[Federal Register Volume 62, Number 214 (Wednesday, November 5, 1997)]
[Notices]
[Pages 59912-59927]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-29138]


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NUCLEAR REGULATORY COMMISSION


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

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 October 10, 1997, through October 24, 1997. 
The last biweekly notice was published on October 22, 1997 (62 FR 
54866).

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 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, MD 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 December 5, 1997, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for

[[Page 59913]]

leave to intervene or who has been admitted as a party may amend the 
petition without requesting leave of the Board up to 15 days prior to 
the first prehearing conference scheduled in the proceeding, but such 
an amended petition must satisfy the specificity requirements described 
above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    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 No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, MD

    Date of amendment request: October 2, 1997.
    Description of amendment request: The amendment request would 
change the Technical Specifications to identify a proposed upgrade of 
the electrical capacity of the No. 1B emergency diesel generator.
    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 Engineered Safety Features (ESF) electrical system provides a 
reliable source of electrical power to the 4.16 kV ESF busses to 
operate the necessary accident mitigation equipment, should offsite 
power be lost. The proposed change to the Technical Specifications was 
prompted by the upgrade of the electrical and mechanical capacity of 
the No. 1B Fairbanks Morse Emergency Diesel Generator (EDG). The 
increased electrical capacity of the No. 1B Fairbanks Morse EDG will 
give the operators greater flexibility in the choice of discretionary 
loads for the mitigation of accidents. This modification necessitates 
changes to the Technical Specifications.
    The ESF electrical system, including the four EDGs, is used to 
mitigate the consequences of an accident. The modification to upgrade 
the capacity of No. 1B EDG will increase the electrical output of the 
EDG, but will not change the configuration of the ESF electrical system 
or any support systems such that the EDGs would become an accident 
initiator. Therefore, the proposed change would not increase the 
probability of an accident previously evaluated.
    The proposed Technical Specifications will continue to demonstrate 
the reliability and capability of the upgraded No. 1B EDG to perform 
its accident mitigation function. The proposed changes to the 
surveillance requirements do not alter the intent or performance of the 
surveillance. Only the electrical loadings changed, reflecting the 
change in the EDG's electrical capacity. Implementation of the proposed 
Technical Specifications will not reduce the ability of No. 1B EDG to 
perform its safety functions. Any auxiliary systems that required 
modification or analysis to support the upgraded ratings of the 1B 
Fairbanks Morse EDG have been determined not to adversely impact 
operation of any other plant systems necessary to mitigate the 
consequences of an accident. Therefore, the proposed change would not 
increase the consequences of an accident previously evaluated.
    Therefore, the proposed change 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.
    The proposed change increases the electrical loading for 
surveillance requirements to reflect the upgrade to the electrical 
capacity of the No. 1B Fairbanks Morse EDG. This change does not add 
any new equipment, modify any interfaces with any existing equipment, 
change the equipment's function, or the method of operating the 
equipment to be modified. The system will continue to operate in the 
same manner as before the capacity upgrades were implemented. The 
modified No. 1B EDG will continue to function as an accident

[[Page 59914]]

mitigator, and will not become an initiator of any accident.
    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 function of the EDG is to provide a reliable source of 
electrical power to the ESF electrical system sufficient to power the 
necessary accident mitigation equipment, should offsite power be lost. 
This safety function is demonstrated by performing the required 
surveillance tests. The proposed changes do not alter the intent or 
method of performance of any of the surveillance tests.
    The proposed change to the Technical Specifications was prompted by 
the upgrade of the electrical and mechanical capacity of the No. 1B 
Fairbanks Morse EDG. The higher electrical capacity will result in an 
increase in the margin between No. 1B EDG's electrical capacities and 
the electrical power required to operate safety-related equipment 
required for safe shutdown or accident mitigation. The increased 
electrical capacity results in the need to increase the electrical 
loadings used in the surveillance tests. The changes in the 
surveillance tests will continue to ensure that the EDG is tested 
appropriately and will continue to perform its safety function. In 
addition, it should be noted that upgrades on identical Fairbanks Morse 
EDGs have already been performed on Unit 2 and have resulted in 
identical changes to the Unit 2 Technical Specifications. Because of 
the increased electrical margin afforded by the upgraded EDG, these 
modifications may be considered an increase in the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Calvert County Library, Prince 
Frederick, MD 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.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, IL; Docket Nos. STN 50-
456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, 
IL

    Date of amendment request: September 8, 1997.
    Description of amendment request: The proposed amendment would 
revise Byron and Braidwood Technical Specification (TS) 4.5.2.b and 
associated bases as they relate to the requirement to vent the 
Emergency Core Cooling System (ECCS) pump casings and discharge piping 
high points outside containment. The change will revise the Unit 1 
requirement for ultrasonic examinations every 31 days to also include 
ultrasonic examination of the piping at the 1CV206 valve for Byron 
(1CV207 valve for Braidwood) if the 1B Chemical and Volume Control (CV) 
pump is idle. These changes are required to align the surveillance 
requirements for Unit 1 with those of Unit 2. In addition, the 
condition that the Unit 1 requirements will be applicable only until 
the end of the current cycle is deleted consistent with the Unit 2 
requirements. With these changes there will no longer be the need to 
maintain separate pages for Unit 1 and Unit 2 requirements.
    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 will align the surveillance requirements for 
both Units 1 and 2 with the installed system design and normal 
operating conditions. No increase in the probability of an accident 
will occur as a result of this change. The conduct of surveillances 
required by the Technical Specifications is not postulated to initiate 
an accident. The level of surveillance performed to date has provided 
confidence that the objective of the current surveillance requirement 
has been met. As such, the proposed change does not result in a 
significant increase in the probability of occurrence of a previously 
analyzed accident.
    The consequences of a previously analyzed accident are not 
increased. Operating experience has shown that the level of 
surveillance performed to date is sufficient to provide confidence that 
no significant voiding has occurred in the affected piping. Ultrasonic 
examinations have confirmed the water solid condition of the piping. 
Although voiding is not expected, evaluation of postulated voided 
conditions confirm that unacceptable dynamic loading would not occur, 
and, therefore, the integrity of the ECCS piping is not compromised. 
Thus, the ECCS will be capable of performing its design function of 
cooling the reactor core and providing shutdown capability following 
initiation of the certain accidents. This will ensure that the 
consequences of a previously analyzed accident are not significantly 
increased.
    Therefore, these proposed revisions do not result in a significant 
increase in the probability or consequences of an accident previously 
analyzed.
    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 create the possibility of a new or 
different kind of accident. ComEd has evaluated the piping 
configuration for the ECCS discharge piping of the ECCS subsystems. A 
specific engineering evaluation of both a voided 2-inch and 8-inch RH 
[Residual Heat Removal] line was performed. This evaluation concluded 
that the piping can withstand the dynamic loads caused by the maximum 
credible air void. Due to the higher-pressure rating and smaller size 
of the SI [Safety Injection] and CV discharge piping, this evaluation 
is considered bounding for the ECCS subsystems. The results of the 
evaluation were submitted for staff review in a letter dated March 12, 
1990, in support of Amendments 47 and 36 to the Operating Licenses for 
Byron and Braidwood, respectively. The proposed changes will not result 
in new failure modes because no new equipment is installed, and 
installed equipment is not operated in a new or different manner. 
Manual venting operations have been performed as permitted by system 
operation and piping configuration. This venting surveillance does not 
apply to subsystems in communication with operating systems because the 
flows and/or pressures prevalent in these systems are sufficient to 
provide confidence that water hammer which could occur from voiding 
would not result in unacceptable dynamic loads from water hammer will 
not occur. Accordingly, this change will not create the possibility of 
a new or different kind of accident.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.

[[Page 59915]]

    The margin of safety is not significantly reduced because the 
proposed change will provide sufficient assurance that excessive 
voiding will not occur. This will assure proper system functioning. 
Venting of the idle subsystems, in conjunction with the operating 
conditions of the subsystems in operation, provides confidence that 
voiding is not present. This has been confirmed by the performance of 
ultrasonic examinations of the piping of interest. This meets the 
objective of the surveillance requirement and thus preserves 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, IL 61010; for 
Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, IL 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, IL 60603.
    NRC Project Director: Robert A. Capra.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, MI

    Date of amendment request: January 18, 1996, as revised October 1, 
1997.
    Description of amendment request: The original proposed amendment 
(January 18, 1996) would have deleted the requirement in Section 6.5.6 
of the Technical Specifications (TS) to perform inservice inspections 
of the primary coolant pump (PCP) flywheels. The October 1, 1997, 
submittal would revise Section 6.5.6 of the TS to lengthen the flywheel 
inspection period to 10 years rather than delete it entirely. The note 
added by Amendment 175 for the deletion of the inspection at the end of 
Cycle 12 would also be deleted. The original submittal was previously 
noticed in the Federal Register on September 11, 1996 (61 FR 47976).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration in 
its original submittal. In its revised submittal the licensee stated 
that the conclusions reached in the original no significant hazards 
consideration determination were still valid because the revised 
submittal just reduces the frequency of the test as opposed to deleting 
it. The original no significant hazards consideration discussion is 
presented below:
    The following evaluation supports the finding that operation of the 
facility in accordance with the proposed change to the Technical 
Specifications would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to the Technical Specifications would delete 
the requirement to perform non-destructive examination of the upper 
flywheel on the PCPs. The fracture mechanics analyses conducted to 
support the change show that a preexisting crack sized just below 
detection level will not grow to the flaw size necessary to result in 
flywheel failure within the life of the plant. This analysis 
conservatively assumes minimum material properties, maximum flywheel 
accident speed, location of the flaw in the highest stress area and a 
number of startup/shutdown cycles eight times greater than expected. 
Since an existing flaw in the flywheel will not grow to the allowable 
flaw size under normal operating conditions or to the critical flaw 
size under LOCA [loss-of-coolant accident] conditions over the life of 
the plant, elimination of inservice inspection for such cracks during 
the plant's life will not involve a significant increase in the 
probability of an accident previously considered.
    The proposed changes do not increase the amount of radioactive 
material available for release or modify any systems used for 
mitigation of such releases during accident conditions. Therefore, 
operation of the facility in accordance with the proposed change to the 
Technical Specifications would 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 previously evaluated.
    The proposed change to the Technical Specifications would not 
change the design, configuration, or method of operation of the plant 
and therefore, operation of the facility in accordance with the 
proposed change to the Technical Specifications would not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the Technical Specifications would not 
result in a significant reduction in the margin of safety. Significant 
conservatisms have been used for calculating the allowable flaw size, 
critical flaw size and crack growth rate in the PCP flywheels. These 
include minimum material properties, maximum flywheel accident speed, 
location of the postulated flaw in highest stress area and a number of 
startup/shutdown cycles eight times greater than expected. Since an 
existing flaw in the flywheel will not grow to the maximum allowable 
flaw size under normal operating conditions or to the critical flaw 
size under LOCA conditions over the life of the plant, elimination of 
inservice inspections for such cracks during the plant's life will 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. In addition, the staff agrees that this analysis bounds the 
conditions in the revised submittal. The editorial change to delete an 
obsolete note has no effect on plant operation or safety and also 
satisfies the three standards of 10 CFR 50.92(c). Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, MI 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, MI 49201.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: October 10, 1997.
    Description of amendment request: The proposed change (TSCR 253) 
would reflect the registered trade name of ``GPU Nuclear'' in the 
operating license for the Oyster Creek Nuclear Generating Station 
(OCNGS) and change the legal name of the operator of OCNGS from GPU 
Nuclear Corporation to GPU Nuclear, Inc. In addition, two minor 
editorial corrections are included.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the facility in accordance with the proposed amendment 
would not involve a significant increase in the probability of 
occurrence or the consequences of an accident previously evaluated. The 
proposed amendment adds to the license and the technical specifications 
the trade name of the

[[Page 59916]]

Owner of Oyster Creek. The change in the legal name of the operator of 
Oyster Creek is an administrative change made to reflect the name 
changes made throughout the GPU family of companies. The name change 
has no impact on plant design or operation.
    Operation of the facility in accordance with the proposed amendment 
would not create the possibility of a new or different kind of accident 
from any accident previously evaluated because no new failure modes are 
created by the proposed changes. The use of a trade name for the Owner 
of Oyster Creek and the change in the legal name of the operator of 
Oyster Creek has no impact on plant design or operation. Thus, there is 
no creation of the possibility of a new or different kind of accident 
from those previously evaluated.
    Operation of the facility in accordance with the proposed amendment 
will not involve a significant reduction in a margin of safety. The 
proposed amendment does not change any operating limits for reactor 
operation.
    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. In addition, the staff has reviewed the licensee's proposed 
editorial changes and determined that they do not effect the 
conclusions of the analysis. 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: Ronald B. Eaton, Acting Director.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 1, Oswego County, NY

    Date of amendment request: October 21, 1997. This notice supersedes 
a previous notice, (62 FR 30625), published June 4, 1997, which was 
based upon the licensee's application for amendment dated May 16, 1997. 
The licensee's application dated October 21, 1997, supersedes the May 
16, 1997, submittal in its entirety.
    Description of amendment request: The proposed amendment would 
change the administrative section of the Technical Specifications (TS) 
regarding the Operations organization. Specifically, TS 6.2.2i 
currently states that ``The Manager Operations, Station Shift 
Supervisor Nuclear and Assistant Station Shift Supervisor Nuclear shall 
hold senior reactor operator licenses.'' This would be changed to state 
``As a minimum, either the Manager Operations or the General Supervisor 
Operations shall hold a senior reactor operator license. The Station 
Shift Supervisor Nuclear and Assistant Station Shift Supervisor Nuclear 
shall hold senior reactor operator licenses.'' In addition TS 6.3.1 
would be revised to indicate an additional exception to the operating 
staff's qualification requirements set forth in American National 
Standard Institute (ANSI) N18.1-1971, ``Selection and Training of 
Nuclear Power Plant Personnel.'' Specifically, this change would 
require that the Manager Operation, in lieu of meeting the senior 
reactor operator (SRO) requirements of ANSI N18.1-1971, shall (1) hold 
an SRO license at the time of appointment, or (2) have held an SRO 
license at Nine Mile Point Nuclear Station Unit 1 or a similar unit, or 
(3) have been certified for equivalent SRO knowledge.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The operation of Nine Mile Point Unit 1 [NMP1], in accordance 
with the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The addition of the position of GSO and the requirement for either 
the GSO or the Manager Operations to have an SRO license is a 
restructuring of the Operations department. The proposed changes are 
administrative changes that provide additional Operations management 
oversight capabilities. Additional restrictions placed on the Manager 
Operations minimum qualification requirements for experience and SRO 
level knowledge for the resulting organization meet the intent of ANSI 
N18.1-1971 and SRP [Standard Review Plan, NUREG-0800] 13.1.1-13.1.3. No 
physical modification of the plant is involved and no changes to the 
methods in which plant systems are operated are required.
    None of the precursors of previously evaluated accidents are 
affected, and no new failure modes are introduced. Therefore, this 
change will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The addition of the position of GSO and the requirement for either 
the GSO or the Manager Operations to have an SRO license is a 
restructuring of the Operations department. The proposed changes are 
administrative changes that provide additional Operations management 
oversight capabilities. Additional restrictions placed on the Manager 
Operations minimum qualification requirements for experience and SRO 
level knowledge ensure the resulting organization meets the intent of 
ANSI N18.1-1971 and SRP 13.1.1-13.1.3. No physical modification of the 
plant is involved and no changes to the methods in which plant systems 
are operated are required. As such, the change does not introduce any 
new failure modes or conditions that may create a new or different 
accident. Therefore, this change does not itself create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The addition of the position of GSO and the requirement for either 
the GSO or the Manager Operations to have an SRO license is a 
restructuring of the Operations department. The proposed changes are 
administrative changes that provide additional Operations management 
oversight capabilities. Additional restrictions placed on the Manager 
Operations minimum qualification requirements for experience and SRO 
level knowledge ensure the resulting organization meets the intent of 
ANSI N18.1-1971 and SRP 13.1.1-13.1.3. No physical modification of the 
plant is involved and no changes to the methods in which plant systems 
are operated are required. As such, this change does not in itself 
adversely affect any physical barrier to the release of radiation to 
plant personnel or to the public. Therefore, the change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 59917]]

    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, NY 
13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: October 7, 1997.
    Description of amendment request: Technical Specifications 4.6.1.1, 
3/4.6.1.2, and 3/4.6.1.3 require the testing of the containment to 
verify leakage limits at a specified test pressure. The proposed 
amendment would (1) modify the list of valves that can be opened in 
Modes 1 through 4, (2) remove a footnote on Type A testing, and (3) 
make editorial changes to the Technical Specifications and associated 
Bases sections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 10 CFR 
50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this conclusion 
is that the three criteria of 10 CFR 50.92(c) are not satisfied. The 
proposed revision does not involve [an] SHC because the revision would 
not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    The proposed change to Technical Specification Surveillance 4.6.1.1 
deletes valves from the list of containment isolation valves that may 
be opened under administrative control. Deleting the valves, which 
means that they are not allowed to be opened under the Limiting 
Condition of Operation, [cannot] cause an accident. The valves being 
added in the steam lines to the steam-driven auxiliary feedwater pump 
can be used to heat the steam lines prior to testing the steam-driven 
auxiliary feed water pump. Heating the steam lines prior to testing the 
steam-driven auxiliary feedwater pump does not increase the likelihood 
of a steam line break.
    The administrative change of replacing the ``-'' with an ``*'' in 
the valve designation can neither cause [an] accident nor affect the 
consequences of any accident.
    The addition of the RHR [residual heat removal] system containment 
isolation valves reflects the fact that these valves can be opened 
during Mode 4 to allow plant heatup and cooldown. Plant heatup and 
cooldown, in accordance with normal plant operation and the Technical 
Specifications, does not increase the likelihood of the above 
accidents.
    The administrative controls include the appropriate considerations 
that containment integrity will be established, when required. By 
establishing containment integrity, the assumptions in the design basis 
analyses are assured. This means that for LOCA [loss-of-coolant 
accident], steam line break and feed line break accidents inside 
containment, there is no effect on their consequences.
    Valves in the steam lines to the steam-driven auxiliary feedwater 
pump are being added to the list of valves allowed to be opened under 
administrative control. This means that these could be open at the 
initiation of an accident. The administrative controls under which 
these valves are opened provides assurance that containment integrity 
will be established, when required. Similarly, for an SGTR [steam 
generator tube rupture], Locked Rotor or Control Rod Ejection event, 
the administrative controls provides assurance that these valves will 
be closed and, therefore, allowing them to be opened will not adversely 
impact the consequences of these events. If failure to close is 
postulated as a single failure for these events, the results would be 
bounded by the analyses described in the FSAR [final safety analysis 
report]. For example, the Locked Rotor accident assumes a stuck open 
steam generator power-operated pressure relief valve (SG PORV). The 
steam released by the assumed single failure of the SG PORV, for the 
twenty minutes until the valve is isolated, would exceed the expected 
releases as a result of failure to close valve 3MSS*V885, 3MSS*V886, or 
3MSS*V887, which are in \1/4\ inch lines. Therefore, allowing these 
valves to be opened under administrative control does not effect the 
consequences of the previously evaluated accidents.
    The FSAR, Section 15.1.5, provides the assumptions on steam 
releases for the consequences of the steam line break accident. The 
steam generator with the broken steam line is assumed to be open to the 
atmosphere for the duration of the event and, therefore, these valves 
being open would not impact that assumption. For the unaffected steam 
generators, steam is assumed released to the atmosphere to remove decay 
heat. These valves are in \1/4\ inch lines which means that any steam 
released via this path would only be a small fraction of decay heat and 
will not adversely affect control of decay heat removal. Therefore, 
whether these valves are open or not will not affect the consequences 
of a steam line break outside containment.
    Allowing the RHR system containment isolation valves to be open, 
under administrative control in Mode 4, does not change the way the 
system is operated. This proposed change to the footnote does not 
change the operators response to an accident in Mode 4. Therefore, the 
addition of these valves does not affect the consequences of the 
previously evaluated accidents.
    The proposed change to Technical Specification Surveillance 
4.6.1.2.a will delete footnote ``*'' which referred to an exemption 
granted by the NRC to permit the Type A test to be delayed until RFO6 
[refueling outage 6]. However, the current extended shutdown has 
significantly delayed RFO6 and NNECO intends to perform the Type A test 
during this midcycle shutdown. The deletion of the footnote does not 
alter the operation of any system or the containment or containment 
airlocks, as assumed for accident analyses.
    Additionally, Technical Specifications 4.6.1.1, 3/4.6.1.2, and 3/
4.6.1.3, and Bases Sections 3/4.6.1.1, 3/4.6.1.2, and 3/4.6.1.3 are 
reworded to provide clarity and consistency. These proposed changes do 
not alter the operation of any system or the containment or containment 
airlocks during accident analyses.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to Technical Specifications 4.6.1.1, 3/
4.6.1.2, and 3/4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2, and 3/
4.6.1.3 do not alter the operation of any system or the containment or 
containment airlocks, during normal operation or as assumed in accident 
analyses.
    Deleting containment isolation valves from the list of those that 
are allowed to be opened under administrative control can not modify 
plant response to an accident. Adding administrative control when the 
RHR system containment isolation valves are opened in Mode 4 for normal 
plant cooldown and heatup can not create a new or different accident. 
Allowing valves to be opened

[[Page 59918]]

to heat the steam lines to the steam-driven auxiliary feedwater pump 
prior to testing does not create the possibility of a new or different 
accident. The administrative change to the valve designation can not 
modify plant response.
    Therefore, the proposed revision does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to Technical Specifications 4.6.1.1, 3/
4.6.1.2, and 3/4.6.1.3, and Bases Sections 3/4.6.1.1, 3/4.6.1.2, and 3/
4.6.1.3 do not alter the design, maintenance or function of any system 
or the containment or the containment airlocks. Additionally, the 
proposed changes do not alter the testing of any system or the 
containment or containment airlocks, or alter any assumption used in 
the accident analyses.
    The considerations associated with administrative control are being 
added to the bases of the technical specification. These considerations 
are identical to those provided in GL 91-08 [Generic Letter 91-08]. 
This means that the changes will maintain the margin of safety. The 
valves that are allowed to be open in the steam lines to the steam-
driven auxiliary feedwater [pump] do not impact the accident analyses 
and therefore do not reduce the margin of safety. The addition of the 
RHR system containment isolation valves reflects the fact that these 
valves are opened for heatup and cooldown in Mode 4. The change adds 
the requirements of administrative controls to these RHR system valves 
in Mode 4, but does not modify the use of these valves. The 
administrative change to the valve designation can not affect the 
margin of safety.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is determined 
that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT, and the Waterford Library, ATTN: Vince Juliano, 49 Rope 
Ferry Road, Waterford, CT.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: October 15, 1997.
    Description of amendment request: Technical Specification 
Surveillances 4.1.2.3.1, 4.1.2.4.1, 4.5.2, 4.6.2.1, and 4.6.2.2 require 
the recirculation spray, quench spray, residual heat removal, 
centrifugal charging, and safety injection pumps to be tested on a 
periodic basis and after modifications that alter subsystem flow 
characteristics. The proposed changes to these surveillances would 
include replacing the specific surveillance pump pressure with a 
statement that the test be conducted in accordance with Specification 
4.0.5, Inservice Testing Program. The proposed changes would also 
include a decrease in the required individual safety injection and 
centrifugal charging pump injection line flow rates, an increase in the 
allowed individual safety injection pump runout flow rate, and 
editorial changes to the surveillances.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 10 CFR 
50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this conclusion 
is that the three criteria of 10 CFR 50.92(c) are not satisfied. The 
proposed revision does not involve an SHC because the revision would 
not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    The Technical Specification changes transfer control of the pump 
developed head requirements for the Centrifugal Charging, Safety 
Injection, Quench Spray, Residual Heat Removal, and Recirculation Spray 
pumps from the Technical Specifications to the Inservice Test program. 
The acceptance criteria will still assure that the safety analysis 
assumptions are valid. The Technical Specification changes reduce the 
minimum flow requirements for the Charging and Safety Injection pumps 
and increase the maximum allowed flow for the Safety Injection pumps. 
Modifying the surveillance requirements [cannot] cause an accident and, 
therefore, [cannot] increase the probability of an accident. The 
revised minimum required flows are consistent with the flows used in 
the accident analyses and, therefore, the change [cannot] increase the 
consequences of any accident. The safety injection pumps are disabled 
such that they [cannot] be a source of mass addition to the RCS 
[reactor coolant system] whenever the cold overpressure system is 
required to be operable. Therefore, the increase in the allowed maximum 
safety injection pump flow has no effect on the cold overpressure 
accident analysis.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes transfer control of the pump developed head 
requirements from the Technical Specifications to the Inservice Test 
program and modify the required flow surveillance values. The 
surveillance values that are used in the Inservice Test program and the 
Technical Specification are consistent with the accident analysis. The 
increase in the allowed maximum safety injection pump flow does not 
impact the cold overpressure accident analysis. The changes do not 
involve any changes to the way that the pumps are operated. The pumps 
will be used post-accident the same way as they are used prior to the 
change.
    Therefore, the proposed revision does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The control of the pump developed head acceptance criteria is being 
transferred from the Technical Specification to the Inservice Test 
program. The acceptance criteria, at a minimum, will assure that the 
design basis analyses are valid. The minimum pump flow surveillance 
requirements in Specification 4.5.2.h are consistent with the 
assumptions of the accident analysis. The maximum allowed Safety 
Injection flow does not exceed the vendor recommendation for maximum 
continuous runout flow. The NPSH [net positive suction head] available 
to the pumps during both the injection and recirculation phases post-
accident

[[Page 59919]]

exceeds the NPSH required at the higher allowed flow. Also, the safety 
injection pumps are disabled so that they [cannot] be an injection 
source when the cold overpressure system is required to be operable 
which means that the increase in maximum flow does not affect the cold 
overpressure accident analysis. Restricting orifices are being 
installed in the injection lines from the safety injection and charging 
pumps to the Reactor Coolant System as required. The restricting 
orifices and the changes to the required flows will allow for resetting 
the throttle position of the existing throttle valves. The sizing of 
the restricting orifices and the associated re-throttling of the 
throttle valves will be in accordance with Regulatory Guide 1.82. The 
proposed changes allow for the setting of the throttle valve positions 
so that the openings will be larger than the sump screen mesh opening 
size while assuring that the design basis flow values are valid.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is determined 
that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT, and the Waterford Library, ATTN: Vince Juliano, 49 Rope 
Ferry Road, Waterford, CT.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Deputy Director: Phillip F. McKee.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, NE

    Date of amendment request: July 25, 1997.
    Description of amendment request: The proposed amendment request 
would revise the Technical Specifications (TS) to implement 10 CFR Part 
50 Appendix J, Option B by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leakage-Test Program,'' with certain 
exceptions detailed in the licensee's application. This revision 
supersedes the staff's description of amendment request that was 
published on October 8, 1997 (62 FR 52586).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change implements Option B of 10 CFR Part 50 Appendix 
J on performance-based containment leakage testing. The proposed change 
does not involve a change to the plant design or operation. As a 
result, the proposed change does not affect any parameters or 
conditions that contribute to the initiation of any accidents 
previously evaluated. The proposed change potentially affects the leak-
tight integrity of the containment structure designed to mitigate the 
consequences of a Loss-of-Coolant Accident (LOCA). The function of the 
containment is to maintain functional integrity during and following 
the peak transient pressures and temperatures and limit fission product 
leakage following the design basis LOCA. Because the proposed change 
does not alter the plant design, only the frequency of measuring Type 
A, B, and C leakage, the proposed change does not directly result in an 
increase in containment leakage.
    Test intervals will be established based on the performance history 
of components being tested. The frequency of monitoring the relatively 
few containment isolation valves and/or containment penetrations 
subject to above normal leakage will not decrease by implementing 
Option B of Appendix J. A performance based program will identify those 
valves and penetrations which must continue to be tested each refueling 
outage.
    The risk resulting from the proposed changes is characterized as 
follows, based primarily on the results contained in NUREG-1493 
``Performance-Based Containment Leakage Test Program,'' the principal 
Technical Support Document used by the NRC as the basis for the 
Appendix J Final Rule:
Type A Testing
    NUREG-1493 found that the effect of containment leakage on overall 
accident risk is minimal since risk is dominated by accident sequences 
that result in failure or bypass of the containment. Industry wide, 
Integrated Leak Rate Tests (ILRTs) have only found a small fraction of 
the leaks that exceed current acceptance criteria. Only three percent 
of all leaks are detectable only by ILRTs, and therefore, by extending 
the Type A testing intervals, only three percent of all leaks have a 
potential for remaining undetected for longer periods of time. In 
addition, when leakage has been detected by ILRTs, the leakage rate has 
been only marginally above existing requirements. The Fort Calhoun 
Station Unit No. 1 Type A testing confirms the industry-wide experience 
that a majority of the leakage experienced during Type A testing is 
through components tested by Type B and C tests.
    NUREG-1493 found that these observations, together with the 
insensitivity of reactor accident risk to the containment leakage rate, 
show that increasing the Type A leakage test intervals would have a 
minimal impact on public risk.
Type B and C Testing
    NUREG-1493 found that while Type B and C tests can identify the 
vast majority (greater than 95 percent) of all potential leakage paths, 
performance-based alternatives to current local leakage-testing 
requirements are feasible without significant risk impacts. The risk 
model used in NUREG-1493 suggests that the number of components tested 
would be reduced by about 60 percent with less than a three-fold 
increase in the incremental risk due to containment leakage. Since, 
under existing requirements, leakage contributes less than 0.1 percent 
of overall accident risk, the overall impact is very small. In 
addition, the NRC's Final Regulatory Impact Analysis concluded that 
while the extended testing intervals for Type B and C tests led to 
minor increases in potential offsite dose consequences, the beneficial 
expected decrease in onsite worker dose received during ILRT and local 
leak rate testing exceeds (by at least an order of magnitude) the 
potential off-site dose consequences.
    Therefore, the proposed change will not result in a significant 
increase in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    There will be no physical alterations to the plant configuration, 
changes to setpoint values, or changes to the implementation of 
setpoints or limits as a result of this proposed change. As a result, 
the proposed change does not affect any of the parameters or conditions 
that could contribute to initiation of any accidents.

[[Page 59920]]

    This change involves the reduction of Type A, B, and C test 
frequency. Except for the method of defining the test frequency, the 
methods for performing the actual tests are not changed. No new 
accident modes are created by extending the testing intervals. No 
safety-related equipment or safety functions are altered as a result of 
this change. Extending the test frequency has no influence on, nor does 
it contribute to, the possibility of a new or different kind of 
accident or malfunction from those previously analyzed. 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 only affects the frequency of Type A, B, and C 
testing. Except for the method of defining the test frequency, the 
methods for performing the actual tests are not changed.
    The frequency of monitoring the relatively few containment 
isolation valves and/or containment penetrations subject to above 
normal leakage will not decrease by implementing Option B of Appendix 
J. A performance based program will identify those valves and 
penetrations which must continue to be tested each refueling outage. 
NUREG-1493 has determined that, under several different accident 
scenarios, the increased risk of radioactivity release from containment 
is negligible with the implementation of these proposed changes.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to containment leakage 
rate. The containment isolation system is designed to limit leakage to 
La, which is stated in the Fort Calhoun Station Unit No. 1 Technical 
Specifications to be 0.1 percent by weight of the containment air per 
24 hours at 60 psig.
    The limitation on containment leakage rate is designed to ensure 
that total leakage volume will not exceed the value assumed in the 
accident analyses at the peak accident pressure. The margin to safety 
for the offsite dose consequences of postulated accidents directly 
related to the containment leakage rate is maintained by meeting the 
1.0 La acceptance criteria. The La value is not being modified by this 
proposed change.
    Except for the method of defining the test frequency, no change in 
the method of testing is being proposed. The Type B and C tests will 
continue to be done at 60 psig or greater. Other programs are in place 
to ensure that proper maintenance and repairs are performed during the 
service life of the primary containment and systems and components 
penetrating the primary containment.
    Therefore, the proposed change will not result in a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, NE 68102.
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: September 3, 1997.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to revise the number of hours 
operating personnel can work in a normal shift. The proposed amendment 
also contains some administrative changes to the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
evaluated?
    A. Establishing operating personnel work hours at, ``an 8 to 12 
hour day, nominal 40 hour week,'' allows normal plant operations to be 
managed more effectively and does not adversely effect performance of 
operating personnel. Overtime remains controlled by site administrative 
procedures in accordance with NRC Policy Statement on working hours 
(Generic Letter 82-12). If 8 hour shifts are maintained in part or 
whole, then acceptable levels of performance from operating personnel 
is assured through effective control of shift turnovers and plant 
activities. No physical plant modifications are involved and none of 
the precursors of previously evaluated accidents are affected. 
Therefore, this change will not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    B. Editorial changes clarify section 6.2.2.g without changing the 
intent or meaning. The proposed change meets the intent of the NRC 
Policy Statement on working hours (Generic Letter 82-12).
    C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the 
intent or meaning of the technical specification sections. 
Clarification to the table notation in section 4.1 related to the 
definition of shift checks to monitor plant conditions will continue as 
intended but are allowed to increase up to at least once per 12 hours. 
This increase is consistent with standard industry practice as 
represented by the Standard Technical Specifications (STS), Reference 
1.
    2. Does the proposed license amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    A. Establishing operating personnel work hours at, ``an 8 to 12 
hour day, nominal 40 hour week,'' allows normal plant operations to be 
managed more effectively and does not adversely effect performance of 
operating personnel. If 8 hour shifts are maintained in part or whole, 
then acceptable levels of performance from operating personnel is 
assured through effective control of shift turnovers and plant 
activities. Overtime remains controlled by site administrative 
procedures in accordance with the NRC Policy Statement on working hours 
(Generic Letter 82-12). No physical modification of the plant is 
involved. As such, the change does not introduce any new failure modes 
or conditions that may create a new or different accident. Therefore, 
operation in accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    B. Editorial changes clarify section 6.2.2.g without changing the 
intent or meaning. The proposed change meets the intent of the NRC 
Policy Statement on working hours (Generic Letter 82-12).
    C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the 
intent or meaning of the technical specification sections. 
Clarification to the table notation in section 4.1 related to the 
definition of shift checks to monitor plant conditions will continue as 
intended but are allowed to increase up

[[Page 59921]]

to at least once per 12 hours. This increase is consistent with 
standard industry practice as represented by the Standard Technical 
Specifications (STS), Reference 1.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    A. Establishing operating personnel work hours at, ``an 8 to 12 
hour day, nominal 40 hour week,'' allows normal plant operations to be 
managed more effectively and does not adversely effect performance of 
operating personnel. If 8 hour shifts are maintained in part or whole, 
then acceptable levels of performance from operating personnel is 
assured through effective control of shift turnovers and plant 
activities. Overtime remains controlled by site administrative 
procedures in accordance with the NRC Policy Statement on working hours 
(Generic Letter 82-12) and is consistent with the Standard Technical 
Specifications. The proposed change involves no physical modification 
of the plant, or alterations to any accident or transient analysis. 
There is no Basis to section 6 of the Technical Specifications, and the 
changes are administrative in nature. Therefore, the change does not 
involve any significant reduction in a margin of safety.
    B. Editorial changes clarify section 6.2.2.g without changing the 
intent or meaning. The proposed change meets the intent of the NRC 
Policy Statement on working hours (Generic Letter 82-12).
    C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the 
intent or meaning of the technical specification sections. 
Clarification to the table notation in section 4.1 related to the 
definition of shift checks to monitor plant conditions will continue as 
intended but are allowed to increase up to at least once per 12 hours. 
This increase is consistent with standard industry practice as 
represented by the Standard Technical Specifications (STS), Reference 
1.
    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, NY 10601.
    Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New 
York, NY 10019.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: September 29, 1997, as supplemented 
October 8, 1997. The September 29 application and October 8, 1997, 
supplement supersede the September 13, 1996, application and its April 
24, 1997, supplement. This notice supersedes the notice published on 
October 9, 1996 (61 FR 197) in its entirety.
    Description of amendment request: The proposed amendment would 
change the Ginna Station Technical Specifications (TSs) which would 
allow referencing of revision of the Ginna Station pressure and 
temperature limits report (PTLR) for the reactor coolant system (RCS) 
pressure and temperature (P/T) limits and low temperature overpressure 
protection (LTOP) limits. The proposed amendment would correct some 
typographical errors in the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes revise Administrative Controls Section 5.6.6.c 
to update the reference to the NRC's approval of the first use of the 
PTLR methodology, update the RCS P/T methodology to the final NRC 
approved version, allow use of ASME Code Case N-514 for LTOP enable 
temperature methodology, and to correct a typographical error. These 
changes complete implementation of Generic Letter 96-03 by referencing 
NRC approved methodology within the Administrative Controls. The 
updated RCS P/T methodology has been generically approved by the NRC 
while the use of ASME Code Case N-514 for LTOP enable temperature 
methodology was previously approved for use at Ginna Station by the 
NRC. As such, these changes are administrative in nature and do not 
impact initiators or analyzed events or assumed mitigation of accident 
or transient events. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an accident 
previously analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed changes 
do not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or changes in the 
methods governing normal plant operation. The proposed changes will not 
impose any new or different requirements. Thus, this change does not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of safety. 
The proposed changes will not reduce a margin of plant safety because 
the methodology have been shown to ensure that the P/T and LTOP limits 
in the PTLR continue to meet all necessary requirements for reactor 
vessel integrity. These changes are administrative in nature since the 
limits were previously relocated to the PTLR under a separate LAR 
[License Amendment Request]. As such, no question of safety is 
involved, and the change does not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Rochester Public Library, 115 
South Avenue, Rochester, NY 14610.
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment requests: December 22, 1995.
    Description of amendment requests: The licensee proposes to delete 
the physical protection program reporting requirement from License 
Condition 2.G, and to clarify in License Condition 2.E that all the 
documents composing the physical protection program plans may not 
contain safeguards information.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 59922]]

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.
    This proposed change is considered an administrative change. It has 
no impact on the probability or consequences of any of the accidents 
previously evaluated. This change revises license conditions for 
clarification and removes the burden of duplicate reporting 
requirements. This change does not affect the physical protection 
program as previously approved by the Nuclear Regulatory Commission 
(NRC). License Condition 2.E is being revised to clarify that the 
physical security, security force training and qualification, and 
safeguards contingency plans may or may not contain safeguards 
information. The security force training and qualification plan does 
not currently contain safeguards information.
    A reporting requirement in License Condition 2.G is being revised 
to remove the reference to License Condition 2.E for the physical 
protection program. The reporting requirements for the physical 
protection program are located in the regulations, 10 CFR 73.71 and 10 
CFR 73 part, Appendix G.
    Therefore, the probability and consequences of an accident 
previously evaluated are not affected by these proposed changes.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change is considered an administrative change. It has 
no impact on equipment, systems, or structures such that a new or 
different kind of accident is created. This change revises license 
conditions to clarify that safeguards information may be located in the 
physical protection program plans and to remove duplicate and 
unnecessary reporting requirements for the physical protection program. 
There is no change associated with the implementation and maintenance 
of the physical protection program as previously approved by the NRC.
    Therefore, the possibility of a new or different kind of accident 
from an accident previously evaluated is not created.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    This proposed change is considered an administrative change only. 
It has no impact on the margin of safety associated with the physical 
protection program. This change revises license conditions to clarify 
the location of safeguards information in the physical protection 
program plans and remove duplicative and unnecessary reporting 
requirements for the physical protection program. The maintenance and 
implementation of the physical protection program is not affected by 
this change.
    Therefore, there will not be a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, CA 92713.
    Attorney for licensee: T.E. Oubre, Esquire, Southern California 
Edison Company, P.O. Box 800, Rosemead, CA 91770.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: October 22, 1997.
    Description of amendment request: The amendment would change the 
Perry Nuclear Power Plant design basis as described in the Updated 
Safety Analysis Report. The change will add a description of the 
temperature control valves and associated bypass lines around the 
Emergency Closed Cooling System heat exchangers. These features are 
designed to ensure operability of the Control Complex Chilled Water 
System under post-accident load conditions, without the need for 
compensatory actions.
    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 amendment is requesting Nuclear Regulatory Commission 
(NRC) review and approval of changes to the Perry Nuclear Power Plant 
(PNPP) Updated Safety Analysis Report (USAR) to incorporate 
descriptions (in the form of text, tables and drawings) of a 
modification to the plant involving two temperature control valves and 
associated temperature elements, and piping segments that have been 
installed in the Emergency Closed Cooling Water (ECC) System. These 
valves, temperature elements, and piping segments were installed to 
increase the overall reliability of the ECC System and the other safety 
related plant systems that it serves, to help ensure that they perform 
their specified safety functions without reliance on manual throttling 
actions.
    The probability of occurrence and the consequences of an accident 
previously evaluated in the USAR are not considered to be increased as 
a result of the temperature control valve modification.
    Based on conformance with the original system design criteria, the 
fact that the ECC System is an accident mitigation system, and that 
this modification does not introduce any new initiators to a previously 
postulated accident, the addition of this temperature control function 
can not increase the probability of occurrence of an accident 
previously evaluated in the USAR. Accidents reviewed involve the Loss 
of Coolant Accident applications described in USAR Chapter 6 with their 
corresponding consequence postulations shown in USAR Chapter 15, 
accident and transient scenarios as described in USAR Chapter 15, 
flooding and rupture postulations as described in USAR Chapter 3, and 
fire protection analyses as described in USAR Chapter 9.
    The modification has been designed, procured, and installed to the 
original design codes and standards. The modification also satisfies 
single failure criteria and does not adversely affect the mitigation 
function of the ECC System. Therefore, the ability to mitigate 
accidents previously evaluated in the USAR is maintained and the 
radiological consequences of such accidents remain unaffected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of previously evaluated 
accidents.
    2. The proposed change would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The modification has been designed to satisfy the requirements of 
the original ECC System. A single failure of the new configuration will 
not result in more than the loss of one respective

[[Page 59923]]

ECC System loop as already analyzed. Analysis of flooding shows no 
scenario greater than the currently bounding event. Missile generation 
is not a concern since no mechanisms conducive to that potential have 
been introduced. From the electrical analysis perspective, analysis has 
shown no adverse effects on the Emergency Diesel Generator loadings or 
other system applications.
    Based on the above discussions, the proposed change would not 
create the possibility of a new or different kind of accident than 
those previously evaluated.
    3. The proposed change will not involve a significant reduction in 
the margin of safety.
    This request does not involve a significant reduction in a margin 
of safety. The modification, including design, procurement, and 
installation, has been performed in accordance with the applicable 
codes, standards, and installation specifications. The modification 
does not change the heat removal capabilities or any previously 
designed parameters of the ECC System. Hence, the ECC System margin of 
safety with respect to safety classification, protection, redundancy, 
heat removal capability, and seismic classification remains unaffected.
    The margins of safety contained in the Technical Specifications and 
the associated Bases also remain unaffected by this modification due to 
conformance with the applicable codes, standards, and installation 
specifications. Specifically, Technical Specification 3.7.10, 
``Emergency Closed Cooling Water (ECCW) System'' and the description in 
the Bases remain unchanged and fully applicable. The following 
Technical Specifications also remain unaffected and applicable: 
3.3.3.2, ``Remote Shutdown System''; 3.7.1, ``Emergency Service Water 
(ESW) System--Divisions 1 and 2''; 3.7.4, ``Control Room Heating, 
Ventilation, and Air Conditioning (HVAC) System''; and the Technical 
Specifications related to Sections 3.8 (Electrical Power Systems), 3.5 
(Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation 
Cooling (RCIC) System) and 3.6 (Containment Systems). On this basis, 
the margins of safety defined in the Technical Specifications remain 
unchanged.
    Therefore, the changes associated with this license amendment 
request do not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

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.

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

    Date of application for amendment: August 27, 1996, as supplemented 
December 18, 1996, January 17, February 18, March 27, April 4, April 
25, April 29, May 30, June 2, June 13, June 18, August 4, August 8, 
September 10, October 2 (RNP RA/97-0216), October 2, (RNP RA/97-0207), 
October 13, and October 21, 1997.
    Brief description of amendment: This amendment addresses a more 
restrictive change proposed by the licensee in minimum allowable 
containment pressure.
    Date of publication of individual notice in Federal Register: 
October 7, 1997 (62 FR 52362).
    Expiration date of individual notice: October 21, 1997.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, SC 29550.

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

    Date of amendment request: September 24, 1997.
    Brief description of amendment request: The proposed amendment 
would add a surveillance requirement in Section 3/4.5.1 to perform a 
monthly valve position verification for each of the four residual heat 
removal crosstie valves.
    Date of publication of individual notice in Federal Register: 
October 6, 1997 (62 FR 52162).
    Expiration date of individual notice: November 5, 1997.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of amendment request: September 29, 1997.
    Brief description of amendment request: The proposed amendment 
would change Technical Specification 3/4.11.1, ``Liquid Effluents--
Concentration.'' The proposed change adds a requirement to perform 
weekly sampling and monthly and quarterly composite analyses of the 
Station Service Water System when the Reactor Auxiliaries Cooling 
System is contaminated.
    Date of publication of individual notice in Federal Register: 
October 6, 1997 (62 FR 52161).
    Expiration date of individual notice: November 5, 1997.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

[[Page 59924]]

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 Nos. 50-325 & 50-324, 
Brunswick Steam Electric Plant, Units 1 & 2, Brunswick County, NC

    Date of amendment request: January 7, 1997, as supplemented on July 
25, 1997, August 27, 1997, and September 15, 1997.
    Brief description of amendment: The amendments correct an error 
involving the transposition of two of the reactor pressure vessel (RPV) 
pressure-temperature (P-T) limits curves between the Technical 
Specifications for the Brunswick Steam Electric Plant, Units 1 and 2 
and update the hydrostatic pressure test limits curves for both units.
    Date of issuance: October 7, 1997.
    Effective date: October 7, 1997.
    Amendment No.: 189 and 220.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11485). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 7, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, NC 28403-3297.

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

    Date of application for amendment: August 27, 1996, as supplemented 
December 18, 1996, January 17, February 18, March 27, April 4, April 
25, April 29, May 30, June 2, June 13, June 18, August 4, August 8, 
September 10, October 2 (RNP RA/97-0216), October 2, (RNP RA/97-0207), 
October 13, and October 21, 1997.
    Brief description of amendment: This amendment addresses a more 
restrictive change proposed by the licensee in minimum allowable 
containment pressure.
    Date of issuance: October 24, 1997.
    Effective date: October 24, 1997.
    Amendment No.: 176.
    Facility Operating License No. DPR-23: Amendment revises the 
License and Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (62 FR 52362 dated October 7, 1997). The 
notice provided an opportunity to submit comments on the Commission's 
proposed NSHC determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by November 6, 
1997, but indicated that if the Commission makes a final NSHC 
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 determination of NSHC are contained in 
a Safety Evaluation dated October 24, 1997.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, SC 29550.

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

    Date of application for amendment: February 21, 1997.
    Brief description of amendment: This amendment adds a specific time 
limit to Technical Specification Table 3.3-3 to place an inoperable 
refueling water storage tank level channel in a bypassed condition.
    Date of issuance: September 30, 1997.
    Effective date: September 30, 1997.
    Amendment No: 74.
    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 Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, NC 27605.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, IL

    Date of application for amendments: March 5, 1997 as supplemented 
October 3, 1997.
    Brief description of amendments: The amendments would revise the 
Technical Specifications by removing the main steamline radiation 
monitor reactor scram function and the main steamline tunnel radiation 
isolation function.
    Date of issuance: October 24, 1997.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 163, 158.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 18, 1997 (62 FR 
19141). The October 3, 1997, submittal provided additional 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 
October 24, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, IL 60450.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 
1, West Feliciana Parish, LA

    Date of amendment request: August 29, 1996, supplemented August 29, 
1996 (proprietary), September 5, and October 8, 1997.
    Brief description of amendment: The amendment eliminates the 
Average Power Range Monitor (APRM) setpoint T-Factor setdown 
requirements and provides for reactivity anomaly calculation 
improvements. The request to decrease the local power range

[[Page 59925]]

monitor (LPRM) calibration frequency will be handled by separate review 
and action.
    Date of issuance: October 10, 1997.
    Effective date: October 10, 1997.
    Amendment No.: 100.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1997 (61 FR 
55032). The Licensee's letters dated August 29, 1996 (proprietary), 
September 5, and October 8, 1997, provided additional clarification and 
corrections to other TSs that would have erroneously referenced the TSs 
being eliminated and did not change the staff's initial no significant 
hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 10, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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

    Date of application for amendment: July 30, 1997, as supplemented 
September 19, and September 24, 1997.
    Brief description of amendment: The amendment reduces current 
technical specification leakage limit from the decay heat removal 
system from 6.0 gallons per hour (gph) to 0.6 gph.
    Date of issuance: October 15, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 205.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45458). The September 19, and September 24, 1997, submittals did not 
affect the initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 15, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

    Date of application for amendment: August 12, 1997, as supplemented 
August 28, September 15, October 3, 9, and 10, 1997.
    Brief description of amendment: The amendment changes the technical 
specifications surveillance requirements for once-through steam 
generator inservice inspection for Cycle 12 operation.
    Date of issuance: October 16, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 206.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45458). The supplemental letters did not affect the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 16, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, TX, Docket 
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda 
County, TX

    Date of amendment request: August 14, 1997, as supplemented 
September 23, 1997. The supplement provided clarifying information 
within the scope of the amendment request and did not change the 
initial no significant hazards consideration determination.
    Brief description of amendments: The amendments revise the allowed 
tolerance of the reactor coolant system volume provided in Technical 
Specification 5.4.2 to account for steam generator tube plugging.
    Date of issuance: October 20, 1997.
    Effective date: October 20, 1997.
    Amendment Nos.: Unit 1--Amendment No. 92; Unit 2--Amendment No. 79.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 26, 1997 (62 FR 
45278). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 20, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear 
Power Station, Unit 1, New London County, CT

    Date of application for amendment: February 7, 1997, as 
supplemented April 3 and September 19, 1997.
    Brief description of amendment: The amendment clarifies the 
requirement for calibration of instrument channels that use resistance 
temperature detectors or thermocouples.
    Date of issuance: October 22, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 90 days.
    Amendment No.: 102.
    Facility Operating License No. DPR-21: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17236). The April 3 and September 19, 1997, letters provided additional 
and clarifying information that did not change the scope of the 
February 7, 1997, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 22, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT, and at the Waterford Library, ATTN: Vince Juliano, 49 Rope 
Ferry Road, Waterford, CT.

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

    Date of application for amendment: June 19, 1997.
    Brief description of amendment: Technical Specification Table 2.2-1 
NOTES 1 and 3 define the values for the constants used in the 
Overtemperature Delta-T and Overpower Delta-T reactor trip system 
instrumentation setpoint calculators. The amendment makes changes to 
the NOTES as well as the associated Bases section.

[[Page 59926]]

    Date of issuance: October 22, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 152.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40852). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 22, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT, and the Waterford Library, ATTN: Vince Juliano, 49 Rope 
Ferry Road, Waterford, CT.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County, MI

    Date of application for amendments: November 6, 1996, as 
supplemented April 10 and October 1, 1997.
    Brief description of amendments: The amendments revise Technical 
Specifications governing the cooling water system and are a partial 
response to the licensee's application. The changes improve plant 
operation based on operational experience with the vertical motor-
driven cooling water pump. The changes also incorporate information 
gathered by the licensee during its self-assessment Service Water 
System Operational Performance Inspection (SWSOPI) completed in late 
1995. The remainder of the licensee's application will be addressed in 
a separate licensing action.
    Date of issuance: October 21, 1997.
    Effective date: October 21, 1997, with full implementation within 
90 days.
    Amendment Nos.: 131 and 123.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4338) The April 10 and October 1, 1997, letters provided clarifying 
information within the scope of the original application and did not 
change the staff's initial proposed no significant hazards 
considerations determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 21, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, MI 
55401.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket No. 
50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County, PA

    Date of application for amendment: June 30, 1997, as supplemented 
by letter dated September 26, 1997.
    Brief description of amendment: Revises the minimum critical power 
ratio (MCPR) safety limit in Section 2.1 of the Technical 
Specifications from 1.07 to 1.11 for two recirculation loops in 
operation. For a single loop in operation, the MCPR will change from 
1.08 to 1.12. The new MCPR safety limits reflect the effect of the new 
General Electric--13 part length fuel design and other Peach Bottom 
core-specific parameters.
    Date of issuance: October 9, 1997.
    Effective date: As of the date of issuance, to be implemented prior 
to startup from Unit 3 refueling outage 3R11.
    Amendment No.: 225.
    Facility Operating License No. DPR-56: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43373).
    The supplemental letter provided 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 October 9, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, PA

    Date of application for amendments: April 9, 1997.
    Brief description of amendments: These amendments revise the TSs to 
clarify existing battery-specific gravity requirements, delete the 
requirement to correct specific gravity values based on electrolyte 
level, and allow the use of charging current measurements to verify the 
battery's state of charge.
    Date of issuance: October 8, 1997.
    Effective date: Both units, as of date of issuance and shall be 
implemented within 30 days.
    Amendment Nos.: 123 and 88.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30643).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 8, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

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

    Date of amendments request: March 7, 1997.
    Brief Description of amendments: The amendments change the 
Technical Specifications for both Farley units to allow operability 
testing for certain containment isolation valves during defueled 
status.
    Date of issuance: October 17, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--130; Unit 2--123.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19834).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 17, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, AL 36302.

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M. 
Farley Nuclear Plant, Unit 1, Houston County, AL

    Date of amendment request: September 3, 1997.
    Brief Description of amendment: The changes reduce the number of 
required incore detectors necessary for continued operation for the 
remainder of Cycle 15 only.
    Date of issuance: October 23, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 131.

[[Page 59927]]

    Facility Operating License No. NPF-2: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 10, 1997 (62 
FR 47695).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 23, 1997.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, AL.

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

    Date of application for amendment: June 20, 1997.
    Brief description of amendment: Modify the Watts Bar Technical 
Specifications (TS) to incorporate the use of Code Case N-514 into the 
methodology for the Pressure-Temperature Limits Report.
    Date of issuance: October 21, 1997.
    Effective date: October 21, 1997.
    Amendment No.: 9.
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: September 10, 1997 (62 
FR 47700).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 21, 1997.
    No significant hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, KS

    Date of amendment request: July 3, 1997, as supplemented by letter 
dated August 20, 1997.
    Brief description of amendment: The amendment revises Surveillance 
Requirements 4.3.1.2 and 4.3.2.2, and Technical Specifications 3/4.3.1 
and 3/4.3.2, and associated Bases Sections B 3/4.3.1 and B 3/4.3.2 to 
eliminate periodic response time testing requirements for selected 
pressure and differential pressure sensors in the reactor trip system 
and engineered safety features actuation system instrumentation 
channels.
    Date of issuance: October 20, 1997.
    Effective date: October 20, 1997, to be implemented prior to 
restart from the ninth refueling outage currently scheduled to start on 
October 4, 1997.
    Amendment No.: 113.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40862).
    The August 20, 1997, supplemental letter provided additional 
clarifying information 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 
October 20, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, KS 66801 
and Washburn University School of Law Library, Topeka, KS 66621.

    Dated at Rockville, Maryland, this 29th day of October 1997.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 97-29138 Filed 11-4-97; 8:45 am]
BILLING CODE 7590-01-P