[Federal Register Volume 65, Number 125 (Wednesday, June 28, 2000)]
[Notices]
[Pages 39956-39966]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-16193]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from June 3, 2000, through June 16, 2000. The 
last biweekly notice was published on June 14, 2000 (65 FR 37420).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By July 28, 2000, 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 electronically from 
the ADAMS Public Library component on the NRC Web site,
http://www.nrc.gov (the Electronic Reading Room). 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

[[Page 39957]]

hearing. The petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner intends to rely to establish those facts or expert opinion. 
Petitioner must provide sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or fact. 
Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. 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 electronically from the ADAMS Public 
Library component on the NRC Web site,
http://www.nrc.gov (the Electronic Reading Room).

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

    Date of amendment request: June 14, 2000.
    Description of amendment request: The requested amendment proposes 
to revise Technical Specification (TS) 5.6.5 to incorporate analytical 
methodologies that are used for core operating limits that have been 
accepted by NRC for referencing in licensing applications.
    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:

    Carolina Power & Light (CP&L) Company has evaluated the proposed 
TS change and has concluded that it does not involve a significant 
hazards consideration. The conclusion is in accordance with the 
criteria set forth in 10 CFR 50.92. The bases for the conclusion 
that the proposed change does not involve a significant hazards 
consideration are discussed below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes in a methodology have been previously 
generically reviewed and approved for use by the NRC for determining 
core operating limits. Analyzed events are assumed to be initiated 
by the failure of plant structures, systems, or components. The core 
operating limits developed in accordance with the new methodologies 
are bounded by the limitations in the NRC acceptance in its safety 
evaluations of the new methodologies. The topical reports associated 
with the new methodologies demonstrate that the integrity of the 
fuel will be maintained during normal operations and that design 
requirements will continue to be met. The proposed change does not 
have a detrimental impact on the integrity of any plant structure, 
system, or component. The proposed change will not alter the 
operation of any plant equipment, or otherwise increase its failure 
probability. Therefore, the probability of occurrence for a 
previously analyzed accident is not significantly increased.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fuel during the analyzed accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. The proposed change to methodology continues 
to meet applicable design and safety analyses acceptance criteria. 
The topical reports associated with the new methodologies 
demonstrate that the integrity of the fuel will be maintained as is 
assumed or is bounded initially in accident analyses. The proposed 
change does not affect the performance of any equipment used to 
mitigate the consequences of an analyzed accident. As a result, no 
analyses assumptions are violated and there are no adverse effects 
on the factors that contribute to offsite or onsite dose as the 
result of an accident. The proposed change does not affect setpoints 
that initiate protective or mitigative actions. The proposed change 
ensures that plant structures, systems, or components are maintained 
consistent with the safety analysis and licensing bases. Based on 
this evaluation, there is no significant increase in the 
consequences of a previously analyzed event.
    Therefore, the proposed change does not involve any increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures, or components. The proposed changes in 
methodology continue to meet applicable criteria for MSLB [main 
steamline break] and LBLOCA [large break loss-of-coolant accident] 
analysis and assure that appropriate criteria are used in future 
safety analyses to establish the acceptability of reload batch fuel 
with regard to mechanical properties. The proposed change does not 
involve a physical alteration of the plant other than allowing for 
fuel design in accordance with NRC approved methodologies. No new or 
different equipment is being installed. No installed equipment is 
being operated in a different manner. There is no alteration to the 
parameters within which the plant is normally operated or in the 
setpoints that initiate protective or mitigative actions. As a 
result no new failure modes are being introduced. There are no 
changes in the methods governing normal plant operation, nor are the 
methods utilized to respond to plant transients altered. Therefore, 
the proposed 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 margin of safety is established through the design of the 
plant structures, systems, and components, through the parameters 
within which the plant is operated, through the establishment of the 
setpoints for the actuation of equipment relied upon to respond to 
an event, and through margins contained within the safety analyses. 
The

[[Page 39958]]

proposed change in the methodologies used for MSLB and LBLOCA 
analyses and the use of the generic design criteria for PWR 
[pressurized-water reactor] fuel designs does not impact the 
condition or performance of structures, systems, setpoints, and 
components relied upon for accident mitigation. The proposed change 
does not significantly, impact any safety analysis assumptions or 
results. Therefore, the proposed change does not result in a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 26, 2000.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Sections 3/4.3.7.1, ``Radiation 
Monitoring Instrumentation,'' 3/4.7.2, ``Control Room and Auxiliary 
Electric Equipment Room Emergency Filtration System,'' and 6.2.F.8, 
``Ventilation Filter Testing Program,'' to eliminate habitability 
system requirements associated with the Auxiliary Electric Equipment 
Room habitability systems.
    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 changes involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The elimination of Auxiliary Electric Equipment Room (AEER) 
habitability system requirements does not affect the precursors or 
initiators of any accidents previously evaluated.
    The current analysis assumes an operator will maintain 
continuous occupancy of the AEER for 30 days following a design 
basis loss-of-coolant-accident (LOCA). This analysis credits 
operation of the AEER habitability system. The resultant dose to the 
operator is within the limits of 10 CFR 50, Appendix A, ``General 
Design Criteria for Nuclear Power Plants,'' General Design Criterion 
(GDC) 19, ``Control Room.'' We have performed an evaluation that 
determined an operator has more than sufficient time to perform all 
required actions in the AEER following a design basis LOCA, when 
directed by the station's emergency operating procedures (EOPs), 
without taking credit for the AEER habitability system and still 
maintain the resultant dose within the limits of GDC 19.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Do the changes create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    The proposed changes do not effect the operation or 
configuration of plant systems, structures, or components. These 
proposed changes do not affect currently analyzed failure modes and 
do not introduce new failure modes.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Do the changes involve a significant reduction in a margin of 
safety?
    The proposed changes will require an operator to be present in 
the AEER in a post-LOCA environment only when necessary to perform 
required actions as directed by the station's EOPs. A time/motion 
study of required AEER actions has determined that the maximum 
cumulative time spent in the AEER is approximately 300 minutes. The 
dose to operators performing the required AEER actions, without 
credit for the AEER filtration system, will continue to be within 
the limits of GDC 19, during and following all design basis 
accidents.
    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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

GPU Nuclear, Inc. and Saxton Nuclear Experimental Corporation, Docket 
No. 50-146, Saxton Nuclear Experimental Facility (SNEF), Bedford 
County, Pennsylvania

    Date of amendment request: April 10, 2000.
    Description of amendment request: The proposed amendment would make 
changes to the organizational and administrative controls for the SNEF 
to reflect changes in GPU Nuclear, Inc. following the sale of the 
Oyster Creek Nuclear Generating Station. The proposed changes to the 
technical specifications (TSs) would (1) replace reference to the 
President of GPU Nuclear and division Vice Presidents with a GPU 
Nuclear Cognizant Officer, (2) replace reference to ``other GPU Nuclear 
personnel'' with ``other GPU Inc, personnel,'' (3) replace reference to 
the ``Radiation Safety Committee'' with the ``TMI2/SNEC Oversight 
Committee,'' (4) replace ``GPU Nuclear audit program procedures'' with 
``approved Quality Assurance Plan procedures,'' and (5) make changes to 
the TSs to reflect changes to NRC organization.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    GPUN has determined that Technical Specifications Change request 
No. 60 involves no significant hazards consideration as defined in 
10 CFR 50.92.
    1. The proposed changes to the SNEC Technical Specifications do 
not involve a significant increase in the probability of occurrence 
or consequences of an accident or malfunction of equipment important 
to safety previously analyzed in the safety analysis report. The 
changes have no impact on plant operations or the release of 
radioactive materials.
    2. The proposed changes to the SNEC Technical Specifications 
will not create the possibility for an accident or malfunction of a 
different type than any previously evaluated in the safety analysis 
report because no plant configuration or operational changes are 
involved.
    3. The changes will not involve a significant reduction in the 
margin of safety as defined in the basis for any technical 
specification for SNEC because no change to operational limits will 
be made.

    The NRC staff has reviewed the analysis of the licensees 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.
    Attorney for the Licensee: Ernest L. Blake, Jr., Shaw, Pittman, 
Potts, and Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Branch Director: Ledyard B. Marsh.

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

    Date of amendment request: February 3, 2000.
    Description of amendment request: NNECO's proposed license 
amendment

[[Page 39959]]

request of February 3, 2000, would add a note to the Millstone 3 Final 
Safety Analysis Report (FSAR) to indicate that the configuration of 
relief valve 3CHS*V62 and isolation valve 3CHS*V61 takes exception to 
American Society of Mechanical Engineers (ASME) Section III code 
requirements for class 2 components. The change does not affect 
existing plant design but rather changes licensing basis information in 
the FSAR to accurately reflect plant configuration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's review is presented 
below:

    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The revision to the Final Safety Analysis Report (FSAR) to 
correctly reflect the current valve configuration to the Chemical 
Volume and Control System (CVCS) will not affect the ability of the 
CVCS to perform its intended safety function. Therefore, this change 
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.
    Since there are no changes in components, component operation, 
or system operation, this change does not create the possibility of 
an accident of a different type.
    3. Involve a significant reduction in a margin of safety.
    Since the FSAR revision does not have anything to do with 
affecting the ability of the CVCS to perform its intended safety 
function, it will not involve a significant reduction in a margin of 
safety.

    Based on the staff's analysis, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: May 12, 2000.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification Section 4.6.E.1.d safety/relief 
valve (SRV) bellows monitoring system test frequency from quarterly to 
once per operating cycle.
    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 amendment will have no impact on the probability or 
consequences of an accident. The BLDS [bellows leak detection 
system] performs a monitoring function only and is not part of the 
reactor pressure boundary.
    The reduced testing frequency for the leak detection monitoring 
function will have no impact on the ability of the pressure switch 
to detect a bellows failure or on the likelihood of bellows failure. 
Experience has shown the pressure switch to be reliable and capable 
of performing its function.
    Reduction in test frequency to once per cycle will still provide 
periodic verification of pressure switch capability. Reduction in 
test frequency to once per cycle will reduce the number of times per 
cycle that SRV operability is impacted by the testing process. This 
will increase the probability that SRV's [sic] would be available to 
mitigate consequences of an accident.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed amendment has the potential to improve reliability 
of the BLDS by removing a requirement which will allow removal of a 
failure path. A reduction in BLDS surveillance test frequency will 
not result in creation of a new or different kind of accident. The 
BLDS performs a monitoring function only. It cannot cause an 
accident as it is not part of the reactor pressure boundary.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Revising the requirement to test this system from quarterly to 
once per cycle will not reduce the margin of safety. The pressure 
switch and pressure boundary components of the BLDS are reliable and 
stable. Therefore, the proposed Technical Specification change does 
not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: March 2, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.6.3, ``Containment Isolation 
Valves.'' The proposed change deletes the asterisk (*) modifying the 
word OPERABLE in the Limiting Condition for Operation and relocates its 
associated footnote at the bottom of the page to immediately following 
the Action Statement. The new note would be reworded to be consistent 
with the wording of NUREG-1431, ``Standard Technical Specifications, 
Westinghouse Plants.'' The Bases associated with this TS would also be 
revised to address the proposed change.
    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 current Salem Technical Specifications allows the use of 
administrative means to unisolate a containment isolation valve on 
an intermittent basis. The proposed change eliminates the potential 
for varying interpretations of the TS footnote by relocating it to 
the ACTION section of the Technical Specifications in accordance 
with the guidance of NUREG 1431, Rev 1 (April 1995) ``Standard 
Technical Specifications Westinghouse Plants (NUREG-1431).'' PSE&G 
[PSE&G] views the proposed change as a change that is editorial in 
nature.
    The proposed change does not delete any existing surveillance 
requirements or delete any requirements from the Limiting Condition 
for Operations (LCOs) or Action Statements, and therefore does not 
reduce the actions that are currently taken in the TS to demonstrate 
operability of plant structures, systems, or components (SSCs). The 
proposed change continues to ensure the operability of the 
containment isolation valves, therefore ensuring that the 
containment atmosphere will be isolated from the outside environment 
in the event of a release of radioactive material to the containment 
atmosphere or pressurization of the containment.
    Since these changes do not modify any SSCs or reduce the current 
requirements for demonstrating operability of these SSCs, the 
proposed changes to the TS do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
in the Safety Analysis Report (SAR).

[[Page 39960]]

    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment eliminates the potential for varying 
interpretations of the TS footnote by relocating it to the ACTION 
section of the Technical Specifications in accordance with the 
guidance of NUREG 1431, Rev 1 (April 1995) ``Standard Technical 
Specifications Westinghouse Plants (NUREG-1431).''
    The proposed change does not alter the physical configuration of 
the plant. The proposed change does not affect any systems, 
structures or components assumed to function in the accident 
analysis, or creates a new or different accident scenario. The 
proposed change to the TS does not affect the ability of the plant 
systems to meet their current TS requirements or design basis 
functions. Therefore, the proposed change does not increase the 
consequences of a malfunction of equipment important to safety 
previously evaluated in the SAR or 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 amendment eliminates the potential for varying 
interpretations of the TS footnote by relocating it to the ACTION 
section of the Technical Specifications in accordance with the 
guidance of NUREG 1431, Rev 1 (April 1995) ``Standard Technical 
Specifications Westinghouse Plants.'' The proposed amendment does 
not change any testing acceptance criteria or modify any protective 
trip setpoint. The proposed change will continue to ensure that the 
containment atmosphere will be isolated from the outside environment 
in the event of a release of radioactive material to the containment 
atmosphere or pressurization of the containment.
    There is no reduction in the current surveillance requirements 
required to demonstrate the operability of plant SSCs. Therefore, 
the proposed changes do not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Nuclear Business Unit--
N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 13, 2000.
    Description of amendment request: The proposed amendments would 
delete Technical Specification (TS) 3/4.1.3.2.2 which is related to 
shutdown and control rod group demand position indication in modes 3, 
4, and 5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS change does not involve any physical changes to 
plant structures, systems or components (SSC). Shutdown margin will 
continue to be maintained as required by plant Technical 
Specifications to ensure the reactor will be maintained sufficiently 
subcritical to preclude inadvertent criticality in the shutdown 
condition. Shutdown and control rod group demand position indication 
is not required to ensure adequate shutdown margin in modes 3, 4 and 
5 and therefore cannot contribute to the initiation of any accident. 
The proposed changes do not change or alter the design assumptions 
for the systems or components used to mitigate the consequences of 
an accident, and the initial conditions and methodologies used in 
the accident analyses remain unchanged. Therefore, accident analyses 
results are not impacted. 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.
    The proposed changes do not involve any physical changes to 
plant structures, systems or components. The safety functions of the 
related structures, systems, or components are not changed in any 
manner, nor is the reliability of any structures, systems, or 
components reduced. No new or different type of equipment will be 
installed by this requested change. Therefore, no new failure modes 
or potential accident initiators are introduced. Therefore, the 
proposed amendments 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.
    Shutdown margin will continue to be maintained in accordance 
with the requirements of TS 3/4.1.1. The reactor will be maintained 
sufficiently subcritical to preclude inadvertent criticality in the 
shutdown condition. Therefore, the proposed amendments do not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Nuclear Business Unit--
N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: June 1, 2000.
    Description of amendment request: The proposed amendments would 
revise the vessel pressure and temperature limit curves that are in the 
Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The changes to the calculational methodology for the pressure 
and temperature (P/T) limits based upon Code Cases N-640 and N-588 
continue to provide adequate margin in the prevention of a non-
ductile type fracture of the reactor pressure vessel (RPV). The code 
cases were developed based upon the knowledge gained through years 
of industry experience. P/T curves developed using the allowances of 
Code Cases N-640 and N-588 indeed yield more operating margin. 
However, the experience gained in the areas of fracture toughness of 
materials and pre-existing undetected defects show that some of the 
existing assumptions used for the calculation of P/T limits are 
unnecessarily conservative and unrealistic. Therefore, providing the 
allowances of the subject code cases in developing the P/T limit 
curves will continue to provide adequate protection against 
nonductile-type fractures of the RPV.
    The evaluation for extending the Unit 1 and Unit 2 P/T limit 
curves to 54 EFPYs was performed using the approved methodologies of 
10 CFR 50, Appendix G, and with the allowances of code cases N-588 
and N-640. The curves generated from these methods ensure the P/T 
limits will not be exceeded during any phase of reactor operation. 
Therefore, the probability of occurrence and the consequences of a 
previously analyzed event are not significantly increased. Finally, 
the proposed changes will not affect any other system or piece of 
equipment designed for the prevention or mitigation of previously 
analyzed events.
    Thus, the probability of occurrence and the consequences of any 
previously analyzed event are not significantly increased as the 
result of the proposed changes.
    2. Do the proposed changes create the possibility of a new or 
different type of accident from any previously evaluated.

[[Page 39961]]

    The proposed changes provide more operating margin in the P/T 
limit curves for inservice leakage and hydrostatic pressure testing, 
non-nuclear heatup and cooldown, and criticality, with the benefits 
being primarily realizable during the pressure tests. The revised 
curves also extend the P/T limit curves to 54 EFPYs. However, 
operation in the ``new'' regions of the curves have been analyzed 
with the new P/T curves providing adequate protection against a 
nonductile-type fracture of the RPV. Otherwise, the proposed changes 
do not result in any new or unanalyzed operation of any system or 
piece of equipment important to safety, and as a result, the 
possibility of a new type event is not created.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    As mentioned previously, the revised P/T curves provide more 
operating margin and thus, more operational flexibility than the 
current P/T curves. With the increased operational margin, a 
reduction in the safety margin results with respect to the existing 
curves. However, the industry experience since the inception of the 
P/T limits in 1974 confirms that some of the existing methodologies 
used to develop P/T curves are unrealistic and unnecessarily 
conservative. Accordingly, ASME Code Cases N-640 and N-588 take 
advantage of the acquired knowledge by establishing more realistic 
methodologies for the development of P/T curves. Therefore, 
operational flexibility is gained and an acceptable margin of safety 
to RPV non-ductile type fracture is maintained.
    The extension of the P/T curves to 54 EFPYs was performed per 
the guidelines of 10 CFR 50, and using code cases N-640 and N-588 
and thus, the margin of safety is not significantly reduced as the 
result of the proposed changes.

    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.
    Attorney for licensee: Ernest L. Blake, Jr., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Richard L. Emch, Jr.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: March 3, 2000.
    Description of amendment request: The proposed amendments would 
revise technical specification (TS) 3.9.4, ``Containment 
Penetrations'', by allowing the equipment hatch to be open during core 
alterations and/or during movement of irradiated fuel within the 
containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes will allow the equipment hatch to be 
open during core alterations and movement of irradiated fuel 
assemblies inside containment. The existing [Vogtle Electric 
Generating Plant] VEGP TS allow the air lock doors to be open during 
core alterations and movement of irradiated fuel assemblies inside 
containment, and the dose analyses for a fuel handling accident 
inside containment remain bounding for the case of [an open 
equipment hatch]. The proposed changes will not alter the manner in 
which fuel is handled or core alterations are performed. Therefore 
the proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes do not create any new failure modes for 
any system or component, nor do they adversely affect plant 
operation. No new equipment will be added and no new limiting single 
failures will be created. The plant will continue to be operated 
within the envelope of the existing safety analyses. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    No. The previously determined radiological dose consequences for 
a fuel handling accident inside containment with the air lock doors 
open remain bounding for the proposed changes. These previously 
determined dose consequences were determined to be well within the 
limits of 10 CFR 100 and they meet the acceptance criteria of 
[Standard Review Plan] SRP Section 15.7.4 and [General Design 
Criteria] GDC 19. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: May 17, 2000.
    Brief description of amendments: The proposed amendment would 
change the Allowable Values specified in Technical Specification Table 
3.3.5-1 to ensure that the 6.9 kilovolt (kV) and 480 volt (V) 
undervoltage relays initiate the necessary actions when required. In 
addition, some unnecessary limits would be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed License Amendment Request includes more restrictive 
Allowable Values for the Preferred offsite source bus undervoltage 
function, the Alternate offsite source bus undervoltage function, 
the 6.9 kv Class 1E bus loss of voltage function, the 6.9 kv Class 
1E bus degraded voltage function, and the 480 V Class 1E bus 
degraded voltage function. These more restrictive values assure that 
all applicable safety analysis limits are being met. The 480 V low 
grid undervoltage relay allowable value is being lowered to the same 
as the 480 V degraded voltage relays which matches its function. 
This is a less restrictive value but the value still assures that 
all applicable safety analysis limits are being met. Lowering of the 
480 V low grid undervoltage allowable value will minimize 
unnecessary actuations that could challenge plant systems. Changing 
the 6.9 kV and 480 V degraded voltage, 480 V low grid undervoltage, 
the 6.9 kV loss of voltage, and the preferred and alternate bus 
undervoltage Allowable Values in the Technical Specifications has no 
impact on the probability of occurrence of any accident previously 
evaluated. Because all accident analyses continue to be met, these 
changes do not impact the consequences of any accident previously 
evaluated.
    Removal of the upper limits for the preferred and alternate bus 
undervoltage and the lower limit for the 6.9 kV Class 1E bus loss of 
voltage relays does not impact the probability of occurrence of any 
accident previously evaluated. None of the accident analyses are 
affected, therefore, the consequences of all previously evaluated 
accidents remain unchanged.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    None of the changes affect plant hardware or the operation of 
plant systems in a way

[[Page 39962]]

that could initiate an accident. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    There were no changes made to any of the accident analyses or 
safety analysis limits as a result of this proposed change. Further, 
the proposed change does not affect the acceptance criteria for any 
analyzed event. Removal of the upper limits for the preferred and 
alternate source bus undervoltage and the lower limit for the 6.9 kV 
Class 1E bus loss of voltage relays does not change the margin of 
safety. Each allowable value, as revised, assures the safety 
analysis limits assumed in the safety analyses as discussed in 
Chapter 15 of the FSAR [Final Safety Analysis Report] is maintained. 
The margin of safety established by the Limiting Conditions for 
Operation also remains unchanged. Thus there is no effect on the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Morgan, Lewis and Bockius, 
1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: May 25, 2000 (ULNRC-04258).
    Description of amendment request: The proposed amendment would 
expand (1) the range of acceptable lift settings for the pressurizer 
safety valves (PSVs), and (2) the tolerance (from +1% to +2%) of the 
as-found, measured lift settings of tested PSVs, to be operable. The 
as-left lift settings, following testing, of the PSVs would not be 
changed from the current range of +1%. The amendment would revise 
Technical Specifications (TS) 3.3.2, ``Engineered Safety Features 
Actuation System (ESFAS) Instrumentation,'' 3.4.10, ``Pressurizer 
Safety Valves,'' and 3.4.11, ``Pressurizer Power Operated Relief Valves 
(PORVs),'' of the Callaway TS. For TS 3.3.2, a new Action H for one or 
more trains inoperable would be added, the note for surveillance 
requirement (SR) 3.3.2.14 would be revised to identify another slave 
relay that the SR would be applicable to, and the automatic PORV 
actuation would be added to Table 3.3.2-1, ``Engineered Safety Features 
Actuation System Instrumentation.'' For TS 3.4.10, the range of 
allowable PSV lift settings in the limiting condition for operation 
(LCO) would be expanded from >2460 and 2510 to >2411 and 2509, and SR 
3.4.10.1 would be revised to state that following testing, the lift 
settings shall be ``within 1% of 2460 psig'' instead of simply ``within 
1%.'' The nominal PSV lift setting would be changed from 2485 psig to 
2460 psig because the maximum PORV lift setting would not be increased 
and the minimum setting would be reduced 59 psig. For TS 3.4.11, 
Actions A and B would be revised to be actions for inoperable PORVs 
either solely due to excessive PORV seat leakage (Action A) or for 
reasons other than excessive seat leakage (Action B), and Action E 
would remain an action for two inoperable PORVs, but would be only for 
reasons other than excessive seat leakage. The licensee also provided 
corrections to the Bases of the TSs and the Callaway Final Safety 
Analysis Report (FSAR) for the above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The pressurizer safety valves (PSVs), in conjunction with the 
Reactor Trip System (RTS), provide overpressure protection for the 
Reactor Coolant System (RCS). The PSV [lift] setpoint is established 
to maintain the RCS pressure below 110% of the system design 
pressure. The proposed change in the minimum allowable PSV setpoint 
could result in a transient being terminated at a pressure that is 
lower than that assumed in the transient's analysis. However, the 
primary system pressure boundary is not challenged by the minimum 
allowable PSV setpoint. Since the maximum allowable PSV setpoint is 
unaffected by the proposed change (other than from round-off, as 
discussed previously [in the application, from 2510 to 2509 psig]), 
the primary system pressure boundary is not challenged by the 
maximum allowable PSV setpoint.
    With a nominal setpoint of 2460 psig and a [as-found] +2% 
setpoint tolerance, the PSV actuation setpoint could potentially 
open at pressures as low as 2410 psig (rounded up in revised LCO 
3.4.10 to 2411 psig). This lower PSV actuation setpoint will reduce 
the margin between the pressurizer PORV and PSV actuation setpoint 
from 125 psi to 75 psi. A 75 psi margin is considered adequate and 
should not challenge the PSVs on Condition I transients.
    The majority of the Callaway PRA [probabilistic risk assessment] 
event trees question the capability of the PORVs to open for RCS 
cooldown and depressurization or for feed and bleed cooling. Some 
event trees question the capability of the PORVs to reclose to 
terminate RCS depressurization and coolant inventory loss. The 
transient-induced ATWS [anticipated transient without scram] event 
trees question the capability of the PSVs to reclose after opening 
for these high pressure transients. The maximum allowable PSV 
setpoint is essentially unchanged; therefore, the proposed change 
will not adversely impact the probability of the PSVs failing open. 
Upgrading the automatic PORV actuation circuitry to fully Class 1E, 
and revising the Technical Specification operability and 
surveillance requirements to demonstrate the operability of the 
automatic PORV actuation circuitry, will enhance valve reliability 
and assure compliance with NRC Generic Letter 90-06. However, it has 
been determined that this plant modification increase the 
probability that the PORVs will inadvertently open and remain open 
if multiple transmitter failures are postulated. With the new safety 
grade PORV 2/4 [two out of four] opening actuation logic, two failed 
high pressurizer pressure channels would result in inadvertent 
opening of both PORVs and the PORVs would remain open until remote-
manually closed. Since two of the four channels available to reclose 
the PORVs are assumed to have failed high, and since closure of the 
PORVs would require a 3/4 logic to close after the modification is 
implemented, there would be no signal to close the PORVs on a low 
pressurizer pressure signal. With the current opening logic, a 
single failed high pressurizer pressure channel would result in 
opening one PORV. However, the current 2/4 closure logic would 
reclose that PORV when pressurizer pressure drops below 
approximately 2200 psia. With the current control logic, three 
failed high pressurizer pressure channels (3/4) are required for 
both PORVs to inadvertently open and remain open. However, the 
consequences of both PORVs inadvertently opening and remaining open 
are bounded by the analysis in FSAR Section 15.6.1, ``Inadvertent 
Opening of a Pressurizer Safety or Relief Valve.'' Since a 
pressurizer safety valve is sized to relieve approximately twice the 
steam flow rate of a pressurizer PORV, and will therefore allow a 
much more rapid depressurization upon opening, the analysis in 
Section 15.6.1 examines the accidental depressurization of the RCS 
associated with an inadvertent opening of a pressurizer safety 
valve. While there is no way to isolate a stuck-open pressurizer 
safety valve, two open PORVs can be remote-manually isolated by 
either closing the PORVs or the PORV block valves. Since there is a 
small impact due to multiple channel failures resulting in an 
increase in the probability of both PORVs inadvertently opening and 
remaining open, it is concluded that the proposed activity increases 
the probability of occurrence of an accident previously evaluated in 
the FSAR. However, multiple failures are required for this 
malfunction and failure modes that result in multiple channels 
failing high are highly unlikely. Therefore, this increase in the 
probability that the PORVs will inadvertently open and remain open 
is considered to be insignificant.

[[Page 39963]]

    All evaluations performed for overpressure transients 
conservatively assume the upper limit of the PSV tolerance as the 
pressure to which the RCS is subjected. It has been determined that 
the design transients are not adversely affected because the 
limiting transients are not sensitive to the pressure tolerance 
change. Although the lower PSV setpoint would result in a lower PSV 
relief flow rate, the slightly lower valve flow rate would be more 
than compensated for by the reduced valve opening pressure. The 
change to the PSV setpoint and setpoint tolerance does not change 
the conclusions of the existing thermal-hydraulic and stress 
analyses for the pressurizer safety and relief system. The design 
function of the valves is not being changed and the conclusions 
documented in the NRC Safety Evaluation of Callaway's response to 
NUREG-0737 Item II.D.1[``Performance Testing of the Pressurizer 
Power-Operated Relief Valve,''] (dated September 10, 1987) are 
unchanged (see also FSAR Section 18.2.5). The PORVs and associated 
discharge piping can accommodate water relief.
    Overall protection system performance will remain within the 
assumptions of the previously performed accident analyses since the 
only hardware changes are associated with making the automatic PORV 
actuation circuitry fully Class 1E. The RTS and Engineered Safety 
Features Actuation System (ESFAS) protection systems will continue 
to function in a manner consistent with the plant design basis. The 
automatic PORV actuation circuitry modification will be performed in 
such a manner that all design, material, and construction standards 
that were applicable to safety-related systems prior to the change 
are maintained.
    The proposed change will not affect the probability of any event 
initiators nor will the proposed change negatively affect the 
ability of any safety-related equipment to perform its intended 
function. Changing the PSV lift setting does not change the 
probability that an event will occur which will result in the PSV 
opening. There will be no degradation in the performance of safety-
related equipment assumed to function during an accident situation. 
There will be no change to normal plant operating parameters.
    Since the FSAR Chapter 15 LOCA [loss-of-coolant accident], SGTR 
[steam generator tube rupture] and MSLB [main steam line break] 
analyses all result in decreasing RCS pressure and do not challenge 
the PSV opening pressure, none of these events are affected by the 
proposed change to the PSV nominal setpoint and the allowable 
setpoint tolerance. Timely operator actions will be taken to 
preclude water relief through the PSVs during an Inadvertent ECCS 
[emergency core cooling system] Actuation at Power event. Water 
relief from the PORVs for the latter event would result in a larger 
discharge of RCS inventory than currently analyzed, wherein operator 
action is assumed to terminate safety injection within 10 minutes 
prior to the pressurizer filling. However, FSAR Figure 15.5-3 in 
Attachment 5 [to the application] demonstrates that DNB [departure 
from nucleate boiling] is not a concern, there will be no fuel 
failures associated with this event, and RCS inventory will be 
directed to the pressurizer relief tank located inside containment. 
Therefore, there will be no impact on offsite radiological 
consequences. None of the other non-LOCA transients are adversely 
affected by the proposed change. Since none of the other FSAR 
Chapter 15 events are adversely affected, the radiological 
consequences of those events are not adversely affected.
    In the Westinghouse reanalysis of the Inadvertent ECCS Actuation 
at Power event, the minimum PSV opening setpoint serves as a limit 
to demonstrate the acceptability of the assumed operator action 
times to assure that the PSVs will not be required to operate while 
the pressurizer is water solid. A lower PSV opening setpoint could 
potentially require earlier operator actions to prevent water relief 
through the PSVs. Simulator exercises for the Inadvertent ECCS 
Actuation at Power event were performed on the Callaway training 
simulator on August 10, 1999 to determine the times required for the 
control room operators to stop the NCP [normal charging pump] and 
unblock the PORVs and assure their availability for automatic 
pressure relief. In all cases, the NCP was stopped within four (4) 
minutes and the PORVs were unblocked and available for automatic 
pressure relief within seven (7) minutes. The reanalysis in 
Attachment 5 [to the application] conservatively credits operator 
actions from the main control room to stop the NCP in six (6) 
minutes and to unblock the PORVs and assure their availability for 
automatic pressure relief in nine (9) minutes. These times include 
all process and instrumentation delays. The revised FSAR Figure 
15.5-2 shows that if operator actions are taken within these time 
frames to terminate NCP flow and to assure at least one PORV is 
available for automatic pressure relief, water relief through the 
PSVs is precluded. Procedure changes and periodic operator 
requalification training will provide assurance that these operator 
actions can be performed within the assumed time constraints.
    Based on the above discussions, the proposed change will not 
involve a significant increase in the probability of occurrence or 
the consequences of an accidently 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 nominal setpoint for the PSVs will be lowered by 1% from 
2485 psig to 2460 psig. The allowable setpoint tolerance will be 
increased from +1% to +2%. The combined effect of these changes 
results in a 2% decrease in the minimum acceptable PSV [lift] 
setpoint from 2460 psig to 2411 psig. The change in the PSV setpoint 
and in the tolerance of the setpoint does not change their ability 
to open on demand. The maximum acceptable PSV setpoint is unaffected 
by this proposed change, other than round-off as discussed 
previously. Since the FSAR accident analyses do not rely on the 
automatic actuation of non-safety related control grade systems or 
components for accident mitigation, a plant modification will make 
the automatic pressurizer PORV pressure relief circuitry fully Class 
1E.
    The proposed change to the PSV nominal setpoint and the 
allowable setpoint tolerance will not prevent the PSVs from 
performing their RCS overpressurization protection function. 
Additionally, the proposed change does not affect the ability of any 
other safety-related equipment to perform its safety function.
    The only hardware changes are associated with making the 
automatic PORV actuation circuitry fully Class 1E. The RTS and 
Engineered Safety Feature Actuation System (ESFAS) protection 
systems will continue to function in a manner consistent with the 
plant design basis. The automatic PORV actuation circuitry 
modification will be performed in such a manner that all design, 
material, and construction standards that were applicable to safety-
related systems prior to the change are maintained. While the 
possibility that the PORVs fail to control RCS pressure, that at 
least one PORV fails to open, and that the operator fails to open 
the block valve and assure the PORV(s) are available for automatic 
pressure relief within the required time frame are all malfunctions 
of a different type than currently analyzed in the FSAR, they do not 
create different accident types. The Class 1E upgrade and changes to 
Emergency Operating Procedure E-0 will provide assurance that the 
reanalysis presented in Attachment 5 [to the application] will bound 
the results of this event which, in turn, is also bounded by the 
results presented in FSAR Section 15.6.1 for an inadvertent PSV 
opening.
    There are no other changes in the method by which any safety-
related plant system performs its safety function. The change will 
not affect the normal method of plant operation.
    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 PSVs, in conjunction with the RTS, provide overpressure 
protection for the RCS. The change in the upper limit of the PSV 
tolerance from +1% to +2%, with a reduction in the nominal setpoint 
from 2485 psig to 2460 psig, does not challenge the upper limit of 
overpressure protection. The maximum opening pressure setpoint is 
unchanged (other than a conservative round-off), and therefore, does 
not impact analyses performed for overpressure transients. The 
change to the PSV setpoint and setpoint tolerance does not change 
the conclusions of the existing thermal-hydraulic and stress 
analyses for the pressurizer safety and relief system. For all non-
LOCA events, the above evaluations support the change in the PSV 
setpoint and setpoint tolerance from 2485 psig +1% to 2460 psig +2%. 
The change in the PSV setpoint and setpoint tolerance also has no 
effect on the RTS and ESFAS trip setpoints.
    The Bases for Technical Specification 3.4.10 states the 
following in the Background section:
    ``The safety valves are designed to prevent the system pressure 
from exceeding the system Safety Limit (SL), 2735 psig, which is 
110% of the design pressure * * * The relief

[[Page 39964]]

capacity for each valve, 420,000 lb/hr at 2485 psig plus 3% 
accumulation, is based on postulated overpressure transient 
conditions resulting from a complete loss of steam flow to the 
turbine. This event results in the maximum surge rate into the 
pressurizer * * * .''
    The locked RCP [reactor coolant pump] rotor and loss of external 
electrical load/turbine trip transient analyses assume PSV actuation 
at 2550 psia. This value is conservatively based on a nominal PSV 
setpoint of 2500 psia plus a 1% setpoint tolerance and a 1% setpoint 
shift (due to the presence of the water seal). The maximum allowable 
PSV setpoint of 2509 psig is unaffected by the proposed change, 
other than a conservative round-off discussed previously. At a 
pressure of 2509 psig, the minimum relief capacity of the safety 
valves would be in excess of 420,000 lb/hr. However, the safety 
analyses for overpressurization events conservatively assume a 
420,000 lb/hr minimum design relief capacity for the PSVs.
    The proposed change does not affect the acceptance criteria for 
any other analyzed event nor is there a change to any other Safety 
Analysis Limit (SAL). The acceptance criteria for the Inadvertent 
ECCS Actuation at Power event will remain the same as currently 
analyzed; however, operator action and automatic PORV actuation will 
be relief upon to demonstrate compliance with that event's 
acceptance criteria.
    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 limits, FQ, FH, LOCA PCT [peak 
cladding temperature], peak local power density, or any other margin 
of safety. The radiological dose consequence acceptance criteria 
listed in the [NRC] Standard Review Plan continue to be met.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

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

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: June 1, 2000.
    Description of amendment request: The proposed amendment would 
permit changes to the Perry Nuclear Power Plant Updated Safety Analysis 
Report (USAR) to incorporate descriptions (in the form of text, tables, 
and drawings) of modifications to the Emergency Service Water (ESW) 
alternate intake sluice gate. The modifications will include (1) 
installation of a safety-related Class 1E selector switch that will be 
used to disable the automatic opening function of the sluice gate 
during warm weather and (2) installation of a non-safety inflatable 
sealing device on the gates between the ESW forebay and the alternate 
intake tunnel. The modifications are designed to increase overall 
reliability of the ESW system and to eliminate undesired operation of 
the ESW pumps.
    Date of publication of individual notice in Federal Register: June 
14, 2000 (65 FR 37414).
    Expiration date of individual notice: July 14, 2000.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: April 24, 2000.
    Brief description of amendment: The amendment allowed a one-time 
extension of some Technical Specification surveillance intervals due to 
elimination of a planned midcycle outage. The surveillances would be 
extended to no later than November 30, 2000.
    Date of issuance: June 12, 2000.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 129.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 2000 (65 FR 
26642).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 12, 2000.
    No significant hazards consideration comments received: No.

[[Page 39965]]

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of application for amendments: March 15, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications to permit plant operation with an ultimate 
heat sink temperature of 100  deg.F.
    Date of issuance: June 13, 2000.
    Effective date: Immediately as of the date of issuance and shall be 
implemented within 30 days.
    Amendment Nos.: 107 and 107.
    Facility Operating License Nos. NPF-72 and NPF-77: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR 
25763).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 13, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: July 27, 1999, as supplemented 
by letters dated October 7, 1999, and May 31, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications by adding a surveillance requirement to verify 
the Keowee out-of-tolerance logic trips and blocks closure of the 
appropriate overhead or underground power path breakers.
    Date of Issuance: June 6, 2000.
    Effective date: As of the date of issuance and shall be implemented 
by November 30, 2000.
    Amendment Nos.: 312, 312 and 312.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the TS.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46429). The supplements dated October 7, 1999, and May 31, 2000, 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 6, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 16, 1999.
    Brief description of amendment: The amendment revised the 
Permanently Defueled Technical Specifications by removing Figure 4.1-1, 
``Site and Exclusion Area Boundaries,'' and incorporating the 
applicable portions of this figure in the Trojan Defueled Safety 
Analysis Report. Other associated administrative changes resulting from 
the deletion of Figure 4.1-1, as well as an administrative change to 
the table of contents, were also made.
    Date of issuance: May 31, 2000.
    Effective date: May 31, 2000.
    Amendment No.: 204.
    Facility Operating License No. NPF-1: The amendment changes the 
Permanently Defueled Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4289).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 31, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 9, 2000.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Limiting Condition for Operation 3.8.2.1 to add two 
new Action Statements for operating conditions where a Class 1E 
battery's electrolyte temperature is below the minimum limit specified 
in TS Surveillance Requirement 4.8.2.1.b.3.
    Date of issuance: June 9, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 127.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12294).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 9, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 28, 1999, as 
supplemented March 17, 2000.
    Brief description of amendment: Revised Technical Specifications 
definitions for Engineered Safety Feature Response Time and Reactor 
Trip System Response Time, to provide for verification of response time 
for selected components, provided that the components and the 
methodology for verification have been previously reviewed and approved 
by the NRC.
    Date of issuance: June 13, 2000.
    Effective date: June 13, 2000.
    Amendment No.: 24.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56534). The March 17, 2000, submittal provided clarifying information 
that did not change the scope of the original request or change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 13, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 8, 1999, as supplemented by 
letters dated April 13, and May 30, 2000.
    Brief description of amendments: The amendments change Technical 
Specification 5.5.11, ``Ventilation Filter Testing Program (VFTP),'' to 
include the requirement for laboratory testing of Engineered Safety 
Feature (ESF) Ventilation System charcoal samples per American Society 
for Testing and Materials D3803-1989 and the application of a safety 
factor of 2.0 to the charcoal filter efficiency assumed in the plant 
design-basis dose analyses. The license amendments also extend the 
implementation date for License Amendment 74, currently June 30, 2000, 
to December 31, 2000.
    Date of issuance: June 12, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 78 and 78.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73101). The April 13, and May 30, 2000, letters provided clarifying 
information that did not change the scope of the November 8, 1999, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 12, 2000.


[[Page 39966]]


    No significant hazards consideration comments received: No.

Viacom Inc., Docket No. 50-22, Test Reactor, Waltz Mill, Pennsylvania

    Date of application for amendment: February 14, 2000 supplemented 
on March 8 and 25, 2000.
    Brief description of amendment: This amendment changes the license 
to reflect the transfer of the licensee for the Test Reactor at Waltz 
Mill from the CBS Corporation to Viacom Inc.
    Date of issuance: May 31, 2000.
    Effective Date: May 4, 2000.
    Amendment No.: 12.
    Facility License No. TR-2: This amendment changes the license.
    Date of Initial notice in Federal Register: February 29, 2000 (65 
FR 10841).
    The Commission has issued a Safety Evaluation for this amendment 
dated April 13, 2000.
    No significant hazards consideration comments received: No.
    Local Public Document: N/A.

    Dated at Rockville, Maryland, this 21st day of June 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-16193 Filed 6-27-00; 8:45 am]
BILLING CODE 7590-01-P