[Federal Register Volume 63, Number 241 (Wednesday, December 16, 1998)]
[Notices]
[Pages 69332-69353]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-33206]



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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. Pub. L. 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 November 20, 1998, through December 4, 1998. 
The last biweekly notice was published on December 2, 1998 (63 FR 
66590).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 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 15, 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.

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    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).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.
    Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland.
    Date of amendments request: November 19, 1998.
    Description of amendments request: The proposed amendment revises 
Technical Specification 3.7.6, ``Service Water (SRW) System'' to allow 
operation of Calvert Cliffs with one SRW plate and frame heat exchanger 
(PHE) secured for maintenance or other reasons, and removing one 
containment air cooler (CAC) from service to enable the affected 
subsystem to remain operable. Specifically, the proposed change adds 
``One SRW heat exchanger inoperable'' as a new condition for Limiting 
Condition for Operation (LCO) 3.7.6. The required actions for the new 
condition are to secure one CAC within one hour and restore the heat 
exchanger to operable condition within 7 days, or be in Mode 3 in 6 
hours and Mode 5 in 36 hours. This limits the effect of one inoperable 
PHE to only one containment cooling train made inoperable by the PHE. 
Consequently, the new action statement introduced in the SRW LCO for an 
inoperable PHE is similar to the one that already exists in the CAC LCO 
for one inoperable containment cooling train.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    None of the systems associated with the proposed revision to the 
Calvert Cliffs Technical Specifications are accident initiators. The 
Saltwater (SW) and SRW systems are used to mitigate the effects of 
accidents analyzed in the Updated Final Safety Analysis Report 
(UFSAR). The SW and SRW Systems provide cooling to safety-related 
equipment following an accident. The CACs are provided with SRW to 
remove heat from the Containment in the event of an accident. They 
support accident mitigation functions; therefore, the proposed 
modification does not increase the probability of an accident 
previously evaluated.
    The proposed revision will provide greater availability of 
safety-related equipment during PHE maintenance activities. It 
ensures that the safety features provided by the SW and SRW, except 
for the isolated CAC, are maintained, i.e., the availability of 
safety-related equipment required to mitigate the radiological 
consequences of an accident described in the UFSAR is enhanced by 
the flexibility provided by this Technical Specification revision.
    Furthermore, the proposed revision will not change, degrade, or 
prevent actions described or assumed in any accident described in 
the UFSAR. The proposed activity will not alter any assumptions 
previously made in evaluating the radiological consequences of any 
accident described in the UFSAR.
    Therefore, the proposed modification does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    None of the systems associated with this modification are 
identified as accident initiators in the UFSAR. The SW and SRW 
Systems and the CACs are used to mitigate the effects of accidents 
analyzed in the UFSAR. None of these functions required of these 
systems have been changed by the proposed revision to the Technical 
Specifications. This activity does not modify any system, structure, 
or component such that it could become accident initiator, as 
opposed to its current role as an accident mitigator.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The safety design basis for the SW and SRW Systems is the 
availability of sufficient cooling capacity to ensure continued 
operation of equipment during normal and accident conditions. The 
redundant cooling capacity of these systems, assuming a single 
failure, is consistent with assumptions used in the accident 
analysis.
    With one SRW subsystem inoperable, the remaining SRW subsystem 
is adequate to perform the heat removal function. However, the 
reliability is reduced because a single failure in the operable SRW 
subsystem could result in loss of SRW function. The proposed change 
will allow continued operation of some SRW-cooled components while a 
PHE is being out-of-service. The second SRW subsystem will still be 
available to perform the SRW function. In addition, the reliability 
of many diesel generator-backed components will be improved since 
the second diesel generator will remain operable while in this 
action statement.
    During a design basis accident, a minimum of one containment 
cooling train (two of the four CACs) and one containment spray 
train, is required to maintain the containment peak pressure and 
temperature, below the design limits. Under the existing Technical 
Specification requirement, with one containment cooling train 
inoperable, the inoperable containment cooling train must be 
returned to operable status within seven days. The remaining 
operable containment spray and cooling units provide iodine removal 
capabilities and are capable of removing at least 100% of the heat 
removal needs after an accident. The seven-day completion time was 
developed taking into account the redundant heat removal 
capabilities afforded by combinations of the containment spray and 
cooling systems, and the low probability of a design basis accident 
occurring during this period. The proposed change to Technical 
Specification 3.7.6 would allow three CACs to remain operable during 
maintenance on a PHE, instead of the two that are maintained under 
the current Technical Specification requirement.

[[Page 69334]]

    For the above reasons, the margin of safety has been preserved, 
and in some cases increased, by the proposed revision to the 
Technical Specifications.
    Therefore, this proposed modification does not significantly 
reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa, Director.

    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 amendments request: November 20, 1998.
    Description of amendments request: On September 9, 1996, a final 
rule amending 10 CFR 50.55a was issued requiring owners to implement, 
by September 9, 2001, the requirements of the 1992 Addenda of the 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code Section XI, Subsections IWE and IWL, as modified and supplemented 
by 10 CFR 50.55a. Baltimore Gas and Electric Company (BGE) have 
developed a program plan to effect the implementation of Subsection IWE 
and IWL. BGE's submittal requests a license amendment in support of the 
program plan. One Technical Specification (TS) change requested is an 
administrative change that removes a TS originally developed from 
Regulatory Guide (RG) 1.35. Compliance with RG 1.35 is not sufficient 
to comply with 10 CFR 50.55a, as amended. The other TS changes request 
the removal from the TSs requirements that are a duplication of 10 CFR 
50.55a.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The Containment Building is a passive safety structure that 
prevents the release of radioactive materials to the environment in 
post-accident conditions. The proposed Technical Specification 
changes delete requirements of the Technical Specifications that 
have been made obsolete by the improvements of the Containment 
Building inspections required by the changes in the regulations. The 
improved inspections required by the American Society of Mechanical 
Engineers Code serve to maintain Containment response to accident 
conditions, by causing the identification and repair of defects in 
the Containment Buildings.
    Relocating existing requirements, eliminating requirements that 
duplicate regulations, and making administrative improvements 
provide Technical Specifications that are easier to use. Because 
existing requirements are controlled by regulation, there is no 
reduction in commitment and adequate control is still maintained. 
Likewise, the elimination of requirements that duplicate regulations 
enhances the usability of the Technical Specifications without 
reducing commitments. Therefore, the proposed changes would not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The Containment Building is a passive safety structure designed 
to contain radioactive materials released from the Reactor Coolant 
System. The performance of the Containment Building is not evaluated 
as the causal factor in any accident at Calvert Cliffs Nuclear Power 
Plant. The proposed Technical Specification changes delete 
requirements of the Technical Specifications that have been made 
obsolete by the improvements of the Containment Building inspections 
required by the changes in the regulations. Revising the Technical 
Specifications, to comply with current regulations and to eliminate 
duplication of requirements, does not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The safety function of the Containment Building is to provide a 
boundary to the release of radioactive material to the environment 
during post-accident conditions. The changes to the Technical 
Specifications incorporate improved inspection techniques and 
criterial to ensure optimum Containment integrity and, therefore, 
optimum containment response in the event of an accident resulting 
in a release of radioactive material from the Reactor Coolant 
System.
    Optimizing containment integrity will result in maintaining the 
margin of safety allowed by the Containment Buildings. Therefore, 
the proposed changes 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 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa, Director.

    CBS Corporation acting through its Westinghouse Electric Company 
Division (licensee), Westinghouse Test Reactor, Waltz Mill Site, 
Westmoreland, Pennsylvania, Docket No. 50-22, License No. TR-2.
    Date of amendment request: September 28, 1998, supplemented on 
November 17, 1998.
    Description of amendment request: CBS Corporation acting through 
its Westinghouse Electric Company Division is the licensee for the 
Westinghouse Test Reactor (WTR) at Waltz Mill, Pennsylvania. The 
licensee is authorized to only possess the reactor and a 
decommissioning plan has been approved. The licensee is planning to 
sell most of its nuclear related facilities to other entities, but will 
retain the WTR. One of the arrangements made with the purchasers of the 
other facilities is that the Westinghouse name will be conveyed with 
these facilities, and because of this arrangement, the licensee 
requests that the license associated with the Westinghouse Test Reactor 
be changed to simply CBS Corporation, to eliminate any reference to the 
name Westinghouse.
    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 
considerations. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). A proposed amendment to a 
license of a facility involves no significant hazards consideration if 
operation of the facility in accordance with the proposed amendment 
would not: (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated; or (2) create the 
possibility of a new or different kind of accident from any accident 
previously evaluated; or (3) involve a significant reduction in a 
margin of safety.
    The staff agrees with the licensee's no significant hazards 
consideration determination submitted on November 17, 1998, for the 
following reason.
    This corporate name change does not involve any change in the 
management, organization, location, facilities equipment, or procedures 
related to the licensed activities under the WTR

[[Page 69335]]

license. The employees responsible for the licensed WTR facility will 
still be responsible, either directly through the CBS Corporation or 
through contractual arrangements for which CBS Corporation is 
ultimately responsible, notwithstanding the new name of the licensee.
    Based on a review of the licensee's analysis, and on the staff's 
analysis detailed above, 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.
    Attorney for license: Lisa A. Campagna, Assistant General Counsel, 
Law Department, CBS Corporation, P.O. Box 355, Pittsburgh, Pennsylvania 
15230.
    NRC Project Director: Seymour H. Weiss.

    Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.
    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 
1 and 2, Will County, Illinois.
    Date of amendment request: October 30, 1998.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) to reduce the spent fuel pool 
(SFP) inadvertent draindown level to account for the effects of 
potential failures of the SFP cooling and skimmer loops.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change to the TS does not involve an increase in the 
probability of an accident previously evaluated. The initial 
conditions of the limiting dewatering incidents involve initiating 
circumstances/failures such as accidental gate openings, gate seal 
failures, or an open transfer tube.
    Specifying a revised inadvertent drain limit which meets the SRP 
[Standard Review Plan, NUREG-0800] acceptance criteria is unrelated 
to the probability of occurrence of the precursors or initiating 
events. These initiators are not affected by the SFP cooling or 
skimmer loop piping/component failure scenarios. There is no change 
being made to the approved design, nor is there any operational 
change being made which would increase the probability of 
occurrence.
    This change to the TS does not involve an increase in the 
consequences of an accident previously evaluated. As documented in 
NUREG-0876, Byron SER, Section 9.1.3, page 9-5, the anti-siphon 
protection design of the SFP cooling and clean-up piping was 
reviewed and found to be acceptable stating that ``all connections 
to the spent-fuel pool are either near the normal water level or are 
provided with antisiphon holes to preclude possible siphon draining 
of the pool water.'' This review is applicable to Braidwood as 
documented in NUREG-1002, Braidwood SER. The anti-siphon attributes 
employed in the SFP skimmer loops at Braidwood, (under consideration 
at Byron), are similar in design as well as their submergence levels 
previously evaluated for the SFP cooling loops. The proposed change 
revises the SFP inadvertent drain limit from approximately 423 feet 
to 410 feet to bound the failure effects of both the SFP cooling and 
skimmer loops, while considering any maloperation or failure 
scenario. The revised value meets the SRP acceptance criteria of 
maintaining at least 10 feet above the active fuel ensuring that 
adequate radiation shielding is maintained as previously analyzed. 
There is no physical or operational change being made which would 
alter the sequence of events, plant response, or conclusions of the 
affected analysis. There is no change in the type or amount of any 
effluents released, and no change in either the Onsite or Offsite 
dose consequences as a result of this change.
    Therefore, based on this evaluation, this proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This change specifically identifies the SFP level sufficient to 
ensure that the SRP acceptance criteria for inadvertent draining are 
met while accounting for the failure effects of both the SFP cooling 
and skimmer loops. Any inadvertent SFP draining due to potential 
failures of the SFP skimmer loops is similar in nature to the 
inadvertent SFP draining effects previously considered due to 
failures of the SFP cooling loops. No new equipment is being 
installed, and no installed equipment is being operated in a new or 
different manner with this change. There is no change in plant 
operation that affects previously evaluated failure modes. This 
change does not represent a new failure mode or accident from what 
has been previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The current TS value does not address inadvertent SFP draining 
due to potential failures of the SFP skimmer loops or cooling 
suction lines as was done for the SFP cooling discharge lines. This 
change specifically identifies the SFP level sufficient to ensure 
that the SRP acceptance criteria for inadvertent draining are met 
while accounting for the failure effects of both the SFP cooling and 
skimmer loops in determining the proposed TS value. The most 
limiting postulated SFP dewatering incidents involve SFP drainage to 
either a dry transfer canal, a dry transfer canal and cask fill 
area, or a dry transfer canal and cask fill area which additionally 
communicates through an open transfer tube to an empty refuel 
cavity. The initial conditions of the dewatering incident analysis 
and resultant water levels over the spent fuel are not affected by 
this SFP skimmer/cooling loop issue because these incident 
initiators are not effected by the SFP cooling or skimmer loop 
failures, thus preserving the previously analyzed and approved 
margin for these dewatering incidents.
    For the less-limiting SFP skimmer/cooling loop failure issue, 
the proposed TS change inadvertent drain limit meets the SRP minimum 
requirement of at least 10 feet above the top of the active fuel 
ensuring that adequate radiation shielding is maintained. This 
change would allow for the conservative acceptance criteria for the 
current UFSAR [Updated Final Safety Analysis Report] design analysis 
to continue to be met.
    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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

    Commonwealth Edison Company, Docket No. 50-374, LaSalle County 
Station, Unit 2, LaSalle County, Illinois.
    Date of amendment request: November 9, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.3.2, ``Isolation Actuation 
Instrumentation'' to add/revise various isolation setpoints for leak 
detection instrumentation. These changes are necessary due to 
modifications to the Reactor Water Cleanup (RWCU) System to restore 
``hot'' suction to the RWCU pumps and due to a re-evaluation of the 
high energy line break analysis. In addition, the amendment would 
eliminate isolation actuation trip functions for the Residual Heat 
Removal (RHR) system steam

[[Page 69336]]

condensing mode and shutdown cooling mode.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    (a) There is no effect on accident initiators so there is no 
change in probability of an accident. A line break in the subject 
areas, would consist of an instantaneous circumferential break 
downstream of the outermost isolation valve of one of these systems. 
The leak detection isolation is only a precursor of a break, and 
thus does not affect the probability of a break.
    (b) There is minimal effect on the consequences of analyzed 
accidents due to changing the leak detection ambient temperature or 
Delta T setpoint and allowable values to detect 25 gpm equivalent 
leakage. The addition of more ambient temperature and T 
leak detection monitoring, along with the addition of the high flow 
break detection will actually decrease the consequences of the 
associated accidents. The worst case accident outside the primary 
containment boundary is a main steam line break which bounds the 
dose consequences of all line breaks and therefore bounds any size 
of leak.
    The deletion of the RHR steam condensing mode isolation 
actuation instrumentation trip functions from the LaSalle Technical 
Specifications does not increase the probability or consequences of 
an accident previously evaluated, because this mode of operation of 
the RHR system has been deleted from the LaSalle design basis and 
the lines that were previously high energy lines are isolated during 
unit operation, including Operational Condition 1 (Run mode), 
Operational Condition 2 (Startup mode), and Operational Condition 3 
(Hot Shutdown).
    The deletion of the RHR shutdown cooling mode leak detection T 
and Delta T isolation actuation instrumentation trip functions from 
the LaSalle Technical Specifications does not increase the 
probability or consequences of an accident previously evaluated, 
because the leak detection is only a precursor of a break, and thus 
does not affect the probability of a break. Also, there are two 
other methods of detecting abnormal leakage and isolating the system 
in Technical Specification trip functions A.6.a, Reactor Vessel 
Water Level--Low, Level 3 and A.6.c, RHR Pump Suction Flow--High. In 
addition, other means to detect leakage from the RHR system, such as 
sump monitoring and area radiation monitoring, are also available. 
In accordance with Technical Specification Administrative 
Requirement 6.2.F.1, LaSalle has a leakage reduction program to 
reduce leakage from those portions of systems outside primary 
containment that contain radioactive fluids. RHR, including piping 
and components associated with the shutdown cooling mode, is part of 
this program, which includes periodic visual inspection of the 
system for leakage. The sump monitoring, radiation monitoring and 
periodic inspections for system leakage makes the probability of a 
leak of 5 gpm going undetected for more than a day very low.
    Also, due to the low reactor pressures (less than 135 psig) at 
which RHR shutdown cooling mode is able to operate, reactor coolant 
makeup and outflow is very low compared to normal plant operation. A 
change in flow balance due to a leak is thus more readily detectable 
with reactor coolant water level changes and makeup flow rate, and 
thus precludes a significant leak going undetected before break 
detection instrumentation would cause automatic isolation.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The purpose of the leak detection system, as it applies to the 
RWCU and RHR system areas, is to provide the capability for leak 
detection and automatic isolation of the system as necessary in the 
event of leakage in these areas. This change maintains this 
capability with at least two different methods of detection of 
abnormal leakage for protection from the flooding concerns of a 
significant leak or line break when the RHR system is operating in 
the shutdown cooling mode, so that redundant systems will not be 
affected.
    This change also maintains or adds primary containment isolation 
logic for the leak detection isolation based on temperature 
monitoring in RWCU areas and break detection based on RWCU pump 
suction flow--high. The additional instrumentation and the 
associated isolation logic is the same or similar to existing 
instrumentation and logic for containment actuation instrumentation, 
so no new failure modes are created in this way.
    Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) Involve a significant reduction in the margin of safety 
because:
    The change to the automatic isolation setpoint for high Delta T 
leak detection in the heat exchanger rooms is based on current 
configuration calculated/analyzed response to a small leak compared 
to a circumferential break. The increased leakage rate in the RWCU 
heat exchanger rooms that is necessary to actuate isolation on 
ambient temperature during winter conditions, does not adversely 
affect the margin of safety. This increased leakage rate is below 
the critical crack leakage rate as represented in UFSAR [Updated 
Final Safety Analysis Report] Figure 5.2-11. Additionally, 
differential temperature leak detection is conservative under these 
same conditions, and will actuate isolation at a leakage rate less 
than the established limit. The leak detection isolation logic is 
unchanged and thus remains single failure proof.
    The addition of automatic primary containment isolation on 
ambient temperature and Delta T-High for the Reactor Water Cleanup 
System (RWCU) Pump, Pump Valve, Holdup Pipe, and Filter/
Demineralizer (F/D) Valve Rooms and the addition of the RWCU Pump 
Suction Flow High line break isolation add to the margin of safety 
with respect to leak detection and line breaks in the RWCU system, 
because the system isolation diversity is increased and the amount 
of system piping monitored for leakage is increased.
    The setpoints for the ambient temperature and Delta T leak 
detection isolations being changed or added and the RWCU pump 
suction flow--high are set sufficiently high enough so as not to 
increase the possibility of spurious actuation. In the event that a 
spurious actuation does occur, little safety significance is 
presented since the RWCU system performs no safety function. The 
setpoints and allowable values for the proposed changes also assure 
sufficient margin to the analytical values and are high enough to 
prevent spurious actuations based on calculations consistent with 
Regulatory Guide 1.105.
    The deletion of the RHR steam condensing mode isolation 
actuation instrumentation does not effect the margin of safety, 
because this mode is no longer utilized by LaSalle in Operational 
Conditions 1, 2, or 3 (Run mode, Startup mode, or Hot Shutdown).
    The elimination of the temperature based trip functions for the 
RHR shutdown cooling mode area is based on the determination that 
temperature is not the appropriate parameter for leak detection as 
it does not provide meaningful indication and will not provide 
setpoints that would be sufficiently above the normal range of 
ambient conditions to avoid spurious isolations.
    There are two other methods of detecting abnormal leakage and 
isolating the system in Technical Specification trip function A.6, 
which are A.6.a, Reactor Vessel Water Level--Low, Level 3 and A.6.c, 
RHR Pump Suction Flow--High. In addition, other means to detect 
leakage from the RHR system, such as sump monitoring and area 
radiation monitoring, are also available. Also, in accordance with 
Technical Specification Administrative Requirement 6.2.F.1, LaSalle 
has a leakage reduction program to reduce leakage from those 
portions of systems outside primary containment that contain 
radioactive fluids. RHR, including piping and components associated 
with the shutdown cooling mode, is part of this program, which 
includes periodic visual inspection of the system for leakage.
    The previous evaluation of diversity of isolation parameters, as 
presented in Table 5.2-8 of the UFSAR remains unchanged. Adequate 
diversity of isolation parameters is maintained because there are at 
least two different methods available to detect and allow isolation 
of the system for a line break, as necessary.
    Therefore, 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

[[Page 69337]]

proposes to determine that the requested amendment involves no 
significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

    Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York.
    Date of amendment request: October 9, 1998.
    Description of amendment request: The proposed amendment would 
revise Section 6.0, administrative controls, of the Technical 
Specifications (TSs). Specifically, TS Sections 6.5.2.1.j, 6.7.1.c, and 
6.8.1.a would be revised to correct typographical errors. In addition, 
TS Section 6.5.2.2 would be revised to change the membership of the 
Nuclear Facility Safety Committee (NFSC). This change would provide 
Consolidated Edison (Con Ed) with the flexibility to obtain industry 
experts outside of Con Ed to perform the duties of Chairman, or Vice 
Chairman, and members of the NFSC.
    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. There is no significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment is administrative in nature. It involves 
a change in 1) the Nuclear Facilities Safety Committee (NFSC) 
Chairman or Vice Chairman to allow the services of an individual 
other than a senior official of the Company, and 2) allowing NFSC 
membership by other than Con Edison employees. In either case, 
concurrence by the Senior Vice President, Nuclear Operations is 
required.
    These changes do not affect possible initiating events for 
accidents previously evaluated or alter the configuration or 
operating of the facility. The Limiting Safety Systems Settings and 
Safety Limits specified in the current Technical Specifications 
remain unchanged. Therefore, the proposed changes to the subject 
Technical Specification would not increase the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any accident previously evaluated has not been created.
    As stated above, the proposed changes are administrative in 
nature. The safety analysis of the facility remains complete and 
accurate. There are no physical changes to the facility, and the 
plant conditions for which the design basis accidents have been 
evaluated are still valid. The operating procedures and emergency 
procedures are unaffected. Consequently, no new failure modes are 
introduced as a result of the proposed changes. Therefore, the 
proposed changes will not initiate any new or different kind of 
accident.
    3. There has been no significant reduction in the margin of 
safety.
    The proposed changes are administrative in nature. Since there 
are no changes to the operation of the facility or physical design 
the Updated Final Safety Analysis Report (UFSAR) design basis, 
accident assumptions, or Technical Specification Bases are not 
affected. Therefore, the proposed changes will not result in a 
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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: S. Singh Bajwa, Director.

    Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan.
    Date of amendment request: November 9, 1998.
    Description of amendment request: The proposed amendment would 
delete the Chemical and Volume Control System (CVCS) operability 
requirements currently in technical specifications (TS) 3.2 and 3.17.6, 
and the associated surveillance testing requirements currently in TS 
4.2 and 4.17. The requirements have been added to the Palisades 
Operating Requirements Manual (ORM).
    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:

    Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes delete certain TS requirements which do not 
meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical 
requirements have been added to a document (the ORM) controlled 
under 10 CFR 50.59.
    10 CFR 50.59 specifically prohibits changes to the facility as 
described in the safety analysis report, and to procedures described 
in the safety analysis report ``if the probability of occurrence or 
the consequences of an accident or malfunction of equipment 
important to safety previously evaluated in the safety analysis 
report may be increased''. Since the conditions which limit changes 
performed under 50.59 are more restrictive than the conditions which 
define changes considered to involve a significant hazards 
consideration, moving of a requirement from the TS to a document 
which is controlled under 50.59 cannot involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    The proposed changes delete certain TS requirements which do not 
meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical 
requirements have been added to a document (the ORM) controlled 
under 10 CFR 50.59.
    10 CFR 50.59 specifically prohibits changes to the facility as 
described in the safety analysis report, and to procedures described 
in the safety analysis report ``if a possibility for an accident or 
malfunction of a different type than any evaluated previously in the 
safety analysis report may be created''. Since the conditions which 
limit changes performed under 50.59 are more restrictive than the 
conditions which define changes considered to involve a significant 
hazards consideration, relocation of a requirement from the TS to a 
document which is controlled under 50.59 cannot create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed changes delete certain TS requirements which do not 
meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical 
requirements have been added to a document (the ORM) controlled 
under 10 CFR 50.59.
    10 CFR 50.59 specifically prohibits changes to the facility as 
described in the safety analysis report, and to procedures described 
in the safety analysis report if the margin of safety is reduced. 
Since the conditions which limit changes performed under 50.59 are 
more restrictive than the conditions which define changes considered 
to involve a significant hazards consideration, relocation of a 
requirement from the TS to a document which is controlled under 
50.59 cannot 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.

[[Page 69338]]

    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423-3698.
    Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: Cynthia A. Carpenter.

    Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina.
    Date of amendment request: July 22 and October 22, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to reflect the licensee's 
planned use of fuel supplied by Westinghouse. The Westinghouse fuel has 
different design characteristics from the fuel currently in use. 
Accordingly, the following changes would need to be made to the TS: 
Figure 2.1.1-1, ``Reactor Core Safety Limits--Four Loops in 
Operation''; various core operating parameters specified by 
Surveillance Requirements 3.2.1.2, 3.2.1.3, and 3.2.2.2; Section 4.2.1, 
``Fuel Assemblies''; and Section 5.6.5, ``Core Operating Limits Report 
(COLR).''
    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:

First Standard

    Implementation of this LAR [license amendment request] would not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. The revised Reactor Core Safety 
Limits Figure further restricts acceptable operation. Moving an 
uncertainty factor from the Improved Technical Specifications to the 
Core Operating Limits Report (COLR) does not exempt this factor from 
regulatory restrictions. COLR parameters are generated by NRC 
approved methods with the intent of ensuring that previously 
evaluated accidents remain bounding. The COLR is submitted to the 
NRC upon implementation of each fuel cycle or when the document is 
otherwise revised. No accident probabilities or consequences will be 
impacted by this LAR.

Second Standard

    Implementation of this LAR would not create the possibility of a 
new or different kind of accident from any previously evaluated. The 
revised Reactor Core Safety Limits Figure further restricts 
acceptable operation. Moving an uncertainty factor from the Improved 
Technical Specifications to the COLR does not exempt this factor 
from regulatory restrictions. Since the parameter in question is not 
being deleted, the possibility of a new or different kind of 
accident from any previously evaluated does not exist.

Third Standard

    Implementation of this LAR would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. Use of the ZIRLOTM 
cladding material has been reviewed and approved in Reference 1 (as 
listed in Chapter 2.1 of Topical Report DPC-NE-2009/DPC-NE-2009P, 
Duke Power Company Westinghouse Fuel Transition Report). 
ZIRLOTM cladding has been extensively used in 
Westinghouse nuclear reactors. The changes proposed in this LAR are 
necessary to ensure that the performance of the fission product 
barriers (cladding) will not be impacted following the replacement 
of one fuel design for another. No safety margin will be 
significantly impacted.

    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: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    NRC Project Director: Herbert N. Berkow.

    Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 
50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana.
    Date of amendment request: November 20, 1998.
    Description of amendment request: The licensee has proposed an 
amendment to Facility Operating License No. NPF-47, Appendix A--
Technical Specifications (TS) Section 3.1.6, ``Control Rod Pattern.'' 
The proposed change will be implemented through the establishment of a 
new specification added to Section 3.10, ``Special Operations.'' The 
proposed specification will be TS Section 3.10.9, ``Control Rod 
Pattern--Cycle 8.'' The new TS 3.10.9 is required due to a current 
plant-specific configuration where 5 control rods have been inserted 
into the reactor core for neutron flux suppression surrounding 2 fuel 
assemblies which have been identified as having possible fuel cladding 
defects. The new requirement is intended to be effective for the 
remainder of the current fuel cycle (Cycle 8), and is in force when rod 
withdrawal operations begin from a condition of 100% rod density to 20% 
rated thermal power (RTP).
    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 request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Accidents analyzed in the SAR have been examined for any impact 
caused by this exception to the [Banked Position Withdrawal 
Sequence] BPWS operation. The limiting event is the [Control Rod 
Drive Accident] CRDA as described in SAR Sections 4.3.2 and 15.4.9. 
The limit on energy addition to the fuel is 280 cal/gm as identified 
in the SRP section 15.4.9. Bank Position Withdrawal Sequence is 
established to reduce maximum incremental control rod worths and 
thus minimize consequences resulting from an accident. The reactor 
will be operated as before using BPWS. Having the current rod 
configuration with 5 rods to minimize impact on the two fuel 
cladding imperfections, in lieu of eight rods inoperable separated 
by two cells, will not affect initiators of a Control Rod Drop 
Accident. In addition, this existing rod configuration has been 
analyzed and the resulting consequences continue to be bounded by 
the licensing evaluations. The insertion of the identified control 
rods will not affect the assumed reactivity insertion time of any 
event. The location of the control rods has been reviewed by GE 
using the NRC approved methodology. Operation within these limits 
will ensure that the consequences of a transient or accident remain 
within the acceptable limits of the evaluation. Specifically, rod 
worths for the proposed configuration are bounded by the rod worths 
allowed for these configurations per TS; thus, the proposed 
configuration is more conservative than that allowed per TS. The 
results confirm all assumed limits are maintained. The proposed 
change ensures that the consequences of abnormal operation and 
accidents are acceptable.
    The additional Technical Specification will control the 
configuration of the plant to that supported by the evaluation. If 
this evaluated configuration is not supported, the plant will be 
required to be placed in a configuration where the Control Rod Drop 
Accident is not applicable, as the current specification requires. 
The plant is therefore maintained within limits as currently 
allowed. With these limits the consequences of an event are not 
increased.
    The probability of an accident is not affected by the proposed 
Technical Specification changes since the operation of systems or 
equipment that could initiate an accident are not affected. 
Therefore, the proposed changes do not significantly increase the 
probability or consequences of any previously evaluated accident.
    (2) The request does not create the possibility of occurrence of 
a new or different

[[Page 69339]]

kind of accident from any accident previously evaluated.
    The proposed changes do not involve any alteration of plant 
hardware or significant change in plant operation. Assuming the 5 
suppression rods are bypassed in lieu of eight rods separated by two 
cells does not affect event initiators or event consequences. No 
plant modifications are required which would affect plant operation. 
Operation with the control rod pattern in the proposed configuration 
will ensure the results of a CRDA will remain within the assumptions 
of the current safety analysis. The system will continue to ensure 
that the limits of control rod worth remain within the assumptions 
of the CRDA. The revised Technical Specifications will continue to 
assure that plant operation is consistent with the assumptions, 
initial conditions, and assumed power distribution and, therefore, 
will not create a new type of accident.
    The proposed Technical Specifications will maintain the plant in 
a configuration supported by evaluation. The response to a CRDA will 
be within current accepted limits and therefore no event of a 
different kind has been created. The proposed Technical 
Specification changes do not introduce any new modes of plant 
operation nor involve new system interactions. Therefore, operation 
with the 5 suppression rods inserted does not create the possibility 
of an occurrence of a new or different kind of accident from any 
accident previously evaluated.
    (3) The request does not involve a significant reduction in a 
margin of safety.
    The proposed Technical Specification and the rod pattern control 
system will continue to ensure the limits of control rod worth 
remain within the assumptions which support the CRDA analysis of 280 
cal/gm maximum energy heat addition to the fuel. This imposed limit 
of 280 cal/gm provides a margin of safety from the experimental 
value of approximately 330 cal/gm at which the fully molten state 
for UO2 occurs. The existing rod configuration with 5 
suppression rods inserted to minimize impact on the two fuel 
cladding imperfections has been analyzed using NRC approved 
methodology. Cycle specific evaluation has confirmed that the 
consequences resulting from a CRDA continues to be bounded by the 
licensing analysis for this event. Since there are no changes in the 
acceptance criteria, the proposed changes will not create a 
reduction in the margin of safety. These limits establish the 
necessary restrictions on power operation and thereby ensure that 
the core is operated within the assumptions and initial conditions 
of the transient and accident analyses.
    As demonstrated in the evaluation, operation within these limits 
will ensure that the margin of safety will be maintained to the same 
level described in the Technical Specifications Bases and the USAR 
and the consequences of the postulated transient or accidents are 
not increased. This limit of 280 cal/gm is not exceeded during any 
transient or postulated accident. Therefore, the proposed Technical 
Specifications to allow startup and continued operation in the low 
power region with these control rods inserted do not involve a 
significant reduction in 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: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Project Director: John N. Hannon.
    Entergy Operations Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana.
    Date of amendment request: July 2, 1998.
    Description of amendment request: The proposed change will modify 
the ACTION Requirements for Technical Specification (TS) 3/4.3.2 for 
the Emergency Feedwater Actuation Signal (EFAS). A change to the TS 
Bases Section 3/4.3.2 has been included to support this change. The 
objective of this change is to add a restriction on the period of time 
a channel of EFAS instrumentation can remain in the tripped condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No
    The proposed revision to the TS changes the allowed outage time 
that a channel of EFAS SGDPI [Steam Generator Differential Pressure 
Instrumentation] can be in the tripped condition from a maximum of 
approximately 18 months when one channel is inoperable and 92 days 
when two channels are inoperable to 48 hours. If a channel were in 
the tripped condition and a single failure occurred (failure of one 
other channel of EFAS SGDPI), an inadvertent EFAS signal would be 
generated. During a Design Basis MSLB [Main Steam Line Break] or FLB 
[Feedwater Line Break] Accident, this single failure would send EFW 
[Emergency Feedwater] to the faulted steam generator. The Waterford 
3 safety analysis assumes that the excess Reactor Coolant System 
(RCS) cooldown and return to power associated with the MSLB will be 
terminated when the faulted steam generator empties. If additional 
EFW were added, the RCS cooldown would be extended and the return to 
power may increase.
    Reducing the time that a channel of EFAS SGDPI can be placed in 
the tripped condition will reduce the probability of this scenario 
occurring during a Design Basis Accident. Since the allowed outage 
time for a channel of EFAS SGDPI is being limited to 48 hours, this 
is considered an off-normal operation and a single failure is not 
required to be postulated during a Design Basis Accident in the 
accident analysis. Reducing the time the channel can be placed in 
the tripped condition and thus, the exposure time to this scenario, 
would not be an accident initiator. The proposed change of being 
more conservative relative to allow[ed] outage time in the tripped 
condition will not affect the assumptions, design parameters, or 
results of any accident previously evaluated.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change does not alter the design or configuration 
of the plant. The proposed change provides a more conservative 
allowed outage time for the channel to be in the tripped condition. 
There has been no physical change to plant systems, structures or 
components nor will the proposed change reduce the ability of any of 
the safety-related equipment required to mitigate Anticipated 
Operational Occurrences or accidents. The configuration required by 
the proposed specification is permitted by the existing 
specification.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change provides a more conservative allowed outage 
time for the channel to be in the tripped condition. By reducing the 
allowed outage time, the probability is reduced that a single 
failure (failure of one channel of EFAS SGDPI with one channel in 
the tripped condition) would occur that would send EFW to the 
faulted steam generator. Therefore, the only change to the margin of 
safety would be an increase. Since the allowed outage time for a 
channel of EFAS SGDPI is being limited to 48 hours, this is 
considered an off-normal operation and a single failure is not 
required to be postulated during a Design Basis Accident in the 
accident analysis. The proposed changes do not affect the limiting 
conditions for operation or their bases.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.


[[Page 69340]]


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street NW., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

    Florida Power and Light Company, et al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida.
    Date of amendment request: December 31, 1997, as supplemented 
November 25, 1998.
    Description of amendment request: The proposed amendment will 
revise the St. Lucie Unit 2 Technical Specifications to permit an 
increase in the allowed Spent Fuel Pool (SFP) storage capacity.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Analyses to support the proposed fuel pool capacity increase 
have been developed using conservative methodology. The analysis of 
the potential accidents summarized below has shown that there is no 
significant increase in the consequences of any accident previously 
analyzed. A review of relevant plant operations has also 
demonstrated that there is no significant increase in the 
probability of occurrence of any accident previously analyzed. This 
conclusion is also discussed below.
    Previously evaluated accidents that were examined for this 
proposed license amendment include: Fuel Handling Accident, Spent 
Fuel Cask Drop Accident, and Loss of all Fuel Pool Cooling.
    There will be no change in the mode of plant operation or in the 
availability of plant systems as a result of this proposed change; 
the systems interfacing with the spent fuel pool have previously 
encountered borated pool water and are designed to interact with 
irradiated spent fuel and remove the residual heat load generated by 
isotopic decay. The proposed amendment does not require a change in 
the maintenance interval or maintenance scope for the fuel pool 
cooling system or for the spent fuel cask crane. The frequency of 
cask handling operations and the maximum weight carried by the crane 
is not increased as a result of the proposed license amendment. 
Thus, there will be no increase in the probability of a loss of fuel 
pool cooling or in the probability of a failure of the cask crane as 
a result of the proposed amendment.
    There will not be a significant increase in the frequency of 
handling discharged assemblies in the fuel pool as a result of this 
change; any handling of fuel in the spent fuel pool will continue to 
be performed in borated water. If the license amendment is approved, 
there will be a one-time repositioning of certain discharged 
assemblies stored in the fuel pool to comply with the revised 
positioning requirements, but the increased pool storage capacity 
will permit the deferral of spent fuel handling associated with cask 
loading operations. Fuel manipulation during the repositioning 
activity will be performed in the same manner as for fuel placed in 
the spent fuel pool during refueling outages. There will be no 
changes in the manner of handling fuel discharged from the core as a 
result of refueling; administrative controls will continue to be 
used to specify fuel assembly placement requirements. The relative 
positions of Region I and Region II storage locations will remain 
the same within the fuel pool. Therefore, the probability of a fuel 
handling accident has not been significantly increased.
    The consequences of a fuel handling accident have been 
evaluated. The radioactive release consequences of a dropped fuel 
assembly are not affected by the proposed increase in fuel pool 
storage capacity. They remain bounded by the results of calculations 
performed to justify the existing St. Lucie Unit 2 fuel storage 
racks and burnup limits. At the limiting fuel assembly burnup, 
radioactive releases from a dropped assembly would be only a small 
fraction of NRC guidelines. The input parameters employed in 
analyzing this event are consistent with the current values of fuel 
enrichment, discharge burnup and uranium content used at St. Lucie 
Unit 2 and with future use of the ``value-added'' fuel pellet 
design. Thus, the consequences of the fuel assembly drop accident 
would not be significantly increased from those previously 
evaluated.
    The capability of the fuel pool cooling system to handle the 
increased number of discharged assemblies has been examined. The 
impact of a total loss of spent fuel pool cooling flow on available 
equipment recovery time and on fuel cladding integrity has also been 
evaluated. For the limiting full core discharge, sufficient time 
remains available to restore cooling flow or to provide an alternate 
makeup source before boiloff results in a fuel pool water level less 
than that needed to maintain acceptable radiation dose levels. 
Analysis has shown that in the event of a total loss of fuel pool 
cooling fuel cladding integrity is maintained. Therefore, the 
consequences of a loss of fuel pool cooling event, including the 
effect of the proposed increase in fuel pool storage capacity, have 
not been significantly increased from previously analyzed results 
for this type of accident.
    The analysis of record pertaining to the radiological 
consequences of the hypothetical drop of a loaded spent fuel cask 
just outside the Fuel Handling Building was examined to determine 
the impact of the increased fuel storage capacity on this accident's 
results. The results of the previously performed analysis were 
determined to bound the conditions described by the proposed license 
amendment, thus the consequences of the cask drop accident would not 
be significantly increased as a result of this change.
    It is concluded that the proposed amendment to increase the 
storage capacity of the St. Lucie Unit 2 spent fuel pool will not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
    In this license amendment FPL [Florida Power & Light Co.] 
proposes to credit the negative reactivity associated with a portion 
of the soluble boron present in the spent fuel pool. Soluble boron 
has always been present in the St. Lucie Unit 2 spent fuel pool; as 
such the possibility of an inadvertent fuel pool dilution has always 
existed. However, the spent fuel pool dilution analysis demonstrates 
that a dilution of the Unit 2 spent fuel pool which could increase 
the pool keff to greater than 0.95 is not a credible 
event. Neither implementation of credit for the reactivity of fuel 
pool soluble boron nor the proposed increase in the fuel pool 
storage capacity will create the possibility of a new or different 
type of accident at St. Lucie Unit 2.
    An examination of the limiting fuel assembly misload has 
determined that this would not represent a new or different type of 
accident. None of the other accidents examined as a part of this 
license submittal represent a new or different type of accident; 
each of these situations has been previously analyzed and determined 
to produce acceptable results.
    The proposed license amendment will not result in any other 
changes in the mode of spent fuel pool operation at St. Lucie Unit 2 
or in the method of handling irradiated nuclear fuel. The spatial 
relationship between the fuel storage racks and the cask crane range 
of motion is not affected by the proposed change.
    As a result of the evaluation and supporting analyses, FPL has 
determined that the proposed fuel pool capacity increase does not 
create the possibility of a new or different type of accident from 
any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    FPL has determined, based on the nature of the proposed license 
amendment that the issue of margin of safety, when applied to this 
fuel pool capacity increase, should address the following areas:

1. Fuel Pool reactivity considerations
2. Fuel Pool boron dilution considerations
3. Thermal-Hydraulic considerations
4. Structural loading and seismic considerations

    The Technical Specification changes proposed by this license 
amendment, the proposed spent fuel pool storage

[[Page 69341]]

configuration and the existing Technical Specification limits on 
fuel pool soluble boron concentration provide sufficient safety 
margin to ensure that the array of fuel assemblies stored in the 
spent fuel pool will always remain subcritical. The revised spent 
fuel storage configuration is based on a Unit 2 specific criticality 
analysis performed using methodology consistent with that approved 
by the NRC. Additionally, the soluble boron concentration required 
by current Technical Specifications ensures that the fuel pool 
keff will be always be maintained substantially less than 
0.95.
    The Unit 2 criticality analysis established that the 
keff of the spent fuel pool storage racks will be less 
than 1.0 with no soluble boron in the fuel pool water, including the 
effect of all uncertainties and tolerances. Credit for the soluble 
boron actually present is used to offset uncertainties, tolerances, 
off-normal conditions and to provide margin such that the spent fuel 
pool keff is maintained less than or equal to 0.95. FPL 
has also demonstrated that a decrease in the fuel pool boron 
concentration such that keff exceeds 0.95 is not a 
credible event.
    Current Technical Specifications require that the fuel pool 
boron concentration be maintained greater than or equal to 1720 ppm. 
This boron value is substantially in excess of the 520 ppm required 
by the uncertainty and reactivity equivalencing analyses discussed 
in this evaluation and the 1266 ppm value required to maintain 
keff less than or equal to 0.95 in the presence of the 
most adverse mispositioned fuel assembly.
    The St. Lucie Unit 2 fuel pool boron concentration will continue 
to be maintained significantly in excess of 1266 ppm; the proposed 
license amendment will not result in changes in the mode of 
operation of the refueling water tank (RWT) or in its use for makeup 
to the fuel pool. Thus, operation of the spent fuel pool following 
the proposed change, combined with the existing fuel pool boron 
concentration Technical Specification limit of 1720 ppm, will 
continue to ensure that keff of the fuel pool will be 
substantially less than 0.95.
    Even if this not-credible dilution event was to occur, no 
radiation would be released; the only consequence would be a 
reduction of shutdown margin in the fuel pool. The volume of 
unborated water required to dilute the fuel pool to a 
keff of 0.95 is so large (in excess of 358,900 gallons to 
dilute the fuel pool to 520 ppm boron) that only a limited number of 
water sources could be considered potential dilution sources. The 
likelihood that this level of water use could remain undetected by 
plant personnel is extremely remote.
    In meeting the acceptance criteria for fuel pool reactivity, the 
proposed amendment to increase the storage capacity of the existing 
fuel pool racks does not involve a significant reduction in the 
margin of safety for nuclear criticality.
    Calculations of the spent fuel pool heat load with an increased 
fuel pool inventory were performed using ANSI/ANS-5.1-1979 
methodology. This method was demonstrated to produce conservative 
results through benchmarking to actual St. Lucie Unit 2 fuel pool 
conditions and by comparison of its results to those generated by a 
calculation using Auxiliary Systems Branch Technical Position 9-2 
methodology. Conservative methods were also used to demonstrate fuel 
cladding integrity is maintained in the absence of cooling system 
forced flow. The results of these calculations demonstrate that, for 
the limiting case, the existing fuel pool cooling system can 
maintain fuel pool conditions within acceptable limits with the 
increased inventory of discharged assemblies.
    Therefore, the proposed change does not result in a significant 
reduction in the margin of safety with respect to thermal-hydraulic 
or spent fuel cooling considerations.
    The primary safety function of the spent fuel pool and the fuel 
storage racks is to maintain discharged fuel assemblies in a safe 
configuration for all environments and abnormal loadings, such as an 
earthquake, a loss of pool cooling or a drop of a spent fuel 
assembly during routine spent fuel handling. The proposed increase 
in spent fuel inventory on the fuel pool and the existing storage 
racks have been evaluated and show that relevant criteria for fuel 
rack stresses and floor loadings have been met and that there has 
been no significant reduction in the margin of safety for these 
criteria.

    The NRC staff has reviewed the licensee's analysis and the changes 
proposed in the November 25, 1998 supplement to the original submittal 
and based on this review, it appears that the three standards of 
50.92(c) continue to be satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

    Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida.
    Date of amendment request: October 27, 1998.
    Description of amendment request: The licensee proposed to change 
Technical Specification (TS) 6.3, Facility Staff Qualifications, in 
order to incorporate qualifications for the Multi-Discipline 
Supervisor. The current TS requires that plant staff meet the 
requirements of the American National Standards Institute (ANSI) N18.1-
1971, which requires non-licensed supervisors to have a high school 
diploma or equivalent and a minimum of 4 years experience in the craft 
or discipline they supervise. The proposed change requires the Multi-
Discipline Supervisor to have, (1) a high school diploma or equivalent, 
(2) a minimum of 4 years of related technical experience, which shall 
include 3 years of power plant experience of which one year is at a 
nuclear power plant, and (3) completed the Multi-Discipline Supervisor 
training 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 facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed changes are administrative in nature addressing 
personnel qualification issues. The Multi-Discipline Supervisor 
(MDS) position will be filled with personnel who are experienced in 
one or more technical disciplines (maintenance, operations, 
engineering, or other related technical discipline). Fundamental 
working knowledge of tasks being performed will be acquired through 
the MDS initial training program. The training concentrates on 
developing the skills and knowledge of an MDS to safely oversee 
tasks for multi-discipline work teams. Therefore, four years 
experience in any related technical discipline or disciplines 
combined with the MDS training program provide adequate technical 
knowledge for proper job oversight. These proposed changes will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because they do not affect 
assumptions contained in plant safety analyses, the physical design 
and/or operation of the plant, nor do they affect Technical 
Specifications that preserve safety analysis assumptions. Therefore, 
the proposed changes do not affect the probability or consequences 
of accidents previously analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The changes being proposed are administrative in nature and do 
not affect assumptions contained in plant safety analyses, the 
physical design and/or modes of plant operation defined in the 
facility operating license, or Technical Specifications that 
preserve safety analysis assumptions. These changes address 
qualification requirements for the MDS position. Since the proposed 
changes do not change the qualifications for those individuals 
responsible for the actual licensed operation of the facility, 
operation of the facility in accordance with the proposed amendments 
would not create the possibility of a new or different kind of 
accident from any accident

[[Page 69342]]

previously evaluated. No new failure mode is introduced due to the 
administrative changes since the proposed changes do not involve the 
addition or modification of equipment nor do they alter the design 
or operation of affected plant systems, structures, or components.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are unchanged by the proposed 
amendments. The proposed changes to add the MDS position have 
management and administrative controls associated with the required 
qualification requirements. The Turkey Point Technical 
Specifications will ensure that any individual filling the MDS 
position has the requisite education, experience, and training. As a 
result, operation of the facility in accordance with the proposed 
changes would 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: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

    GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey.
    Date of amendment request: November 5, 1998.
    Description of amendment request: The proposed Technical 
Specification change will modify the safety limits and surveillances of 
the LPRM and APRM systems and related Bases pages to ensure the APRM 
channels respond within the necessary range and accuracy and to verify 
channel operability. In addition, an unrelated change to the Bases of 
Specification 2.3 is included to clarify some ambiguous language.
    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 technical specification changes to the limits 
and surveillance requirements of the LPRM and APRM systems are 
provided to ensure the APRM channels respond within the necessary 
range and accuracy and to verify channel operability. If one or more 
monitored parameters exceeded their specified limits, the RPS 
initiates a reactor scram signal to preserve the integrity of the 
fuel cladding and the Reactor Coolant System and minimize the energy 
that must be absorbed following a loss of coolant accident. 
Therefore, the probability of occurrence or the consequences of an 
accident previously evaluated in the [safety analysis report] SAR 
will not increase as a result of these changes.
    2. The proposed technical specification changes to the limits 
and surveillance requirements of the LPRM and APRM systems are 
provided to ensure the APRM channels respond within the necessary 
range and accuracy and to verify channel operability. The proposed 
changes are designed to ensure the APRM system responds in a manner 
that ensures the safety limits, limiting safety system settings, 
limiting conditions for operations, as well as design parameters for 
the APRM system and individual components are continuously met. 
Therefore, the proposed activity does not create the possibility for 
an accident or malfunction of a different type than any previously 
identified in the SAR.
    3. The proposed change does not involve a significant reduction 
in the margin of safety. When the APRMs exceed their specified 
limits, the RPS initiates a reactor scram signal to preserve the 
integrity of the fuel cladding and the Reactor Coolant System and 
minimize the energy that must be absorbed following a loss of 
coolant accident. The proposed changes are designed to assure the 
APRM system responds in a manner that ensures the safety limits, 
limiting safety system settings, limiting conditions for operations, 
as well as design parameters for the APRM system and individual 
components are continuously met. Therefore, the margin of safety 
will not be 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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    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.

    GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania.
    Date of amendment request: November 25, 1998.
    Description of amendment request: The proposed amendment will 
change the surveillance specification for Once Through Steam Generator 
(OTSG) inservice inspections for TMI-1 Cycle 13 refueling outage 
examinations which would be applicable for the next operating cycle 
only, Operating Cycle 13.
    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:

    A. The proposed changes do not represent a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed flaw disposition strategy, based on measurable eddy 
current parameters of axial and circumferential extent for Inside 
Diameter (ID) Initiated Inter-Granular Attack (IGA), will continue 
to provide high confidence that unacceptable flaws that do not have 
the required structural integrity to withstand a postulated MSLB 
[main steam line break] are removed from service. The axial and 
circumferential length limits for eddy current ID degradation 
indications meet the Draft Regulatory Guide 1. 121 * * * acceptance 
criteria for margin to failure for MSLB-applied differential 
pressure and axial tube loads. The capability for detection of flaws 
is unaffected; and the identification of tubes that should be 
repaired or removed from service is maintained. The operation of the 
OTSGs or related structures, systems, or components is otherwise 
unaffected. Therefore, neither the probability nor consequences of 
[an] SGTR [steam generator tube rupture] is significantly increased 
either during normal operation or due to the limiting loads of [an] 
MSLB accident.
    Neither the change in voltage normalization for the eddy current 
examinations, nor the administrative change in clarification of the 
reporting requirements, as described above, could significantly 
affect the probability of occurrence or consequences of any accident 
previously evaluated. These changes are administrative only.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because there are no hardware changes involved nor changes to any 
operating practices. These changes involve only the OTSG tube 
inservice inspection surveillance requirements, which could only 
affect the potential for OTSG primary-to-secondary leakage. The 
proposed changes continue to impose flaw length limits for ID IGA to 
assure tube structural and leakage integrity, as confirmed by 12R 
(and post 12R) tube pull sample examinations and pressure testing.
    In addition, neither the change in voltage normalization for the 
eddy current examinations nor the administrative change in the 
description of the reporting requirements, as described above, could 
possibly create the possibility of an accident

[[Page 69343]]

of a new or different type from any previously evaluated. These 
changes are included only to modify the plant's eddy current 
normalization to the industry standard, and clarify the reporting 
period for submittal of the OTSG inspection results to the NRC 
[Nuclear Regulatory Commission]. Therefore, these changes do not 
create the potential for any other kind of accident different from 
those that have been evaluated.
    C. These proposed changes do not involve a significant reduction 
in a margin of safety because the margins of safety defined in Draft 
Regulatory Guide 1. 121 * * * are retained. The probability of 
detecting degradation is unchanged since the bobbin coil eddy 
current methods will continue to be the primary means of initial 
detection and the probability of leakage from any indications left 
in service remains acceptably small. The strategy for dispositioning 
ID initiated IGA will continue to provide a high level of confidence 
that tubes exceeding the allowable limits for tube integrity are 
repaired or removed from service.
    In addition, neither the change in voltage normalization for the 
eddy current examinations nor the administrative change in the 
description of the reporting requirements, as described above, could 
significantly affect a margin of safety. These changes are 
administrative in nature and are included only to align TMI-1's 
voltage normalization to the industry standard, and clarify the 
reporting period, respectively.

    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.
    Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut.
    Date of amendment request: November 10, 1998.
    Description of amendment request: The proposed changes would modify 
Technical Specifications 3.3.1.1, ``Reactor Protective 
Instrumentation,'' and 3.3.2.1, ``Engineered Safety Feature Actuation 
System Instrumentation'' to restrict the time a reactor protection or 
engineered safety feature actuation channel can be in the bypass 
position to 48 hours, from an indefinite period of time. Most of these 
proposed changes were originally submitted in a letter dated May 14, 
1998. The licensee withdrew its original request and submitted a new 
request in its November 10, 1998, letter.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with 10CFR50.92, NNECO [Northeast Nuclear Energy 
Company] has reviewed the proposed changes and has concluded that 
they do not involve a significant hazards consideration (SHC). The 
basis for this conclusion is that the three criteria of 
10CFR50.92(c) are not compromised. The proposed changes do not 
involve an SHC because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to restrict the time [* * *] reactor 
protection or engineered safety feature actuation channels can be in 
the bypass position to 48 hours, from an indefinite period of time, 
has no effect on the design of the Reactor Protection System (RPS) 
or the Engineered Safety Feature Actuation System (ESFAS) and does 
not affect how these systems operate. In addition, this will 
minimize the susceptibility of these systems to the remote 
possibility of fault propagation between channels. However, this 
proposed change will require an inoperable pressurizer high pressure 
reactor protection channel to be placed in the tripped condition 
within 48 hours. With a pressurizer pressure channel in the tripped 
condition, the high failure of a second pressurizer pressure channel 
would initiate a reactor trip and open both pressurizer power 
operated relief valves (PORVs). Opening the pressurizer PORVs would 
result in an undesired loss of primary coolant. Thus, this change 
will increase the probability of occurrence of a previously 
evaluated accident. However, this would not place the plant in an 
unanalyzed condition since FSAR [Final Safety Analysis Report] 
Section 14.6.1 analyzes the inadvertent opening of both PORVs, the 
release of reactor coolant can be terminated by closure of the PORV 
block valves from the control room, and the Emergency Operating 
Procedures provide guidance on how to address this situation. 
Therefore, this change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change to increase the time a second RPS or ESFAS 
channel can be removed from service (from 2 hours to 48 hours), 
provided one of the inoperable channels is placed in the tripped 
condition, has no effect on the design of the RPS or ESFAS and does 
not affect how these systems operate. These systems will still 
function as designed to mitigate design basis accidents. However, 
this change will also impact the probability of occurrence of a 
previously evaluated accident since it will allow a second 
pressurizer high pressure reactor protection channel to be placed in 
the tripped condition for 48 hours instead of the current 2 hour 
time limit. The impact of this change is bounded by the proposed 
change to require an inoperable pressurizer high pressure reactor 
protection channel to be placed in the tripped condition after 48 
hours as previously discussed. Therefore, this change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change to apply a more restrictive action statement 
to the loss of turbine load reactor trip function has no effect on 
the design of this trip function and does not affect how this trip 
function operates. Also, this trip function is not assumed to 
operate to mitigate any design basis accident. Therefore, this 
change does not significantly increase the probability or 
consequences of an accident previously evaluated.
    The proposed change to require a channel calibration every 18 
months for the loss of turbine load reactor trip function and for 
the wide range logarithmic neutron flux monitors has no effect on 
the design of either the loss of turbine load reactor trip function 
or the wide range logarithmic neutron flux monitors. Also, neither 
of these are assumed to operate to mitigate any design basis 
accident. Therefore, this change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change to exclude the neutron detectors from the 
channel calibration requirement has no effect on the design of the 
neutron detectors and has no significant effect on how these 
detectors operate. The detectors are passive devices with minimal 
drift. In addition, slow changes in the sensitivity of the linear 
power range flux detectors is compensated for by performing the 
daily calorimetric calibration and the monthly calibration using the 
incore detectors. These detectors will still function as designed to 
mitigate design basis accidents. Therefore, this change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change to add the license amendment numbers to 
Technical Specification Page 3/4 3-9 will not result in a technical 
change to the Millstone Unit No. 2 Technical Specifications. The RPS 
will continue to function as before. Therefore, this change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change to correct the surveillance requirement 
referenced in an action statement has no effect on the design of the 
ESFAS and does not affect how this system operates. The ESFAS will 
still function as designed to mitigate design basis accidents. 
Therefore, this change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change to add a reference to the reactor coolant 
pump low speed reactor trip function to a note that states this trip

[[Page 69344]]

may be bypassed <5% power, and that the bypass must be automatically 
removed [greater than or equal to] 5% will not affect this reactor 
trip function. This bypass capability currently exists in the design 
of the Millstone Unit No. 2 RPS, and is the same bypass feature 
referenced for the reactor coolant flow low reactor trip function. 
Both of these reactor trip functions provide protection for a 
reduction in RCS [reactor coolant system] flow. The addition of this 
note will not result in any technical change to the Millstone Unit 
No. 2 RPS. The RPS will continue to function as before. Therefore, 
this change does not significantly increase the probability or 
consequences of an accident previously evaluated.
    The proposed change to correct the power level high trip 
setpoint on Technical Specification Page 2-4 will not result in any 
change to the actual plant setpoint for this RPS trip function. As a 
result of this proposed change, the setpoint listed on Page 2-4 will 
agree with the setpoint previously approved by the NRC, and 
currently used by the RPS. The change has no effect on the design of 
the RPS and does not affect how this system operates. The RPS will 
still function as designed to mitigate design basis accidents. 
Therefore, this change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The information added to the Bases of the affected Technical 
Specifications to provide a discussion of how the RPS and ESFAS are 
affected by the proposed changes, the effect the action statements 
have on the operation of the RPS and ESFAS, and to discuss the 
impact of surveillance testing on RPS operability will have no 
effect on equipment operation. The RPS and ESFAS will continue to 
function as designed to mitigate design basis accidents. Therefore, 
this change does not significantly increase the probability or 
consequences of an accident previously evaluated.
    Thus, this License Amendment Request does not impact the 
probability of an accident previously evaluated nor does it involve 
a significant increase in the consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. They will not alter assumptions 
made in the safety analysis and licensing basis. The RPS and the 
ESFAS will still function as designed to mitigate design basis 
accidents.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will not reduce the margin of safety since 
they have no impact on any safety analysis assumption. The proposed 
changes do not decrease the scope of equipment currently required to 
be operable or subject to surveillance testing, nor do the proposed 
changes affect any instrument setpoints or equipment safety 
functions.
    The effectiveness of Technical Specifications will be maintained 
since the changes will not alter the operation of any RPS or ESFAS 
function. In addition, most of the changes are consistent with the 
Calvert Cliffs RPS and ESFAS Technical Specifications model provided 
in Enclosure 3 of the NRC correspondence dated April 16, 1981 [R. A. 
Clark letter to W. G. Council, Evaluation of the Reactor Protection 
System Inoperable Channel Condition at Millstone Nuclear Power 
Station, Unit No. 2, dated April 16, 1981] and with the new, 
improved Standard Technical Specifications (STS) for Combustion 
Engineering plants (NUREG-1432).
    Therefore, there is no significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, 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.

    Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota.
    Date of amendment request: November 25, 1997, as supplemented 
September 25 and November 11, 1998. The September 25, 1998, supplement 
incorrectly references the original request as October 31, 1997, rather 
than November 25, 1997.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications for the condensate storage tank 
(CST) low level suction transfer setpoint for the high pressure coolant 
injection (HPCI) and reactor core isolation cooling (RCIC) systems to 
allow removing one CST from service for maintenance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed setpoint change and temporary level switch cross 
connection will not affect the way the suction transfer equipment 
functions, introduce new failure modes, or significantly increase 
the probability of failure of this equipment.
    A slight increase in the probability of failure of the CST 
suction low level automatic transfer function may result, however, 
during plant operation with one CST in service and the CST low level 
transfer switches temporarily cross connected. This temporary 
modification preserves the redundancy of the automatic level 
transfer logic and allows HPCI and RCIC to remain aligned to the 
condensate storage system.
    When the switches are cross connected, sections of piping and 
instrument tubing will be shared by both level switches. The 
probability that freezing or plugging of a common section of piping 
or tubing will disable both switches will be slightly higher than 
during two CST operation with the level switch piping in its normal 
configuration.
    The level switches would be cross connected at infrequent 
intervals to permit prudent and timely CST preventive maintenance 
and at the same time continue to provide HPCI and RCIC with a source 
of reactor makeup quality water. In the unlikely event of a spurious 
actuation of either system, only high quality water would be 
injected into the reactor vessel.
    Overall, the possibility of freezing or plugging of piping and 
tubing associated with the automatic transfer level switches has 
been shown to be very small, with or without the temporary level 
switch cross connection in place. During periods of operation with 
one CST, we believe the small additional opportunity for level 
instrument failure due to freezing or plugging is more than 
compensated for by the benefits of maintaining a high quality source 
of water to the HPCI and RCIC pumps.
    The proposed level switch cross connection will not affect the 
way the suction transfer equipment functions. The cross connection 
tubing will be evaluated for seismic loads equivalent to the 
existing instrument piping. Rupture of the tubing will not prevent 
the function of the level switches from being accomplished and no 
other equipment important to safety is impacted by these changes.
    Technical Specification and other specified margins of safety 
are effectively increased by the proposed changes. The HPCI/RCIC low 
CST level suction transfer level is being adjusted upward in the 
conservative direction.
    The changes do not present the opportunity for a new release 
path for radioactive material.
    These changes have no impact on the protection of the health and 
safety of the public.
    (2) The proposed amendment will not create the possibility of a 
new or different

[[Page 69345]]

kind of accident from any accident previously analyzed.
    No system, structure, or component (SSC) described in the USAR 
[Updated Safety Analysis Report] as important to safety is affected 
by these changes except for the low level CST HPCI/RCIC suction 
transfer function. Postulated malfunctions related to the proposed 
changes to the low level switches are bounded by the failure of the 
HPCI system, which has been previously evaluated in the USAR. The 
RCIC system is not relied upon to mitigate any USAR design basis 
accident.
    No new types of credible events could be identified which could 
be created by the proposed setpoint change and level switch cross 
connection. No new failure modes are associated with the proposed 
changes [sic].
    (3) The proposed amendment will not involve a significant 
reduction in the margin of safety.
    No margin of safety is reduced. Technical Specification and 
other specified margins of safety are effectively increased by the 
proposed activities. The HPCI/RCIC low CST level suction transfer 
setpoint is being adjusted upward in the conservative direction. 
Cross connecting the level switches associated with this transfer 
will preserve the redundancy built into the logic during extended 
outages of one CST. A small additional reduction in the reliability 
of the automatic transfer logic due to possible freezing or plugging 
of common instrument piping results when the level switches are 
temporarily cross connected during infrequent periods of operation 
with one CST in service. This small reduction in reliability of the 
automatic transfer function is fully compensated for by the ability 
to perform necessary and prudent preventive maintenance on the CSTs 
while at the same time supplying the HPCI and RCIC systems with 
water from the preferred high quality source.

    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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

    Northern States Power Company, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota.
    Date of amendment requests: November 25, 1998.
    Description of amendment requests: The proposed amendments would 
modify the technical specifications (TS) (TS 3.2 and Table 3.5-2B) to 
allow limited inoperability of boric acid storage tank (BAST) level 
channels and transfer logic channels to provide for required testing 
and maintenance of the associated components.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not affect any system that is a 
contributor to initiating events for previously evaluated design 
basis accidents. Therefore, the proposed changes do not involve a 
significant increase in the probability of an accident previously 
evaluated. The proposed Actions 34, 35 and 36 will allow limited 
continued plant operation with portions of BAST to RWST [refueling 
water storage tank] transfer instrumentation inoperable. However, 
because the proposed actions place time limits on inoperability 
comparable to those already approved for use in the Prairie Island 
Technical Specifications the proposed changes do not involve a 
significant increase in the consequences of an accident previously 
evaluated. The remaining proposed changes to Table TS.3.5-2B and to 
Specification 3.2B are administrative in nature. The changes to 
Table 3.5-2B incorporate design information on the BAST to RWST 
transfer instrumentation which clarifies the operability 
requirements for the instrumentation. The changes to Specification 
3.2.B add a reference to Table TS.3.5-2B. Therefore, because of the 
administrative nature of the changes, they do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not alter the design or function of any 
plant component and do not install any new or different equipment. 
The proposed changes do not alter the operation of any plant 
component in a manner which could lead to a new or different kind of 
accident. Therefore the possibility of a new or different kind of 
accident from those previously analyzed has not been created.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The proposed Actions 34, 35 and 36 will allow limited continued 
plant operation with portions of the BAST to RWST transfer 
instrumentation inoperable. However, because the proposed actions 
place time limits on inoperability comparable to those already 
approved for use in the Prairie Island Technical Specifications the 
proposed changes do not involve a significant reduction in the 
margin of safety. The remaining proposed changes to Table TS.3.5-2B 
and to Specification 3.2.B are administrative in nature. The changes 
to Table 3.5-2B incorporate design information on the BAST to RWST 
transfer instrumentation which clarifies the operability 
requirements for the instrumentation. The changes to Specification 
3.2.B add a reference to Table TS.3.5-2B. Therefore, because of the 
administrative nature of the changes, they do not involve a 
significant reduction in the margin of safety.

    NRC staff has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California.
    Date of amendment request: September 3, 1998.
    Description of amendment request: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 to change TS 3.4.9.1, ``Reactor Coolant 
System--Pressure/Temperature Limits,'' Figure 3.4-2, ``Reactor Coolant 
System Heatup Limitations--Applicable Up to 12 EFPY,'' and Figure 3.4-
3, ``Reactor Coolant System Cooldown Limitations--Applicable Up to 12 
EFPY,'' to extend the applicability up to 16 effective full power years 
(EFPY). The affected TS Bases would also be appropriately revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to Figures 3.4-2 and 3.4-3 of Technical 
Specification (TS) 3.4.9.1 and the associated Bases adjust the 
reactor

[[Page 69346]]

coolant system (RCS) heatup and cooldown pressure/temperature (P/T) 
limits to permit operation through 16 effective full power years 
(EFPY). The 16 EFPY P/T limits are more restrictive than the current 
limits; this accounts for an expected incremental increase in 
reactor vessel embrittlement, and assures the reactors will continue 
to be operated within acceptable stresses and at temperatures for 
which the reactor vessel metal exhibits ductile properties. The P/T 
limits developed for 16 EFPY were determined in accordance with 10 
CFR 50, Appendix G, and maintain the same margins of safety as the 
current limits. The proposed changes will not impact the probability 
of overpressurization or brittle fracture of the vessel, and 
therefore will not impact the consequences of an accident.
    The present low temperature overpressure protection (LTOP) 
pressure and enable temperature setpoints were reviewed and found to 
be acceptable and conservative for use through 16 EFPY, based on use 
of ASME Code Case N-514, which provides acceptable margins to the 
prevention of vessel overpressurization and brittle fracture. 
Therefore, there is no change to the consequences of accidents 
previously analyzed. Since no changes are proposed in the actual 
LTOP setpoints, nor any physical alteration of the LTOP system, nor 
a change to the method by which the LTOP system performs its 
function, there would be no change to the probability of an accident 
previously evaluated. The proposed change to the Bases incorporates 
use of ASME Code Case N-514, which will benefit DCPP by not 
resulting in a reduced RCS P/T window and reduced power-operated 
relief valve (PORV) pressure setpoint for LTOP. This maintains the 
current level of operator flexibility during heatup and cooldown, 
and prevents an increase in the probability of an accident 
associated with an inadvertent PORV actuation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to TS 3.4.9.1, ``Reactor Coolant System--
Pressure/Temperature Limits,'' do not involve any physical 
alteration to any plant system or change the method by which any 
safety-related system performs its function. The changes to TS 
3.4.9.1 account for the effects of an incremental increase in 
reactor vessel embrittlement and are requested in order to restrict 
future reactor operation to within acceptable stress levels and 
temperature regimes in accordance with 10 CFR 50, Appendix G, 
requirements. These changes are needed to maintain the current P/T 
limit margins of safety as defined by 10 CFR 50, Appendix G, and 
ASME XI, Appendix G, for operation through 16 EFPY. The possibility 
of a new kind of accident such as catastrophic failure of the 
reactor vessel is prevented by maintaining acceptable margins of 
safety.
    The present LTOP pressure setpoint was reviewed and found to be 
acceptable and conservative for the extension of the P/T curves to 
16 EFPY.
    Additionally, the proposed changes will not affect the ability 
of the LTOP system to provide pressure relief at low temperatures, 
thereby maintaining the LTOP design basis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes to TS 3.4.9.1 adjust the RCS heatup and 
cooldown P/T limits to permit operation through 16 EFPY. The P/T 
limits have been determined in accordance with 10 CFR 50, Appendix 
G, and include the safety margins with regard to brittle fracture 
required by the ASME Section XI, Appendix G, which maintain the same 
margins of safety as the current limits.
    The LTOP setpoints were reevaluated using the requirements of 
ASME Code Case N-514. This code case was developed to provide the 
necessary margins of safety for the prevention of reactor vessel 
overpressurization and brittle fracture. The LTOP evaluation results 
conclude the current LTOP setpoints are conservative for operation 
through 16 EFPY. In addition, avoiding an unnecessary reduction in 
the LTOP, the PORV pressure setpoint prevents an increase in the 
likelihood of an inadvertent PORV actuation.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room Location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas & 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    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: October 29, 1998.
    Description of amendment request: The proposed change will relocate 
Technical Specification 3/4.7.9 requirements for Snubbers and the 
associated Bases to the Technical Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates requirements and surveillances for 
Technical Specification 3/4.7.9 that do not meet the criteria for 
inclusion in Technical Specifications as identified in 10 CFR 
50.36(c)(2)(ii). The affected components are not assumed to be 
initiators of analyzed events and are not assumed to mitigate 
accident or transient events. The requirements and surveillances for 
these affected systems and components will be relocated from the 
Technical Specifications to the Technical Requirements Manual, which 
is incorporated in the STP UFSAR and will be maintained pursuant to 
10 CFR 50.59. In addition, the Snubber operability is 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. The associated changes to the Index are administrative. 
Therefore, the change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change relocates requirements and surveillances 
applicable to snubbers which 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. The associated changes to the Index are administrative. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change relocates requirements and surveillances for 
snubbers, that do not meet the 10 CFR 50.36(c)(2)(ii) criteria for 
inclusion in Technical Specifications. The change will not reduce a 
margin of safety since it has no impact on any safety analysis 
assumptions. In addition, the relocated requirements and 
surveillances for the affected structure, system, component, or 
variable remain the same as the existing Technical Specifications. 
Since any future changes to these requirements or the surveillance 
procedures will be evaluated per the requirements of 10 CFR 50.59, 
there will be no reduction in a margin of safety. The associated 
changes to the Index are administrative and have no potential effect 
on the margin of safety.

[[Page 69347]]

    The proposed change is also consistent with the Westinghouse 
Plants Standard Technical Specification, NUREG-1431 approved by the 
NRC Staff, revising the Technical Specifications to reflect the 
approved content 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, NW, Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

    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: October 29, 1998.
    Description of amendment request: The proposed change will relocate 
Specification 3/4.3.4, ``Turbine Overspeed Protection,'' to the 
Technical Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates the requirements of specification 
3/4.3.4, ``Turbine Overspeed Protection,'' that do not meet the 
criteria for inclusion in Technical Specifications as identified in 
10 CFR 50.36(c)(2)(ii). The specification is not related to any 
assumed initiators of analyzed events and are not assumed to 
mitigate accident or transient events. The requirement to perform 
the testing is not altered by the proposed change. The requirements 
of the limiting condition for operation and surveillance testing 
will be relocated from the Technical Specifications to the Technical 
Requirements Manual, which is incorporated in the STP UFSAR and will 
be maintained pursuant to 10 CFR 50.59. In addition, the 
surveillance testing details 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. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change relocates the requirements of specification 
3/4.3.4, ``Turbine Overspeed Protection,'' that do 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. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change relocates the requirements of specification 
3/4.3.4, ``Turbine Overspeed Protection,'' that do not meet the 10 
CFR 50.36 criteria for inclusion in Technical Specifications. The 
change will not reduce a margin of safety since it has no impact on 
any safety analysis assumptions. In addition, the relocated 
requirements applicable to the turbine overspeed protection remain 
the same as the existing Technical Specifications requirements. 
Since any future changes to these requirements or the surveillance 
procedures will be evaluated per the requirements of 10 CFR 50.59, 
there will be no reduction in a margin of safety.
    The proposed change is also consistent with the Westinghouse 
Plants Standard Technical Specification, NUREG-1431 approved by the 
NRC Staff. Revising the Technical Specifications to reflect the 
approved content, 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, NW, Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

    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: October 29, 1998.
    Description of amendment request: The proposed change will relocate 
descriptive details of Surveillance Requirement 4.8.1.1.2.g, regarding 
maintenance of the diesel generator fuel oil storage tanks (DGFOSTs), 
to the Technical Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates descriptive details of 
surveillance requirement 4.8.1.1.2.g that do not meet the criteria 
for inclusion in Technical Specifications as identified in 10 CFR 
50.36(c)(3). The affected descriptive testing details are not 
related to any assumed initiators of analyzed events and are not 
assumed to mitigate accident or transient events. The requirement to 
perform the testing is not altered by the proposed change. The 
descriptive details of the surveillance testing will be relocated 
from the Technical Specifications to the Technical Requirements 
Manual, which is incorporated in the STP UFSAR and will be 
maintained pursuant to 10 CFR 50.59. In addition, the surveillance 
testing details 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. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change relocates descriptive details of 
surveillance testing applicable to the DGFOSTs, which do not meet 
the criteria for inclusion in Technical Specifications as identified 
in 10 CFR 50.36(c)(3). 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. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change relocates descriptive details of the 
surveillance testing applicable to the DGFOSTs, that do not meet the 
10 CFR 50.36 criteria for inclusion in Technical Specifications. The 
change will not reduce a margin of safety since it has no impact on 
any safety analysis assumptions. In addition,

[[Page 69348]]

the relocated surveillance testing details for the DGFOSTs remain 
the same as the existing Technical Specifications. Since any future 
changes to these requirements or the surveillance procedures will be 
evaluated per the requirements of 10 CFR 50.59, there will be 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 Specifications to 
reflect the approved NUREG-1431 content 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, NW, Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri.
    Date of application request: October 31, 1997, as supplemented by 
letter dated September 29, 1998. This notice supersedes the staff's 
proposed no significant hazards consideration determination evaluation 
for the requested changes that was published on January 14, 1998 (63 FR 
2283).
    Description of amendment request: The proposed amendment 
application would change Tables 3.3-3, 3.3-4, and 4.3-2 of the 
technical specifications (TS) to revise the engineered safety feature 
actuation system (ESFAS) Functional Unit 6.f, Loss of Offsite Power-
Start Turbine-Driven Pump. Table 3.3-2 would be revised to create 
separate functional units for the analog and digital portions of the 
ESFAS function associated with starting the turbine-driven auxiliary 
feedwater pump (TDAFP) upon a loss of offsite power. Table 3.3-4 would 
be revised to create separate functional units for the analog and 
digital portions of the ESFAS function associated with starting the 
TDAFP upon a loss of offsite power. Table 4.3-2 would be revised to 
create separate functional units for the analog and digital portions of 
the ESFAS function associated with starting the TDAFP upon a loss of 
offsite 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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The recognition that different 
operability and surveillance requirements apply to analog vs. 
digital circuitry does not impact any previously analyzed accidents. 
The proposed change will not affect any of the analysis assumptions 
for any of the accidents previously evaluated. The proposed change 
does not alter the current method or procedures for meeting the 
surveillance requirements in Table 4.3-2. The proposed change will 
not affect the probability of any event initiators nor will the 
proposed change affect the ability of any safety-related equipment 
to perform its intended function. There will be no degradation in 
the performance of nor an increase in the number of challenges 
imposed on safety-related equipment assumed to function during an 
accident situation. Therefore, the proposed 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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The separation of analog and digital portions of 
Functional Unit 6.f will not impact the normal method of plant 
operation.
    The operability requirements, ACTION Statement, and surveillance 
requirements for the analog portion, new Functional Unit 6.f.1), are 
identical to those of Functional Unit 8.a. The requirements for the 
digital portion, new Functional Unit 6.f.2), are consistent with the 
current Technical Specifications, other than the new ACTION 
Statement 39 provisions that eliminate the transient imposed on the 
plant from a 3.0.3 shutdown and the performance of a refueling 
interval TADOT [Trip Actuating Device Operational Test]. There is no 
safety benefit associated with shutting the plant down under LCO 
3.0.3, if both logic trains were inoperable, when considering the 
fact that the pump is allowed to be inoperable for 72 hours. This 
unnecessary shutdown would be detrimental to plant safety. The 
``new'' TADOT requirement is a reflection of current plant testing 
practice. These changes do not change any ESFAS design standards and 
are appropriate for digital functions such as this. No new accident 
scenarios, transient precursors, failure mechanisms, or limiting 
single failures are introduced 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.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the acceptance criteria for 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection functions. There will be no impact 
on any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: William H. Bateman.

    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri.
    Date of application request: July 30, 1998.
    Description of amendment request: The proposed amendment 
application would change Table 4.3-2 of the technical specifications 
(TS) by adding a table notation to clarify that verification of the 
time delays associated with engineered safety feature actuation system 
(ESFAS) Functional Units 8.a and 8.b, ``Loss of Power,'' is only 
performed as part of the channel calibration.
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The protection systems will continue 
to function in a manner consistent with the plant design basis. The 
proposed change will not affect any of the analysis assumptions for 
any of the accidents previously evaluated. Neither the Trip 
Setpoints and Allowable Values in Technical Specification Table 3.3-
4 nor the response times listed in FSAR [Final Safety

[[Page 69349]]

Analysis Report] Table 16.3-2 are affected. The proposed change will 
not affect the probability of any event initiators nor will the 
proposed change affect the ability of any safety-related equipment 
to perform its intended function. There will be no degradation in 
the performance of nor an increase in the number of challenges 
imposed on safety-related equipment assumed to function during an 
accident situation. There will be no change to normal plant 
operating parameters or accident mitigation capabilities. Therefore, 
the proposed 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.
    There are no hardware changes associated with this license 
amendment nor are there any changes in the method by which any 
safety-related plant system performs its safety function. The normal 
manner of plant operation is unchanged. Verification of the time 
delays need not be performed on a monthly basis when response time 
testing is performed on an alternating 18 month basis per the 
provisions of Technical Specifications 4.3.1.2 and 4.3.2.2 and the 
verification of LOCA [loss-of-coolant accident] and shutdown 
sequencer timing and analog channel time constant calibrations are 
performed on a refueling frequency. No new accident scenarios, 
transient precursors, failure mechanisms, or limiting single 
failures are introduced as a result of this change. There will be no 
adverse effect or challenges imposed on any safety-related system 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.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits or limiting safety system settings are determined nor will 
there be any effect on those plant systems necessary to assure the 
accomplishment of protection functions. There will be no impact on 
the overpower limit, DNBR [Departure from Nucleate Boiling Ratio] 
limits, FQ, Nuclear Enthalpy Rise Hot Channel Factor, 
LOCA PCT [Peak Clad Temperature], peak local power density, or any 
other 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: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: William H. Bateman.

    Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339. 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia.
    Date of amendment request: November 18, 1998.
    Description of amendment request: The proposed amendments would 
make changes to the North Anna Power Station (NAPS), Unit 1 and 2, 
Technical Specifications (TS) Surveillance Requirement (SR) 4.7.13.1, 
``Groundwater Surveillance Requirements'' and related Table 3.7-6, 
``Allowable Groundwater Levels--Service Water Reservoir.'' The change 
in the SR requests that the measuring device numbers assigned to 
piezometers be eliminated from the TS SR in order to avoid redundancy, 
and eliminate confusion as well as the need to initiate TS changes 
whenever new piezometers are added, older devices are replaced or 
abandoned in-place. The proposed change in groundwater threshold levels 
will raise the allowable groundwater levels to those consistent with 
the allowable levels in the ``Stability of Service Water Reservoir 
(SWR) Slope Under Increased Phreatic Surface'' calculations.
    Basis for proposed no significant hazards consideration 
determination: as required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards, which is 
presented below:

    Specifically, operation of the North Anna Power Station in 
accordance with the proposed TS Change Request will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, since: (a) 
removing non-safety related SWR piezometer device numbers from the 
TS and raising TS allowable groundwater surface threshold elevation 
levels in the southeast section of the SWR will have no effect on 
the way the safety-related Service Water System was designed to 
operate, (b) Periodic Test Procedures will continue to identify all 
open-tube piezometers and require that they be monitored in order to 
obtain as much information as possible regarding changing 
groundwater levels, (c) sufficient redundancy will continue to exist 
since at least two (2) open-tube (standpipe-type) piezometers, not 
subject to mechanical failure, have been installed in each of the 
three (3) SWR zones to meet the TS Surveillance Requirement that 
``at least one measurement per zone be available'' and (d) recent 
calculations have confirmed that raising the allowable water level 
in the southeast section of the SWR will not affect the stability of 
the SWR dike as indicated in the original design basis calculation.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, since: (a) the frequency of 
piezometer monitoring and the intent of monitoring groundwater 
surface threshold elevations in order to maintain stability of the 
SWR slope have not changed, (b) no physical modification to the 
plant or new mode of plant operation is involved, (c) changes are 
consistent with the assumptions made in the Safety Analyses and 
original design basis calculation and (d) failure of the SWR dike 
and ensuing loss of service water was the most serious accident 
postulated and considered credible. Operation of the SWR is not 
being changed. Therefore, a new or different kind of accident is 
[not] created by the change in groundwater level. In addition, since 
both the SWR and Lake Anna reservoir provide redundant sources of 
service water, failure of the SWR is not considered as a credible 
accident.
    3. Involve a significant reduction in a margin safety, since: 
(a) increasing the allowable phreatic surface in the SE section of 
the SWR dike will not lower the factor of safety with respect to the 
stability of the SWR as defined by the original design basis 
calculation, (b) the margin to failure of the SWR dike has been 
proven by calculation to have not been reduced as defined by the 
original design basis calculation and (c) subject changes will not 
impact the performance of structures, systems or components relied 
upon for accident mitigation or any safety analysis assumptions, 
therefore the margin of safety is not changed by the proposed 
[change] in groundwater level at the SWR.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for Licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Project Director: Herbert N. Berkow.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the

[[Page 69350]]

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

    Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina.
    Date of application for amendment: February 27, 1997, as 
supplemented August 24, 1998.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 3/4.4.5, ``Steam Generators,'' by adding sleeve 
installation as an alternative to tube plugging for repairing degraded 
steam generators.
    Date of issuance: November 23, 1998.
    Effective date: November 23, 1998.
    Amendment No.: 85.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17225).
    The August 24, 1998, supplemental letter provided clarifying 
information only, and 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 November 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

    Commonwealth Edison Company, Docket No. 50-254, Quad Cities Nuclear 
Power Station, Unit 1, Rock Island County, Illinois.
    Date of application for amendment: August 14, 1998, as supplemented 
by letters dated October 13 and November 23, 1998.
    Brief description of amendment: The amendment changes the Quad 
Cities Technical Specifications (TS) to reflect the use of Siemens 
Power Corporation ATRIUM-9B fuel. Specifically the amendment 
incorporates the following into the TS: (a) new methodologies that will 
enhance operational flexibility and reduce the likelihood of future 
plant derates, (b) administrative changes that eliminate the cycle 
specific implementation of ATRIUM-9B fuel and adopt Improved Standard 
Technical Specification language where appropriate, and (c) changes to 
the Minimum Critical Power Ratio.
    Date of issuance: December 3, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 182.
    Facility Operating License No. DPR-29: The amendment revised the 
TSs. Public comments requested as to proposed no significant hazards 
consideration: Yes (63 FR 59588 dated November 4, 1998). This notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by December 4, 1998, but indicated that if the Commission makes 
a final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final no significant hazards consideration 
determination are contained in a Safety Evaluation dated December 3, 
1998.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

    Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida.
    Date of application for amendment: October 31, 1997, as 
supplemented December 13, 1997, February 27 and April 24, 1998.
    Brief description of amendment: The amendment proposed to revise 
the Final Safety Analysis Report (FSAR) to reflect changes to the 
credited methodology for boron precipitation prevention, as approved by 
the NRC.
    Date of issuance: November 30, 1998.
    Effective date: November 30, 1998.
    Amendment No.: 171.
    Facility Operating License No. DPR-72: Amendment revised the 
Operating License to reflect the change to the FSAR.

    Date of initial notice in Federal Register: November 12, 1997 (62 
FR 60731). The supplemental letters contained clarifying information 
that did not change the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

    Florida Power and Light Company, et al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida.
    Date of application for amendment: October 29, 1998.
    Brief description of amendment: The amendment revised the 
terminology used in the St. Lucie Plant Technical Specifications (TS) 
relative to the implementation and automatic removal of certain 
protection system trip bypasses to ensure that the meaning of explicit 
terms used in the TS are consistent with the intent of the stated 
requirements.
    Date of Issuance: November 24, 1998.
    Effective Date: November 24, 1998.
    Amendment No.: 98.
    Facility Operating License No. NPF-16: Amendment revised the TS.
    Date of initial notice in Federal Register: November 5, 1998 (63 FR 
59809).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

    GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear

[[Page 69351]]

Generating Station, Ocean County, New Jersey.
    Date of application for amendment: July 21, 1998.
    Brief description of amendment: The amendment (1) revises Technical 
Specification (TS) 6.2.2.2(a) to provide flexibility to accommodate 
unexpected absence of on-duty shift crew members, (2) eliminates 
reference to the Manager, Plant Operations in Specification 6.2.2.2(j) 
as the position has been eliminated, (3) reduces the maximum time in 
which to forward audit reports to the responsible manager from 60 days 
to 30 days, (4) replaces the term ``Vice President'' with the term 
``Corporate Officer'' in several places in Section 6, and (5) corrects 
several typographical errors.
    Date of Issuance: November 30, 1998.
    Effective date: November 30, 1998, to be implemented within 30 days
    Amendment No: 203.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 26, 1998 (63 FR 
45525).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated November 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

    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: October 8, 1998.
    Brief description of amendments: The amendments would revise the 
Technical Specification Section 3.4.1.3, ``Reactor Coolant System--
Shutdown,'' and its associated bases to provide separate requirements 
for the Reactor Coolant system in MODE 4, MODE 5 with the reactor 
coolant loops filled, and MODE 5 with the reactor coolant loops not 
filled.
    Date of issuance: November 27, 1998.
    Effective date: November 27, 1998, with full implementation within 
30 days.
    Amendment Nos.: 224 and 208.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1998 (63 FR 
57322).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 27, 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. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York.
    Date of application for amendment: June 19, 1998, as supplemented 
November 6, 1998.
    Brief description of amendment: This amendment changes Technical 
Specification 3.2.2 and the associated Bases to update pressure-
temperature operating curves and tables for continued plant operation 
up to 28 effective full-power years.
    Date of issuance: November 25, 1998.
    Effective date: As of the date of issuance to be implemented before 
core operation exceeds 18 effective full-power years.
    Amendment No.: 164.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 29, 1998 (63 FR 
40557)
    The November 6, 1998, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
November 25, 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. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York.
    Date of application for amendment: November 25, 1998, as 
supplemented November 27, 1998.
    Brief description of amendment: This change adds a note to certain 
specific containment isolation valves listed in Table 4.4-1. The note 
permits the licensee to operate Indian Point Unit 3 for the remainder 
of the current cycle (Cycle 10) without pneumatic leakage rate testing 
of these isolation valves. These valves have been leakage rate tested 
in the past using water pressurized with nitrogen gas. Without this 
emergency amendment, there would have had to delay its resumption of 
plant operation at power until the Technical Specifications required 
test was performed.
    Date of issuance: November 27, 1998.
    Effective date: As of the date of issuance to be implemented 
immediately.
    Amendment No.: 184.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, and final determination 
of no significant hazards consideration, are contained in a Safety 
Evaluation dated November 27, 1998.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

    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: August 3, 1998, as supplemented 
October 20, 1998.
    Brief description of amendment: The amendment provides for 
application of the existing minimum critical power ratio safety limit 
to Cycle 14 operation.
    Date of issuance: November 25, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 246.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48264).
    The October 20, 1998, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 25, 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.

    Southern California Edison Company, et al., Docket No. 50-362, San 
Onofre Nuclear Generating Station, Unit No. 3, San Diego County, 
California.
    Date of application for amendment: September 22, 1998.

[[Page 69352]]

    Brief description of amendment: The proposed amendment would modify 
the Technical Specifications (TS) to change the parameter used to 
establish and remove the bypasses for high reactor power trips. The 
parameter would be changed from the current ``THERMAL POWER'' to 
logarithmic power.
    Date of issuance: November 23, 1998.
    Effective date: November 23, 1998.
    Amendment Nos.: 136.
    Facility Operating License No. NPF-15: The amendments revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56259).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

    Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama.
    Date of application for amendments: June 12 and August 14, 1998 
(TS-390).
    Brief description of amendments: Changes the technical 
specifications (TS) to accommodate surveillance intervals to be 
compatible with a 24-month fuel cycle.
    Date of issuance: November 30, 1998.
    Effective date: November 30, 1998.
    Amendment Nos.: 235, 255, 215.
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the TS.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48269).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 30, 1998.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611.

    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 21, 1996 (TS 96-03).
    Brief description of amendments: The amendments revise the SQN 
Technical Specification (TS) 3.7.1.3 to extend the limiting condition 
for operation of the condensate storage tanks to Mode 4 when steam 
generator is relied upon for heat removal.
    Date of issuance: November 19, 1998.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 238 and 228.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TSs.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52967).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 19, 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: April 30, 1998 (TS 98-01).
    Brief description of amendments: The amendments revise the SQN 
Technical Specification Surveillance Requirement 4.4.3.2.1.b by 
changing the mode requirement to allow power-operated relief valve 
stroke testing in Modes 3, 4, and 5 with a steam bubble in the 
pressurizer rather than only in Mode 4.
    Date of issuance: November 19, 1998.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 239 and 229.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38204).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 19, 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 No. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee.
    Date of application for amendment: May 6, as supplemented June 5, 
1998.
    Brief description of amendment: The requested changes would allow 
an increase in the limit, up to 5.0 percent, for the U-235 enrichment 
of new (unirradiated) fuel stored in the new fuel storage racks and 
limit the fuel storage locations to assure that k-effective values are 
met.
    Date of issuance: December 1, 1998.
    Effective date: December 1, 1998.
    Amendment No.: 15.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43214).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 1, 1998.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.

    The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear, 
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 50-
440 Perry Nuclear Power. Plant, Unit 1, Lake County, Ohio.
    Date of application for amendment: September 3, 1998.
    Brief description of amendment: This amendment revised Technical 
Specification 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,'' 
by increasing the Division 3 Diesel Generator fuel oil level 
requirements to account for (1) a rounding error in the calculation, 
and (2) the unusable volume due to vortex formation at the eductor 
suction nozzle located in the fuel oil storage tank.
    Date of issuance: November 23, 1998.
    Effective date: November 23, 1998.
    Amendment No.: 94.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53960).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.
    The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear, 
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 50-
440 Perry Nuclear Power Plant, Unit 1, Lake County, Ohio.
    Date of application for amendment: August 28, 1997.
    Brief description of amendment: This amendment revised Pressure-

[[Page 69353]]

Temperature (P/T) Limits contained in Technical Specification 3.4.11 as 
a result of the Reactor Vessel Material Surveillance Program 
Requirements contained in Appendix H of 10 CFR Part 50.
    Date of issuance: December 2, 1998.
    Effective date: December 2, 1998.
    Amendment No.: 95.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61846).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

    Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin.
    Date of application for amendment: April 15, 1998 as supplemented 
by letters dated August 13, 1998, September 28, 1998, and November 24, 
1998.
    Brief description of amendment: The amendment incorporates changes 
to TS 2.1, ``Safety Limits'' and TS 3.10, ``Control Rod and Power 
Distribution Limits.'' These changes revise the power distribution 
peaking factor limits and limits operating parameters related to the 
Minimum Departure from Nucleate Boiling Ratio (MDNBR) in support of 
cycle 23 fuel and reload changes. A change associated with the fuel and 
reload changes, is the removal, from the current licensing basis, of 
the fuel pool turbine missile hazards analysis
    Date of issuance: December 2, 1998.
    Effective date: December 2, 1998.
    Amendment No.: 142.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1998 (63FR25120 
).
    The supplemental submittals did not affect the initial 
determination of no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

    Dated at Rockville, Maryland, this 9th 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. 98-33206 Filed 12-15-98; 8:45 am]
BILLING CODE 7590-01-P