Heat Flux 
Hot Channel Factor in the Core Operating Limits Report and to provide 
clarification to the currently applicable methodology. It revises the 
Design Features Section 4.2.1 to provide clarification of the types of 
zirconium alloy filler rod material that have received previous NRC 
approval and to clarify that the application shall be NRC approved. 
Section 4.2.1 is revised to clarify that the analyses performed to 
verify compliance with the fuel safety design bases shall be cycle 
specific. 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 
administrative changes do not affect the manner by which the plant is 
operated and no new equipment will be installed. The proposed 
administrative changes will not impose any new or different 
requirements. All original design and performance criteria continue to 
be met, and no new failure modes have been created for any system, 
component, or piece of equipment. Thus, these changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    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 methodology has been shown to meet all applicable design criteria 
and ensure that all pertinent licensing basis acceptance criteria are 
met. As such, no question of safety is involved, and the changes do not 
involve a significant reduction in a margin of safety.

Evaluation of Less Restrictive Changes

    The less restrictive change [related to the fuel cladding material 
(TS 4.2.1)] does not involve a significant hazards consideration as 
discussed below:
    (1) Operation of Ginna Station in accordance with the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The Westinghouse 
14 x 14 VANTAGE + fuel assemblies containing fuel rods fabricated with 
ZIRLO alloy meet the same fuel assembly and fuel rod design bases as 
Westinghouse 14 x 14 OFA [Optimized Fuel Assembly] fuel assemblies in 
the other fuel regions. In addition, the 10 CFR 50.46 criteria will be 
applied to the fuel rods fabricated with ZIRLO alloy. The use of these 
fuel assemblies will not result in a change to the proposed Ginna 
Westinghouse 14 x 14 OFA reload design and safety analysis limits. The 
ZIRLO alloy is similar in chemical composition and has similar physical 
and mechanical properties as that of Zircaloy-4. Thus the cladding 
integrity is maintained and the structural integrity of the fuel

[[Page 71973]]

assembly is not affected. The ZIRLO clad fuel rods improve corrosion 
resistance and dimensional stability. The use of ZIRLO does not impact 
the radiological consequences of accidents previously evaluated in the 
Safety Analysis. The RCS [reactor coolant system] isotopic inventory is 
negligibly impacted; therefore, changes in postulated releases from the 
RCS or the secondary systems are negligible. Assumptions of fuel 
melting in the radiological analyses are not based on the type of fuel 
cladding. For those accidents where fuel melting is postulated to occur 
(control rod ejection, locked [seized] RCP rotor), the amount of fuel 
undergoing melting and clad damage using ZIRLO clad is bounded by the 
current values used in the Safety Analysis. Therefore, the probability 
or consequences of an accident previously evaluated is not 
significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The Westinghouse 
14 x 14 VANTAGE + fuel assemblies containing fuel rods fabricated with 
ZIRLO alloy will satisfy the same design bases as that used for 
Westinghouse 14 x 14 OFA fuel assemblies in the other fuel regions. 
Since the original design criteria is being met, the fuel rods 
fabricated with ZIRLO alloy will not be an initiator for any new 
accident. All design and performance criteria will continue to be met 
and no new single failure mechanisms have been created. In addition, 
the use of these fuel assemblies does not involve any alterations to 
plant equipment or procedures which would introduce any new or unique 
operational modes or accident precursors. Therefore, the possibility 
for a new or different kind of accident from any accident previously 
evaluated is not created.
    (3) Operation of Ginna Station in accordance with the proposed 
change does not involve a significant reduction in a margin of safety. 
The Westinghouse 14 x 14 VANTAGE + fuel assemblies containing fuel rods 
fabricated with ZIRLO alloy do not change the proposed Ginna 
Westinghouse 14 x 14 OFA reload design and safety analysis limits. The 
use of these fuel assemblies containing fuel rods fabricated with ZIRLO 
alloy will take into consideration the normal core operating conditions 
allowed in the Technical Specifications. For each cycle reload core, 
these fuel assemblies will be specifically evaluated using approved 
reload design methods and approved fuel rod design models and methods 
as specified in Technical Specifications. This will include 
consideration of the core physics analysis peaking factors and core 
average linear heat rate effects. In addition, the 10 CFR 50.46 
criteria will be applied each cycle to the fuel rods fabricated with 
ZIRLO alloy. Analyses or evaluations will be performed each cycle to 
confirm that 10 CFR 50.46 will be met. Therefore, the margin of safety 
as defined in the Bases to the Ginna Technical Specifications is not 
significantly reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: 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 Project Director: S. Singh Bajwa.

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 23, 1998.
    Description of amendment requests: The proposed change would revise 
the Technical Specifications (TS) to (1) reinstate the log power 
reactor trip at or above 4E-5% RATED THERMAL POWER (RTP); (2) reinstate 
reactor trips for Reactor Coolant Flow--Low (RCS flow), the Local Power 
Density--High (LPD), and the Departure from Nucleate Boiling Ratio--Low 
(DNBR); (3) remove the word ``automatically'' from notes (a) and (d) of 
Table 3.3.1-1 to clarify that the manual enable of the trip is 
permissible; and, (4) clarify that the setpoints on Table 3.3.1-1 are 
set relative to logarithmic power, not thermal power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change to TS 3.3.1 does not adversely impact 
structure, system, or component design or operation in a manner which 
would result in a change in the frequency of occurrence of accident 
initiation. SCE has re-analyzed the relevant accidents and established 
that accident consequences are not significantly increased by the 
proposed changes to the bypass-permissive and enable setpoints. The 
reactor trip bypass and automatic enable functions are not accident 
initiators. Consequently, the proposed TS change will not significantly 
increase the probability of accidents previously evaluated. Therefore, 
this amendment request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    No new or different accidents result from changing the reactor trip 
bypass-permissive and automatic enable setpoints. Introducing an 
uncertainty band for the enable setpoints delays the mitigation action 
of the reactor trip for the design basis analysis for the events that 
credit this trip. The enable setpoint itself does not cause any 
accident. Therefore, the amendment request does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    SCE [Southern California Edison Company] has re-analyzed the 
accidents and determined that the consequences of the accidents are 
within their acceptance criteria under the proposed amendment so that 
the margin of safety that bounds the setpoint in both directions 
remains intact. The analyses are relatively insensitive to the reactor 
trip automatic enable setpoints, and no significant reduction in the 
margins of safety ensues from the relatively minor proposed changes to 
the bypass-permissive and enable setpoints, nor from establishing 
allowable values for these points.
    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, Irvine, California 92713.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, P.O. Box 800, Rosemead, California 91770.

[[Page 71974]]

    NRC Project Director: William H. Bateman.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: November 23, 1998.
    Description of amendment request: The proposed amendment relocates 
descriptive design information from Technical Specification 3/4.7.1.1 
(Table 3.7-2), regarding orifice sizes for main steam line Code safety 
valves, to the 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:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change relocates the orifice size design information 
for the main steam line Code safety valves, found in Table 3.7-2, that 
does not meet the criteria for inclusion in Technical Specifications as 
identified in 10 CFR 50.36(c)(2)(ii). The affected descriptive design 
information is not related to any assumed initiators of analyzed events 
and is not assumed to mitigate accident or transient events. The 
limiting condition for operation for the main steam line Code safety 
valves is not altered by the proposed change. The orifice size design 
information will be relocated from Table 3.7-2 of Specification 3/
4.7.1.1 to the Bases section for that same Technical Specification and 
will be maintained pursuant to 10 CFR 50.59. In addition, surveillance 
testing details for this Technical Specification are addressed in 
existing surveillance procedures, which are also controlled by 10 CFR 
50.59, and subject to the change control provisions imposed by plant 
administrative procedures, which endorse applicable regulations and 
standards. Therefore, the change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change relocates the orifice size design information 
for the main steam line Code safety valves, found in Table 3.7-2, that 
does not meet the criteria for inclusion in Technical Specifications as 
identified in 10 CFR 50.36(c)(2)(ii). The change does not involve a 
physical alteration of the plant (no new or different type of equipment 
will be installed) or make changes in the methods governing normal 
plant operation. The change will not impose different requirements, and 
adequate control of information will be maintained. This change will 
not alter assumptions made in the safety analysis and licensing basis. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change relocates the orifice size design information 
for the main steam line Code safety valves, found in Table 3.7-2, that 
does not meet the criteria for inclusion in Technical Specifications as 
identified in 10 CFR 50.36(c)(2)(ii). The change will not reduce a 
margin of safety since it has no impact on any safety analysis 
assumptions. In addition, the relocated orifice size design information 
remains the same as the existing Technical Specifications. Since any 
future changes to this orifice size information (that will be located 
in the Bases section) will be evaluated per the requirements of 10 CFR 
50.59, there is no reduction in a margin of safety.
    The proposed change is also consistent with the Westinghouse Plants 
(Improved) Standard Technical Specification, NUREG-1431, approved by 
the NRC Staff. Revising the Technical Specification to reflect the 
approved content of NUREG-1431 ensures no significant reduction in the 
margin of safety. Therefore, the 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

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: November 11, 1998.
    Brief description of amendments: The proposed amendments revise 
core safety limit curves and Overtemperature N-16 reactor trip 
setpoints based on analyses of the core configuration and expected 
operation for Comanche Peak Steam Electric Station (CPSES) Unit 2, 
Cycle 5. The changes apply equally to CPSES Units 1 and 2 licenses 
since the Technical Specifications are combined.
    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?
    A. Revision to the Unit 2 Core Safety Limits
    Analyses of reactor core safety limits are required as part of 
reload calculations for each cycle. TU Electric has performed the 
analyses of the Unit 2, Cycle 5 core configuration to determine the 
reactor core safety limits. The methodologies and safety analysis 
values result in new operating curves which, in general, permit plant 
operation over a similar range of acceptable conditions. This change 
means that if a transient were to occur with the plant operating at the 
limits of the new curve, a different temperature and power level might 
be attained than if the plant were operating within the bounds of the 
old curves. However, since the new curves were developed using NRC 
approved methodologies which are wholly consistent with and do not 
represent a change in the Technical Specification BASES for safety 
limits, all applicable postulated transients will continue to be 
properly mitigated. As a result, there will be no significant increase 
in the consequences, as determined by accident analyses, of any 
accident previously evaluated.
    B. Revision to Unit 2 Overtemperature N-16 Reactor Trip Setpoints
    As a result of changes discussed, the Overtemperature reactor trip 
setpoint has been recalculated. These trip setpoints help ensure that 
the core safety limits are protected and that all applicable limits of 
the safety analysis are met.
    Based on the calculations performed, no significant changes to the 
safety

[[Page 71975]]

analysis values for Overtemperature reactor trip setpoint were 
required. The f(delta I) trip reset function was revised due to less 
top-skewed axial power distributions predicted for this cycle. The 
analyses performed show that, using the TU Electric methodologies, all 
applicable limits of the safety analysis are met. This setpoint 
provides a trip function which allows the mitigation of postulated 
accidents and has no impact on accident initiation. Therefore, the 
changes in safety analysis values do not involve an increase in the 
probability of an accident and, based on satisfying all applicable 
safety analysis limits, there is no significant increase in the 
consequences of any accident previously evaluated.
    In addition, sufficient operating margin has been maintained in the 
overtemperature setpoint such that the risk of turbine runbacks or 
unnecessary reactor trips due to upper plenum flow anomalies or other 
operational transients will be minimized, thereby, reducing potential 
challenges to the plant safety systems.
    C. Administrative changes to reflect plant nomenclature
    Changes to the N-16 trip setpoint equation are for clarification 
only to more accurately reflect CPSES plant nomenclature. This change 
is administrative in nature and does not increase in the probability or 
consequences of an accident previously evaluated.

Summary

    The changes in the amendment request apply NRC approved 
methodologies to changes in safety analysis values, new core safety 
limits and new N-16 setpoint and parameter values to assure that all 
applicable safety analysis limits have been met. The potential for an 
operational transient to occur has not been affected and there has been 
no significant impact on the consequences of any accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve the calculation of new reactor core 
safety limits and overtemperature reactor trip setpoint resets. As 
such, the changes play an important role in the analysis of postulated 
accidents but none of the changes effect plant hardware or the 
operation of plant systems in a way that could initiate an accident. 
Changes to the N-16 trip setpoint equation are for clarification only 
to more accurately reflect CPSES plant nomenclature. 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?
    In reviewing and approving the methods used for safety analyses and 
calculations, the NRC has approved the safety analysis limits which 
establish the margin of safety to be maintained. While the actual 
impact on safety is discussed in response to question 1, the impact on 
margin of safety is discussed below:
    A. Revision to the Unit 2 Reactor Core Safety Limits
    The NRC-approved TU Electric reload analysis methods have been used 
to determine new reactor core safety limits. All applicable safety 
analysis limits have been met. The methods used are wholly consistent 
with Technical Specification BASES 2.1 which is the bases for the 
safety limits. In particular, the curves assure that for Unit 2, Cycle 
5, the calculated DNBR is no less than the safety analysis limit and 
the average enthalpy at the vessel exit is less than the enthalpy of 
saturated liquid. The acceptance criteria remains valid and continues 
to be satisfied; therefore, no change in a margin of safety occurs.
    B. Revision to Unit 2 Overtemperature N-16 Reactor Trip Setpoints
    Because the reactor core safety limits for CPSES Unit 2, Cycle 5 
are recalculated, the Reactor Trip System instrumentation setpoint 
values for the Overtemperature N-16 reactor trip setpoint which protect 
the reactor core safety limits must also be recalculated. The 
Overtemperature N-16 reactor trip setpoint helps prevent the core and 
Reactor Coolant System from exceeding their safety limits during normal 
operation and design basis anticipated operational occurrences. The 
most relevant design basis analysis in Chapter 15 of the CPSES Final 
Safety Analysis Report (FSAR) which is affected by the Overtemperature 
reactor trip setpoint is the Uncontrolled Rod Cluster Control Assembly 
Bank Withdrawal at Power (FSAR Section 15.4.2). This event has been 
analyzed with the new safety analysis value for the Overtemperature 
reactor trip setpoint to demonstrate compliance with event specific 
acceptance criteria. Because all event acceptance criteria are 
satisfied, there is no degradation in a margin of safety.
    The nominal Reactor Trip System instrumentation setpoint values for 
the Overtemperature N-16 reactor trip setpoint (Technical Specification 
Table 2.2-1) are determined based on a statistical combination of all 
of the uncertainties in the channels to arrive at a total uncertainty. 
The total uncertainty plus additional margin is applied in a 
conservative direction to the safety analysis trip setpoint value to 
arrive at the nominal and allowable values presented in Technical 
Specification Table 2.2-1. Meeting the requirements of Technical 
Specification Table 2.2-1 assures that the Overtemperature reactor trip 
setpoint assumed in the safety analyses remains valid. The CPSES Unit 
2, Cycle 5 Overtemperature reactor trip setpoint is not significantly 
different from the previous cycle, and thus provides operational 
flexibility to withstand mild transients without initiating automatic 
protective actions. Although the value of the f(delta I) trip reset 
function setpoint is different, the Reactor Trip System instrumentation 
setpoint values for the Overtemperature N-16 reactor trip setpoint are 
consistent with the safety analysis assumptions which have been 
analytically demonstrated to be adequate to meet the applicable event 
acceptance criteria. Thus, there is no reduction in a margin of safety.
    Using the NRC approved TU Electric methods, the reactor core safety 
limits are determined such that all applicable limits of the safety 
analyses are met. Because the applicable event acceptance criteria 
continue to be met, there is no significant reduction in the margin of 
safety.
    C. Administrative changes to reflect plant nomenclature
    Changes to the N-16 trip setpoint equation are for clarification 
only to more accurately reflect CPSES plant nomenclature. This change 
is administrative in nature and has no impact on the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: John N. Hannon.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: December 10, 1998.

[[Page 71976]]

    Description of amendment request: The licensee proposed to correct 
an error in the technical specifications by changing to the use of 
``hydrogen, balance air'' rather than the incorrect ``hydrogen balance 
nitrogen'' for calibration of the Augmented Offgass System hydrogen 
monitors.
    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:
    Based on the criteria for defining a significant hazards 
consideration in 10CFR50.92, operation of VYNPS in accordance with this 
change would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated, because:
    The proposed change is purely administrative in nature--correcting 
instrument calibration requirements to conform the Technical 
Specification with the instrument manufacturer's recommendations. The 
change has no effect on plant hardware, plant design, safety limit 
setting, or plant system operation and therefore does not modify or add 
any initiating parameters that would significantly increase the 
probability or consequences of an accident previously evaluated. This 
change to the Technical Specifications is a correction of an error 
which occurred when the particular Technical Specification was issued. 
The function of this surveillance requirement remains unchanged.
    No new modes of operation are introduced by the proposed change 
such that adverse consequences would result. Accordingly, the 
consequences of previously analyzed accidents are not affected by this 
proposed change.
    The Augmented Off-Gas (AOG) System hydrogen monitors do not serve a 
reactor safety function. In this context, the determination of no 
significant hazards consideration defined in 10CFR50.92 is made based 
on the ``accident previously evaluated'' being a postulated hydrogen 
detonation within the off-gas system downstream of the hydrogen 
recombiners. The hydrogen monitors do not mitigate the consequences of 
an accident, but rather function to preclude a hydrogen explosion 
within the off-gas system. The function of the Augmented Off-Gas System 
hydrogen monitors to prevent a hydrogen detonation is not affected by 
this change.
    (2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated, because:
    Since this change merely corrects Technical Specification wording 
to reflect the actual manufacturer's recommended gas mixture to be used 
for calibrating these instruments, no new or different types of 
accidents are created. Since the calibration gas mixture has a very low 
(approximately 2%) hydrogen concentration, its use does not introduce 
the possibility of fires, explosions, or other hazards which might 
adversely affect safety-related equipment. Therefore, use of the proper 
calibration gas does not create the possibility of a new or different 
kind of accident.
    This change does not affect the operation of any systems or 
components, nor does it involve any potential initiating events that 
would create any new or different kind of accident. Therefore, the 
proposed change does not create the possibility of a new or different 
kind of accident from any previously evaluated for the Vermont Yankee 
Nuclear Power Station.
    (3) Involve a significant reduction in a margin of safety, because:
    This proposed change involving the specification of the correct 
calibration gas mixture ensures that the off-gas system hydrogen 
monitors are properly calibrated and therefore preserve the margin of 
safety in precluding a hydrogen explosion in the off-gas system. 
Administratively changing this specification only establishes the 
appropriate calibration gas for the actual, installed hydrogen 
monitors. Changing the specification to reflect correct practice will 
not reduce the margin of safety.
    The proposed change does not affect any equipment involved in 
potential initiating events or safety limits. Therefore, it is 
concluded that the proposed 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Project Director: Cecil O. Thomas.

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

    Date of amendment request: July 30, 1998 (TSCR 206).
    Description of amendment request: The purpose of the proposed 
amendments is to incorporate changes to the Technical Specifications to 
more clearly define the requirements for Service Water (SW) System 
operability.
    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 Point Beach Nuclear Plant in accordance with 
the proposed amendment[s] does not result in a significant increase in 
the probability or consequences of any accident previously evaluated.
    The Service Water System is primarily a support system for systems 
required to be operable for accident mitigation. Portions of the SW 
system supplying the containment fan coolers also function as part of 
the containment pressure boundary under post accident conditions. 
Failures within the SW system are not an initiating condition for any 
analyzed accident.
    Analyses performed demonstrate that under the Technical 
Specifications allowable configurations, the SW system will continue to 
perform all required functions. The SW system is capable of supplying 
the required cooling water flow to systems required for accident 
mitigation. That is, the SW system removes the required heat from the 
containment fan coolers and residual heat removal heat exchangers 
ensuring containment pressure and temperature profiles following an 
accident are as evaluated in the FSAR [final safety analysis report]. 
This in turn ensures that environmental qualification of equipment 
inside containment is maintained and thus function as required post-
accident.
    SW system response post accident is within all design limits for 
the system. Transient and steady state forces within the system remain 
within all design and operability limits thereby maintaining the 
integrity of the system inside containment and the integrity of the 
containment pressure boundary. Assumptions dependent on containment 
pressure profile for containment leakage assumed in the radiological 
consequence analyses remain valid.
    In addition, removing required heat from containment ensures that 
cooling

[[Page 71977]]

of the reactor core is accomplished for long-term accident mitigation.
    Therefore, operation of the SW system as proposed will not result 
in a significant increase in the probability or consequences of any 
accident previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes do not alter the way in which the SW system 
performs its design functions nor the design limits of the system. The 
proposed changes do not introduce any new or different normal operation 
or accident mitigation functions for the system. Therefore, no new 
accident initiators are introduced by the proposed changes. Operation 
of SW system as proposed cannot result in a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction in a 
margin of safety.
    Analyses performed in support of the proposed amendments 
demonstrate that the SW system continues to perform its function as 
assumed and credited in the accident analyses and radiological 
consequence analyses performed for the Point Beach Nuclear Plant. 
Therefore, the analyses and results are not changed. All analysis 
limits remain met. The SW system continues to be operated and responds 
within all design limits for the system. Therefore, operation of the 
Point Beach Nuclear Plant in accordance with the proposed amendments 
cannot result in a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: September 23, 1998 (TSCR 209).
    Description of amendment request: The purpose of the proposed 
amendments is to remove the test requirements for snubbers from the 
Technical Specifications (TS). These requirements are already included 
in the Point Beach Nuclear Plant In-Service Inspection Program.
    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 Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of an accident previously evaluated.
    These changes do not involve a significant increase in the 
probability of an accident previously evaluated because no such 
accidents are affected by the proposed revisions to delete TS 15.4.3. 
The proposed TS change does not introduce any new accident initiators.
    Initiating conditions and assumptions are unchanged and remain as 
previously analyzed for accidents in the PBNP Final Safety Analysis 
Report. The proposed TS change does not involve any physical changes to 
systems or components, nor does it alter the typical manner in which 
the systems or components are operated. Therefore, these changes do not 
increase the probability of previously evaluated accidents.
    As noted above, the snubber testing requirements included in the 
ASME/ANSI OM-4 Code are more comprehensive and in general more 
conservative than the snubber testing requirements currently contained 
in TS 15.4.13.
    These changes do not involve a significant increase in the 
consequences of an accident or event previously evaluated because the 
source term, containment isolation or radiological releases are not 
being changed by these proposed revisions. The snubber program ensures 
that snubbers function as required, therefore related systems continue 
to function as designed and analyzed. Existing system and component 
redundancy and operation is not being changed by these proposed 
changes. The assumptions used in evaluating the radiological 
consequences in the PBNP Final Safety Analysis Report are not 
invalidated. Therefore, these changes do not affect the consequences of 
previously evaluated accidents.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    These changes do not introduce nor increase the number of failure 
mechanisms of a new or different type than those previously evaluated 
since there are no physical changes being made to the facility. As 
noted above, the snubber testing requirements included in the ASME code 
in general are more comprehensive than the snubber testing requirements 
currently contained in TS 15.4.13 and provide the requisite level of 
assurance of snubber operability. The design and design basis of the 
facility remain unchanged. The plant safety analyses remain unchanged. 
Therefore, the possibility of a new or different kind of accident from 
any accident previously evaluated is not introduced.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not involve a significant reduction in a 
margin of safety.
    The proposed changes do not involve a significant reduction in the 
margin of safety because existing component redundancy is not being 
changed by these proposed changes. There are no changes to the initial 
conditions contributing to accident severity or consequences, and 
safety margins established through the design and facility license 
including the Technical Specifications remain unchanged. Therefore, 
there are no significant reductions in a margin of safety introduced by 
[these] proposed amendment[s].
    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: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

[[Page 71978]]

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

    Date of amendment request: October 7, 1998 (TSCR 207).
    Description of amendment request: The purpose of the proposed 
amendments is to incorporate changes to the Technical Specifications 
(TS) to ensure the 4 kV bus undervoltage input to reactor trip is 
controlled in accordance with the design and licensing basis for the 
facility. One additional administrative change is requested which 
removes the footnote related to the definition of Rated Power in TS 
15.1.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. Operation of the Point Beach Nuclear Plant [PBNP] in accordance 
with the proposed amendments will not create a significant increase in 
the probability or consequences of an accident previously evaluated.
    The changes proposed ensure the Point Beach Nuclear Plant continues 
to be operated in accordance with the design and licensing basis for 
the facility.
    The first change removes a footnote qualifying the definition of 
Rated Power as applied to PBNP Unit 2. This restriction was eliminated 
with the replacement of Unit 2 steam generators as approved by 
Amendments 173 and 177, dated July 1, 1997. The analyses for those 
amendments were performed based on the minimum flow requirements 
specified in Technical Specification 15.3.1.G.3. The note should have 
been deleted from the Technical Specifications at that time. 
Elimination of this note does not result in a change in the operation 
of PBNP from that analyzed and approved in Amendments 173 and 177. 
Therefore, this change is administrative and cannot result in an 
increase in probability or consequences of an accident previously 
evaluated.
    The second change modifies the Limiting Condition For Operation 
[LCO] for the undervoltage reactor trip protection function. This trip 
function is the primary protective function credited in the complete 
loss of flow event analysis in the Final Safety Analysis Report (FSAR) 
Section 14.1.8. As a primary protective function, this trip is required 
to be single failure proof as stipulated in proposed IEEE 279-1968 
documented in FSAR Section 7.2. This change ensures that this 
protective feature is maintained in a condition where single failure 
considerations are satisfied. When single failure criteria cannot be 
met, appropriate action is stipulated to shutdown the unit placing it 
in a condition where the protective function is no longer required. 
Therefore, this change ensures PBNP is operated in accordance with its 
design and licensing basis and cannot result in an increase in the 
probability or consequences of an accident previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The changes proposed by this request remove a footnote qualifying 
the definition of rated power as it applies to PBNP Unit 2 operation, 
and modify the LCO related to the undervoltage reactor trip protective 
function to ensure this function is maintained as required by the PBNP 
design and licensing basis. These changes are in agreement with 
approved analyses. These changes do not introduce any new accident 
initiators or alter the response of the PBNP Units to previously 
analyzed accidents. Therefore, operation of PBNP in accordance with the 
proposed changes cannot result in a new or different kind of accident 
from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not create a significant reduction in a 
margin of safety.
    Operation of the PBNP in accordance with the proposed amendments is 
within the bounds of approved design and licensing basis of the 
facility. The design and licensing basis establish appropriate margins 
of safety. Since operation of the PBNP remains within the approved 
design and licensing basis of the facility, a reduction in a margin of 
safety cannot result.
    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: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: November 18, 1998
    Description of amendment request: The proposed amendment would 
revise the pressure/temperature (P/T) limits and the low-temperature 
overpressure protection (LTOP) requirements in the facility technical 
specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    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 
consequence of an accident previously evaluated.
    Failure of a reactor vessel is not an accident that has been 
previously evaluated; design provisions ensure that this is not a 
credible event. Since the potential consequences of a reactor vessel 
failure are so severe, industry and governmental agencies have worked 
together to ensure that failure will not occur. Compliance with 10 CFR 
50 Appendix G and H ensures that failure of a reactor vessel will not 
occur. The proposed changes do not impact the capability of the reactor 
coolant pressure boundary piping (i.e., no change in operating 
pressure, materials, seismic loading, etc.) and therefore do not 
increase the potential for the occurrence of a LOCA [loss-off-coolant 
accident].
    The LTOP setpoint, revised enabling temperature, and revised P/T 
limits reflected in proposed Figures TS 3.1-1 and TS 3.1-2 ensure that 
the Appendix G pressure/temperature limits are not exceeded, and 
therefore, ensure that RCS integrity is maintained. The changes do not 
modify the reactor coolant system pressure boundary, nor make any 
physical changes to the facility design, material, construction 
standards, or setpoints. The reactor coolant system full power 
operating pressure (2235 psig) is not being changed by this proposed 
amendment. The LTOP valve setpoint remains at less than or equal to 500 
psig. The LTOP enabling temperature based on Figure

[[Page 71979]]

TS 3.1-2 is 200 deg.F and is consistent with ASME Code Case N-514 
guidance of RTNDT + 50 deg.F. The revised enabling 
temperature is lower than the 355 deg.F value in the current TS. 
However, the allowable combination of Appendix G pressures and 
temperatures (refer to the 0 deg.F isothermal cooldown limit) is 
greater for the revised limit curves. The combination of greater 
allowable Appendix G pressure and temperature limits and lower enabling 
temperature produces a larger operating window. A larger operating 
window reduces the likelihood of inadvertently lifting the LTOP relief 
valve while maneuvering the plant through the knee of the P-T curve 
during startup and shutdown. The probability of an LTOP event occurring 
is independent of the pressure-temperature limits for the RCS [reactor 
coolant system] pressure boundary and enabling temperature. Therefore, 
the probability of a[n] LTOP event is not increased.
    The revised heatup and cooldown limit curves and LTOP enabling 
temperature were developed using test results from unirradiated and/or 
irradiated specimens that represent the KNPP [Kewaunee Nuclear Power 
Plant] reactor vessel beltline circumferential weld, closure head 
flange, and intermediate forging. The circumferential beltline weld and 
intermediate forging are the most limiting materials in the reactor 
coolant pressure boundary due to the effects of neutron irradiation 
which cause the flow properties to increase and the toughness to 
decrease. 10 CFR 50, Appendix G states that the metal temperature of 
the closure flange regions must exceed the material unirradiated 
RTNDT by at least 120 deg.F for normal operation and 
90 deg.F for hydrostatic pressure tests and leak tests when the 
pressure exceeds 20 percent of the preservice hydrostatic test 
pressure. Drop weight and Charpy V-notch testing of IP3571 weld metal 
and the intermediate forging material has been performed and used for 
derivation of the revised PTS [pressurized thermal shock] assessment, 
the proposed Appendix G heatup and cooldown limit curves, and the 
corresponding LTOP system enabling temperature. The revised limit 
curves and corresponding LTOP enabling temperature have been developed 
using accepted engineering practices, methods derived from the ASME 
Boiler and Pressure Vessel Code, criteria set forth in NRC Regulatory 
Standard Review Plan 5.3.2, and 10 CFR 50.61. Utilization of the 
revised heatup and cooldown limit curves and corresponding LTOP 
enabling temperature ensures adequate fracture toughness for ferritic 
materials of the pressure-retaining components of the reactor coolant 
pressure boundary. These limit curves provide adequate margins of 
safety during any condition of normal operation, including anticipated 
operational occurrences and system hydrostatic tests, and low 
temperature overpressure protection (corresponding to isothermal events 
during low temperature operations (i.e., less than or equal to 
200 deg.F)) thus ensuring the integrity of the reactor coolant pressure 
boundary.
    The changes do not adversely affect the integrity of the RCS such 
that its function in the control of radiological consequences is 
affected. Radiological off-site exposures from normal operation and 
operational transients, and faults of moderate frequency do not exceed 
the guidelines of 10 CFR 100. In addition, the changes do not affect 
any fission product barrier. The changes do not degrade or prevent the 
response of the LTOP relief valve or other safety-related systems to 
previously evaluated accidents. In addition, the changes do not alter 
any assumption previously made in the radiological consequence 
evaluations nor affect the mitigation of the radiological consequences 
of an accident previously evaluated. Therefore, the consequences of an 
accident previously evaluated will not be increased.
    Thus, operation of KNPP in accordance with the PA does not involve 
a significant increase in the probability or consequences of any 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of accident 
from any previously evaluated.
    Since the potential consequences of a reactor vessel failure are so 
severe, industry and governmental agencies have worked together to 
ensure that failure will not occur. Compliance with 10 CFR 50 Appendix 
G and H ensures that failure of a reactor vessel will not occur. The 
proposed heatup and cooldown limit curves have been constructed by 
combining the most conservative pressure-temperature limits derived by 
using material properties of the intermediate forging, closure head 
flange, and beltline circumferential weld to form a single set of 
composite curves. With NRC approval to use Code Case N-588, the 
intermediate forging and closure head flange become the controlling 
materials for development of the heatup limit curve and the cooldown 
limit curves at low temperatures. At high temperatures, the 
circumferential weld continues to be limiting for development of the 
cooldown limit curves. Use of conservative pressure-temperature limits 
derived by using material properties of the intermediate forging, 
closure head flange, and beltline circumferential weld to form a single 
set of composite curves, does not modify the reactor coolant system 
pressure boundary, nor make any physical changes to the LTOP setpoint 
or design. Proposed Figures TS 3.1-1 and TS 3.1-2 were prepared in 
accordance with regulatory and code requirements and were derived using 
more conservative material property basis and more limiting 
requirements of neutron exposure projections thru 33 EFPY [effective 
full-power years] instead of 20 EFPY.
    The revised LTOP system enabling temperature and the proposed 
Appendix G pressure temperature limitations were prepared using methods 
derived from the ASME Boiler and Pressure Vessel Code and the criteria 
set forth in NRC Regulatory Standard Review Plan 5.3.2. The changes do 
not cause the initiation of any accident nor create any new credible 
limiting failure for safety-related systems and components. The changes 
do not result in any event previously deemed incredible being made 
credible. As such, it does not create the possibility of an accident 
different than previously evaluated.
    The changes do not have any adverse effect on the ability of the 
safety-related systems to perform their intended safety functions. The 
combination of higher allowable Appendix G pressure and temperature 
limits and lower enabling temperature produces a larger operating 
window. The ASME Section XI, Working Group on Operating Plant Criteria 
(WGOPC) has prepared a technical bases document for Code Case N-514. 
The technical bases document is contained in Attachment 3 of Reference 
1. This technical bases document provides justification for enabling 
the LTOP system at temperatures less than 200 deg.F or at coolant 
temperatures corresponding to a reactor vessel metal temperature less 
than RTNDT + 50 deg.F, whichever is greater.
    WGOPC, which has responsibility for Appendix G of Section XI, has 
considered the burden and safety impact imposed by the LTOP criteria, 
and has developed Code guidelines for determining the LTOP set-point 
pressure and the required enabling temperature. These guidelines will 
relieve some operational restrictions, yet provide adequate margins 
against failure for the reactor vessel. Further, by relieving the 
operational restrictions, these guidelines result in a reduced

[[Page 71980]]

potential for activation of pressure relieving devices, thereby 
improving plant safety. Thus, a slightly larger operating window at 
KNPP is viewed to reduce the likelihood of inadvertently lifting the 
LTOP relief valve while maneuvering the plant through the knee of the 
P-T curve during startup and shutdown. The new LTOP operating window 
(i.e., less than or equal to 200 deg.F) is within the existing 
operating band for the residual heat removal system; operating 
procedures allow the LTOP system to be placed into service at 
<400 deg.F. At KNPP, as long as the LTOP relief valve is operable, the 
LTOP system is enabled anytime the residual heat removal system is in 
communication with the reactor coolant system.
    The proposed changes do not make physical changes to the plant or 
create new failure modes. Thus, the PA does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) Involve a significant reduction in the margin of safety.
    The proposed Appendix G pressure temperature limitations and LTOP 
enabling temperature were prepared using methods derived from the ASME 
Boiler and Pressure Vessel Code, including Code Cases N-514 and N-588, 
and the criteria set forth in NRC Regulatory Standard Review Plan 
5.3.2. Reference 1 to this letter provides information to support NRC 
approval to use Code Case N-514 and Code Case N-588 for the KNPP PTS 
evaluation, development of the heatup and cooldown limit curves, and 
establishment of the LTOP system enabling temperature. These documents 
and practices along with the calculational limitations specified in 10 
CFR 50.61 are an acceptable method for implementing the requirements of 
10 CFR 50 Appendices G and H.
    Use of the methodology set forth in the ASME Boiler and Pressure 
Vessel Code, NRC Regulatory Standard Review Plan 5.3.2., 10 CFR 50.61, 
and 10 CFR 50 Appendices G and H ensures that proper limits and safety 
factors are maintained. Thus, the PA does not involve a significant 
reduction in the margin of safety.
    The revised heatup and cooldown limit curves and LTOP system 
enabling temperature were prepared using drop weight and Charpy V-notch 
data for the beltline weld, closure head flange, and intermediated 
forging material along with practices described herein and methods 
derived from the ASME Boiler and Pressure Vessel Code and 10 CFR 50.61. 
The safety factors and margins used in the development of the limit 
curves and LTOP system enabling temperature meet the criteria set forth 
by these documents. Application of low leakage core designs decreases 
the rate of shift in transition temperature from ductile to nonductile 
behavior. The revised limit curves and LTOP enabling temperature 
provide adequate margins of safety during any condition of normal 
operation, including anticipated operational occurrences and system 
hydrostatic tests, and low temperature overpressure protection 
(corresponding to isothermal events during low temperature operations 
(i.e., less than or equal to 200 deg.F)). With the preparation of the 
revised limit curves in accordance with the latest criteria and 
guidance, this PA ensures that proper limits and safety factors are 
maintained.
    Thus, the PA does not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Project Director: Cynthia A. Carpenter.

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

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

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of application for amendment: May 15, 1998, as supplemented 
September 25 and October 13, 1998.
    Brief description of amendment: The amendment would revise 
Technical Specification 5.5, ``Storage of Unirradiated and Spent Fuel'' 
to reflect a planned modification to increase the number of fuel 
assemblies that can be stored in the spent fuel pool from 2776 to 4086.
    Date of publication of individual notice in Federal Register: 
November 24, 1998 (63 FR 64973).
    Expiration date of individual notice: December 24, 1998.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Notice of 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)

[[Page 71981]]

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: October 16, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.3.1 ``Reactor Protective System (RPS) 
Instrumentation-Operating'' and TS 3.3.2, ``Reactor Protective System 
(RPS) Instrumentation-Shutdown,'' to clarify an inconsistency between 
the TS wording and the design bases as described in the TS Bases and 
the Updated Final Safety Analysis Report. Specifically, the change 
replaces the operating bypass input process variable, Thermal Power, in 
Footnotes (a), (b), and (d) of Table 3.3.1 and in the Note to Limiting 
Condition for Operation 3.3.2 with Nuclear Instrument Power.
    Date of issuance: December 8, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 229 & 204.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1998 (63 FR 
57320).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated December 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendment: April 25, 1996, as supplemented 
on September 5, 1996, August 8, 1997, March 26, July 31, and August 24, 
1998.
    Brief description of amendment: This amendment revises Technical 
Specifications (TSs) 3/4.5.F.1, ``Core and Containment Cooling 
systems'' to extend the allowed outage time (AOT) for the emergency 
diesels, TSs 3.9.B.1 and 3.9.B.4, ``Auxiliary Electrical System'' to 
reduce the AOT from 7 days to 3 days and reduce the AOT for the 
combination of an EDG and startup transformer or shutdown transformer 
from 72 hours to 48 hours, and add Configuration Risk Management 
Program in TS 5.5, ``Programs and Manuals'' of Section 5.0 
``Administrative Controls''. Various TS pages were re-numbered in 
Section 5.0. In addition, TSs 3.9, ``Auxiliary Electrical System,'' and 
3.9.A, ``Auxiliary Electrical Equipment,'' have been reformatted to be 
consistent with TS 3.9.B approved in a previous amendment. The 
associated Bases sections have also been changed to reflect the new 
TSs.
    Date of issuance: December 11, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 179.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50934).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 11, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: July 15, 1997, as supplemented 
March 3, April 13, June 16, October 26, and November 5, 1998.
    Brief description of amendments: The amendments revised the 
Technical Specifications to add new requirements for the main steamline 
break instrumentation and resolved issues related to Inspection and 
Enforcement Bulletin 80-04.
    Date of Issuance: December 7, 1998.
    Effective date: As of the date of issuance to be implemented 
coincident with implementation of the improved Technical 
Specifications.
    Amendment Nos.: 234--Unit 1; 234--Unit 2; 233--Unit 3.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50001).
    The March 3, April 13, June 16, October 26, and November 5, 1998, 
letters provided clarifying information that did not change the scope 
of the July 15, 1997, 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 December 7, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

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

    Date of application for amendment: September 24, 1998, as 
supplemented November 3, 1998.
    Brief description of amendment: This amendment revised technical 
specification 3.1.2.8 in two places to change the term ``contained 
volume'' to usable volume.'' This change eliminates the potential for a 
non-conservative interpretation of the specification values for the 
Refueling Water Storage Tank and Boric Acid Storage Tank and thereby 
eliminates the need for temporary administrative controls, which have 
been used correctly to properly interpret the specification values as 
usable volumes.
    Date of issuance: December 14, 1998.
    Effective date: Effective immediately, to be implemented within 30 
days.
    Amendment No: 95.
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59591).
    The November 3, 1998, letter did not change the initial proposed no 
significant hazards consideration determination or expand the amendment 
request beyond the scope of the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 14, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of application for amendment: August 17, 1998.

[[Page 71982]]

    Brief description of amendment: The amendment reduces the load at 
which diesel generators are tested.
    Date of issuance: December 14, 1998.
    Effective date: December 14, 1998.
    Amendment No.: 118.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53949).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 14, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.

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

    Date of application for amendments: August 1, 1997.
    Brief description of amendments: The amendments delete a portion of 
a technical specifications surveillance test requirement that specifies 
that the steam driven auxiliary feedwater pumps be tested ``when the 
secondary steam supply pressure is greater than 310 psig.'' This 
removes any misunderstanding that the secondary steam pressure must be 
just above 310 psig for this test.
    Date of issuance: December 10, 1998.
    Effective date: December 10, 1998, with full implementation within 
45 days.
    Amendment Nos.: 225 and 209.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68308).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 10, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station Unit No. 2, Oswego County, New York

    Date of application for amendment: February 5, 1998.
    Brief description of amendment: This amendment changes the 
Technical Specifications to update the terminology and references to 10 
CFR 50.55a(f) and (g) consistent with the 1989 edition of Section XI of 
the American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code, and consistent with the second 10-year interval of the Inservice 
Inspections and Inservice Testing Program Plans.
    Date of issuance: December 3, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 84
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 11, 1998 (63 FR 
11920).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 3, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendments: August 8, 1996, as supplemented 
June 30, 1997 and August 26, 1998.
    Brief description of amendments: The amendments eliminate the 
response time testing requirements for selected sensors and specified 
instrument loops for (1) the reactor protection system, (2) the 
isolation system, and (3) the emergency core cooling system.
    Date of issuance: December 14, 1998.
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 132 and 93.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57489).
    The June 30, 1997 and August 26, 1998, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 14, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: July 10, 1998, as supplemented 
October 16, 1998.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.6/4.6 and associated bases to relocate portions of 
the reactor coolant chemistry to the Updated Final Safety Analysis 
Report and to applicable plant procedures. Changes to the relocated 
requirements will be controlled by the provisions of 10 CFR 50.59.
    Date of issuance: December 1, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 247.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 29, 1998 (63 FR 
40560).
    The October 16, 1998, submittal fell with the scope of, and did not 
change, the initial proposed finding of no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: March 30, 1998, as supplemented 
on October 27, 1998.
    Brief description of amendment: The amendment revises the 
definition of logic system functional tests, and revises test frequency 
requirements for certain instrumentation.
    Date of issuance: December 11, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 248.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19978).
    The October 27, 1998, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration.

[[Page 71983]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 11, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: August 12, 1998, as 
supplemented on October 12, 1998. The October 12, 1998, letter provided 
clarifying information that did not change the initial proposed no 
sigificant hazards consideration determination.
    Brief description of amendments: The amendments revise TS 3/
4.6.1.3, ``Containment Air Locks,'' to change the action statements for 
an inoperable air lock. The amendments also revise TS Bases 3/4.6.1.2, 
``Containment Leakage,'' to correct an editorial error and TS Bases 3/
4.6.1.3, ``Containment Air Locks,'' to provide additional details 
regarding the air locks.
    Date of issuance: December 2, 1998.
    Effective date: December 2, 1998.
    Amendment Nos: 215 and 195.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48265).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 2, 1998
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: December 31, 1997, as supplemented by 
letter dated September 11, 1998.
    Brief Description of amendments: The amendments revised the 
Technical Specifications (TSs) to change the intermediate range neutron 
flux reactor trip setpoint and allowable value, and delete the 
reference to the reactor trip setpoints in TS 3.10.3, ``Special Test 
Exceptions--Physics Tests,'' and TS 3.10.4, ``Special Test Exceptions--
Reactor Coolant Loops.''
    Date of issuance: December 8, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1--140; Unit 2--132.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6998).
    The September 11, 1998, letter provided clarifying information that 
did not change December 31, 1997, application or the initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: September 20, 1996 (TS 96-09).
    Brief description of amendments: The amendments change the 
Technical Specifications to clarify the types of work shifts that are 
acceptable when considering the requirements to ensure overtime is not 
heavily used on a routine basis by unit staff.
    Date of issuance: December 7, 1998.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 240 and 230.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59596).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 7, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: August 22, 1998, as 
supplemented on August 27 and October 8, 1998 (TS 96-08). The August 
27, 1998, amendment request superseded the original (August 22, 1998) 
request in its entirety.
    Brief description of amendments: The amendments revise the Sequoyah 
Nuclear Plant Technical Specifications by extending the allowed outage 
time for the SQN emergency diesel generators from 72 hours to 7 days.
    Date of issuance: December 16, 1998.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 241 and 231.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52969), superseded by a second notice on September 9, 1998 (63 FR 
48270). The October 8, 1998, letter provided clarifying information 
that did not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 16, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: July 28, 1998, as supplemented 
October 16, 1998. The October 16, 1998, letter was administrative in 
nature and did not change the initial no significant hazards 
consideration determination.
    Brief description of amendments: The amendments revise the 
Technical Specifications to change the Emergency Diesel Generator 
section to be consistent with station procedures associated with 
steady-state conditions.
    Date of issuance: December 10, 1998.
    Effective date: December 10, 1998.
    Amendment Nos.: 216 and 197.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48272).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 10, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special

[[Page 71984]]

Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

    Dated at Rockville, Maryland, this 23rd day of December 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-34440 Filed 12-29-98; 8:45 am]
BILLING CODE 7590-01-P