[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