[Federal Register Volume 59, Number 109 (Wednesday, June 8, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-13765]
[[Page Unknown]]
[Federal Register: June 8, 1994]
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NATIONAL FOUNDATION ON THE ARTS AND THE HUMANITIES
U.S. NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 16, 1994, through May 26, 1994. The last
biweekly notice was published on May 25, 1994 (59 FR 27049).
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 Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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, 11555 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By July 8, 1994, 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 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 20555, and at the local public document
room for the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: December 30, 1993
Description of amendment requests: The proposed amendment would
modify Tables 3.3-4 and 3.3-5 of Technical Specification 3/4.3.2,
``Engineered Safety Features Actuation System Instrumentation,'' to
provide clarification of settings for undervoltage relay trip values
for the Class IE 4.16 kV electrical bus. The proposed amendment would
also add Figure 3.3-1, ``LOSS OF VOLTAGE RELAY (GE IAV) TIME VS VOLTAGE
CURVE,'' to clarify the relay setpoints and methodology.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1--Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed TS amendment does not significantly increase the
probability of an accident previously evaluated. The methodology
remains the same.
Clarifying the minimum acceptable voltage and allowing more
conservative values of the 4.16 kV bus undervoltage trip value will
ensure that 4.16 kV ESF [Engineered Safety Features] bus voltages
are sufficient to provide adequate voltage to equipment necessary
for accident response.
Standard 2--Create the possibility of a new or different kind of
accident from any accident previously analyzed.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously analyzed.
No new or different methodology is being proposed.
The minimum undervoltage relay setpoints are specified and the
new wording will allow a more conservative setpoint to ensure
voltage levels to equipment powered by the 4.16 kV ESF bus.
The proposed amendment also adds a figure to clarify the relay
setpoints and methodology.
Standard 3--Involve a significant reduction in a margin of
safety.
The margin of safety as defined in the TS will be increased by
specifying the minimum and allowing more conservative values for the
4.16 kV ESF bus undervoltage relay trip values. This clarification
will ensure that the trip setpoint adequately protects that
equipment powered by the 4.16 kV ESF bus from a potentially damaging
degraded voltage condition.
The NRC staff has reviewed the licensees' analysis and, based on
that 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: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: March 28, 1994
Description of amendment requests: The proposed amendment would
change Technical Specification 3/4.3.7.1.3 Condensate Storage Tank. The
licensee proposed to change the minimum condensate storage tank (CST)
indicated level from 25 feet to 29.5 feet to ensure that the CST
contains sufficient volume of water. In addition, the licensee proposed
to make an editorial change to the Unit 3 Technical Specification 3/
4.3.7.1.3, from ``with a level'' to ``with an indicated level,'' to be
consistent with the Units 1 and 2 Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1--Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This amendment request does not involve a significant increase
in the probability or consequences of an accident previously
evaluated based on the safety analysis for the CST [condensate
storage tank] minimum indicated level. The proposed change increases
the minimum indicated CST water level from 25 feet to 29.5 feet.
Increasing the minimum indicated CST water level ensures that the
requirements of BTP [Branch Technical Position] RSB [Reactor System
Branch] 5-1 and UFSAR [Updated Final Safety Analysis Report] Section
9.2.6 continue to be met. Therefore, this proposed change ensures
that the consequences of an accident previously evaluated are not
affected.
Standard 2--Create the possibility of a new or different kind of
accident from any accident previously analyzed.
This amendment request does not create the possibility of a new
or different kind of accident from any accident previously analyzed
since the minimum water volume of 300,000 gallons is maintained by
this change. The change in the CST minimum required water level does
not change the operation of any plant equipment while ensuring that
the required 300,000 gallons are available. Since this change does
not affect the operation of plant equipment and ensures that the
minimum CST water inventory is maintained, the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously analyzed.
Standard 3--Involve a significant reduction in a margin of
safety.
The margin of safety presently provided is not reduced by the
proposed change in the CST minimum required water level. The
proposed change ensures that the CST volume of 300,000 gallons is
available to satisfy the requirements of BTP RSB 5-1, UFSAR Section
9.2.6 and the BASES for Technical Specification 3/4.3.7.1.3.
Therefore, since the minimum required CST level in maintained, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensees' analysis and, based on
that 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: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: May 20, 1994
Description of amendment request: The proposed amendment will
change the Technical Specifications (TS) to (1) add an exception to TS
6.3.1 regarding the requirement for the Manager - Operations position
to hold a Senior Reactor Operator's (SRO's) license, (2) add a new TS
6.3.2 to describe the qualifications for the Manager - Operations and
Manager - Shift Operations positions. This new section will require the
Manager-Operations to hold or have held an SRO license for either the
H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson), or a
similar plant and require the Manager-Shift Operations to hold an SRO
license at the Robinson plant, and (3) renumber the sections to allow
for the insertion of a new TS 6.3.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The total number of senior reactor operator
licensed personnel on shift remains unchanged. The change to provide
a middle level of management will lessen the burden of the daily
shift operations placed on the Manager - Operations. This will allow
for more effective overall management of the Operations Unit.
Requiring the Manager shift Operations to hold an SRO license will
assure that supervision of operator activities continues to be
performed by a senior licensed individual. The proposed changes to
the Operations management organization do not involve physical
alterations of the plant configuration or changes in setpoints or
operating parameters. Therefore, there would be no increase in the
probability or consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As described above, these changes affect the organization
of the Operations Unit. They do not represent any appreciable change
in the current methodologies; they merely update the TS to reflect
current personnel organization configuration and standards. The
proposed changes to the Operations management organization do not
involve physical alterations of the plant configuration or changes
in setpoints or operating parameters. Therefore, the changes
proposed do not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety. The changes proposed do not
reduce the number of senior reactor operator licensed personnel on
shift. The changes to the Operations organization, as reflected in
the proposed change, will enhance the overall effectiveness of plant
operations and will serve to improve nuclear safety. There are no
changes to the plant configuration or changes in setpoints or
procedures. Therefore, the changes proposed have no affect on the
facility's 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: Hartsville Memorial Library,
Home and Fifth Avenues, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: May 10, 1994
Description of amendment request: The proposed amendment would
remove component lists from Technical Specification (TS) sections in
accordance with the guidance provided in Generic Letter (GL) 91-08
dated May 6, 1991. Related TS which reference the lists are also
modified.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change will not result in any hardware or operating
changes. The proposed change is based upon Generic Letter 91-08 and
merely removes component lists, removes details relating to the
component lists, provides clarifying information supporting the
removal of the component listings, or removes details (which are
considered administrative) that are no longer applicable to the TS.
The removal of tabular component listings from the TS does not
impact affected component OPERABILITY requirements. TS will continue
to require the components to be OPERABLE. Action statements and
surveillance requirements for the components will also remain in the
TS. The tabular component lists will be relocated to plant
procedures which will be controlled in accordance with the
provisions specified in the Administrative Controls Section of the
TS. Therefore, this change is administrative in nature and does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. Further, the proposed changes do
not alter the design, function, or operation of the components
involved and therefore, do not affect the consequences of any
previously evaluated accident.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed changes will not impose any different
operational or surveillance requirements. The changes propose to
relocate these component lists to plant procedures whereby adequate
control of information is maintained. Further, as stated above, the
proposed changes do not alter the design, function, or operation of
the components involved and therefore, no new accident scenarios are
created.
3. The proposed changes do not involve a significant reduction
in a margin of safety. The proposed change will not reduce a margin
of safety because it has no impact on any safety analysis
assumption. The proposed change does not alter the scope of
equipment currently required to be OPERABLE or subject to
surveillance testing nor does the proposed change affect any
instrument setpoints or equipment safety functions. 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 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: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226
NRC Project Director: Ledyard B. Marsh
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: July 15, 1993
Description of amendment request: The revision proposed by
Technical Specification Change Request (TSCR) No. 211 to the Technical
Specifications would remove a footnote to specification 3.24, ``Reactor
Coolant Inventory Trending System,'' and would revise that
specification to be consistent with the revised Babcock & Wilcox
revised Standard Technical Specifications, issued as NUREG-1430.
The July 15, 1993, request supersedes the request of September 5,
1991, published in the Federal Register on November 13, 1991 (56 FR
57697).
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 revised Limiting Condition for Operation represents no
increase in the probability of occurrence or consequences of an
accident previously evaluated.
The TSCR represents no change to the physical configuration or
operation of the Reactor Coolant Inventory Trending System.
As stated in the Bases for the existing Technical Specification
3.24, the system is not a system required to mitigate accidents. It
may be useful to have the system operable, but adverse impact does
not result if it is not operable; other useful information for
monitoring inadequate core cooling is available. The change proposed
is in accordance with the Limiting Condition for Operation in NUREG
1430.
2. The revised Limiting Condition for Operation does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
As identified above, the TSCR represents no change to the
physical configuration or operation of the Reactor Coolant Inventory
Trending System.
3. The revised Limiting Condition for Operation does not involve
a significant reduction in a margin of safety.
The margin of safety for the proposed Limiting Condition for
Operation is no different from that for existing Technical
Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: April 11, 1994
Description of amendment request: The purpose of the request is to
revise the Technical Specifications (TS) by relocating the detailed
inspection criteria, methods and frequencies of the containment tendon
surveillance program to the Final Safety Analysis Report (FSAR) and
providing a direct reference to the existing tendon surveillance
program in the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed amendment only relocates the
tendon surveillance program detailed requirements and criteria to
the FSAR, consistent with the BWOG [Babcock & Wilcox Owners Group]
Revised Standard Technical Specifications. The proposed amendment
does not affect the requirement to verify the containment structural
integrity in accordance with the inservice tendon surveillance
program. The proposed Technical Specification specifies that the
tendon surveillance program conforms to the recommendations of U.S.
NRC Regulatory Guide 1.35, which ensures the effectiveness of the
program is not reduced and containment structural integrity is not
affected. Therefore, this change does not increase the probability
of occurrence or the consequences of an accident previously
evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The tendon
surveillance program is required to conform to the recommendations
of U.S. NRC Regulatory Guide 1.35. Therefore, the effectiveness of
the surveillance program is maintained, thus providing continued
assurance of containment structural integrity.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The performance of the inservice tendon surveillance program
is maintained in conformance with the recommendations of U.S. NRC
Regulatory Guide 1.35, thus providing continued assurance of
containment structural integrity. Therefore, it is concluded that
operation of the facility in accordance with the proposed amendment
does not involve a reduction in a margin of safety as defined in the
basis of any Technical Specification.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: April 19, 1994
Description of amendment request: The purpose of the request is to
delete the quality assurance (QA) audit program frequency requirements
from the Technical Specifications (TS) and to utilize the Operational
Quality Assurance (OQA) Plan as the controlling document.
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. These changes do not affect the function of any system or
component. Therefore, they do not increase the probability of
occurrence or the consequences of an accident previously evaluated
in the SAR [Safety Analysis Report].
2. These changes do not involve a physical change to plant
configuration and they do not affect the performance of any
equipment. Therefore, they do not create the possibility of a new or
different kind of accident or malfunction of a different type than
previously identified.
3. The shifting of the audit frequency requirements from the
Technical Specifications to the OQA Plan and the extension of the
maximum interval between audits of certain areas do not change the
activities to be audited nor the scope of individual audits.
Furthermore, audit frequencies are not associated with the marginof
safety in the bases of any Technical Specification. Therefore, the
margin of safety in not affected by this change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: April 14, 1994.
Description of amendment request: The licensee proposes to revise
Technical Specification 5.3.1, ``Fuel Assemblies.'' The amendment would
permit fuel assembly reconstitution to restore the usefulness of fuel
assemblies containing damaged or leaking fuel rods.
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 a previously evaluated
accident.
Fuel assemblies containing filler rods will be shown to meet the
current nuclear, mechanical, and thermal-hydraulic design limits on
a cycle-specific basis. Since fuel assemblies containing filler rods
will be shown to meet the current nuclear, mechanical, and thermal-
hydraulic design limits on a cycle-specific basis, there is no
impact on the design basis of the plant.
Replacement of fuel rods with fuel rods containing natural
uranium or fuel rods from fresh or burned assemblies will be
evaluated by South Texas Project's internal 10 CFR 50.59 review
process. Reconstituted fuel assemblies with fuel rods containing
natural uranium and/or fuel rods from fresh or burned assemblies
will be analyzed using the normal design methodology as described in
Section 6.9.1.6 of the Technical Specifications.
Since current reload core design limits will be met by fuel
assemblies using zirconium alloy or stainless steel filler rods, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind accident from any previously evaluated.
The use of filler rods does not involve any alteration to plant
equipment or procedures which would introduce any new or unique
operational modes or accident precursors. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The use of filler rods does not change the reload design or
safety analysis limits for a reload core. Their use will be
evaluated on a cycle-specific basis using NRC-accepted fuel rod
configurations and analysis techniques. Since the safety analysis
limits are unaffected and since the modified fuel assemblies will be
shown to meet existing design limits on nuclear, mechanical, and
thermal-hydraulic parameters, 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: James E. Lyons, Acting
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: April 29, 1994
Description of amendment request: The licensee proposes to revise
Technical Specification 3.7.1.1, ``Turbine Cycle - Safety Valves.'' The
amendment would change the maximum allowable power range neutron flux
high setpoint when one or more main steam safety valve (MSSV) is
inoperative. The Bases for Technical Specification 3.7.1.1 would also
be changed to reflect the new algorithm used to calculate the revised
setpoint values.
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 a previously evaluated
accident. The Main Steam Safety Valves [MSSVs] are only actuated
after a transient has occurred. Therefore, revising the maximum
power level with inoperable MSSVs would not increase the probability
of a previously evaluated accident. Following a loss of load/turbine
trip, the revised Technical Specification Table 3.7-1 maximum
allowable power range neutron flux high setpoints would ensure that
the maximum power level allowed for operation with inoperable MSSVs
is below the heat removing capability of the operable MSSVs. This
would ensure that the design limit of 110% overpressurization is not
exceeded. Therefore, there is no increase in the consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change would not alter the design, configuration,
or method of operation of STP [South Texas Project]. For this
reason, as well as the reasons stated in response to Criterion 1
above, the proposed change would 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 the margin of safety.
The revised Table 3.7-1 setpoint values were calculated using a
conservative method where the maximum power level allowed for
operation with inoperable MSSVs is below the heat removing
capability of the operable MSSVs. Using the revised maximum plant
operating power levels will ensure that the secondary system
pressure will be limited to within 110% of its design pressure.
Therefore, since the design criteria will continue to be met, 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: Suzanne C. Black
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: April 27, 1994
Description of amendment request: The amendment would delete
Technical Specifications Surveillance Requirement 4.4.1.1.1, which
requires that the reactor recirculation pump discharge valve be
demonstrated operable by performing a full-stroke test of the valve
prior to reactor thermal power exceeding 25% of rated reactor thermal
power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed change to delete the TS Surveillance Requirement
(SR) does not require any modifications to the plant or equipment,
and will not impact the operation of the reactor recirculation
system. The reactor recirculation system will continue to function
as designed to maintain reactor pressure boundary integrity and to
provide sufficient flow through the reactor core to remove heat from
the fuel. This proposed TS change does not affect the operation of
any Emergency Core Cooling System (ECCS) or other plant equipment
important to safety. The purpose of this TS SR is to satisfy an ECCS
operability requirement for Boiling Water Reactor (BWRs) where the
reactor recirculation system piping serves as the injection flowpath
to the reactor pressure vessel for the Low Pressure Coolant
Injection (LPCI) system, an ECCS. Each LPCI subsystem, at LGS, has
an independent flowpath which does not rely on the reactor
recirculation system piping for injecting to the reactor pressure
vessel. The reactor recirculation system pump discharge valves do
not perform an active safety-related function and are classified as
passive safety-related components designed to maintain the reactor
pressure boundary integrity during reactor recirculation pump
maintenance activities. These valves are not normally used during
plant operations except to establish normal shutdown cooling, a
manually initiated non-safety related function. These valves are not
used to mitigate the consequences of design bases accidents.
Therefore, the proposed change does not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This proposed TS does not require physical changes to the plant
or equipment, and does not impact any design or functional
requirements of the reactor recirculation system, LPCI system, or
other plant systems important to safety. The purpose for this TS SR
is to satisfy an ECCS operability requirement for BWRs where the
reactor recirculation system piping serves as the injection flowpath
to the reactor pressure vessel for the LPCI system. The LPCI system
at LGS has an independent flowpath for injecting to reactor pressure
vessel and does not rely on the reactor recirculation system piping
as part of the injection flowpath. Since the intent of this
requirement is to support LPCI operation, and the LPCI system design
function is accident mitigation, eliminating this TS SR has no
impact on the types of accidents that could occur. The reactor
recirculation system pump discharge valves do not perform an active
safety-related function and are classified as passive safety-related
components designed to maintain the reactor pressure boundary
integrity during reactor recirculation pump maintenance activities.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed change to delete the TS SR does not involve a
change to the physical design or functional requirements of the
reactor recirculation system, LPCI system, or other plant system
important to safety. The reactor recirculation system will continue
to function as designed to maintain reactor pressure boundary
integrity and to provide sufficient flow through the reactor core to
remove heat from the fuel. This proposed TS change does not impact
the safety-related operation of the LPCI system. The LPCI system
will continue to function as designed to mitigate the consequences
of an accident. These valves do not perform an active safety-related
function and are classified as passive safety-related components
designed to maintain the reactor pressure boundary integrity during
reactor recirculation pump maintenance activities.
Therefore, the proposed TS change does not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Charles L. Miller
Philadelphia Electric Company, Public Service Electric and Gas
Company,Delmarva Power and Light Company, and Atlantic City
Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: April 27, 1994
Description of amendment request: These technical specifications
(TS) changes are being proposed to support the implementation of
proposed Modification 5274 which is intended to replace the Peach
Bottom Atomic Power Station (PBAPS), Unit 2 Containment Atmospheric
Dilution (CAD) System and Containment Atmospheric Control (CAC) System
analyzers. This modification was performed on PBAPS, Unit 3 during the
previous Unit 3 outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The design function and operation of the CAC and CAD Systems,
which are supported by the operation of the containment monitoring
system, have not been altered as a result of these changes. The CAC
System monitors the content of oxygen during startup and normal
operation and the CAD System is utilized to monitor the content of
hydrogen and oxygen during post-LOCA [loss-of-coolant accident]
operation. The monitoring of these variables will continue to
mitigate the consequences of accidents previously evaluated.
Additionally, no accident precursors will be impacted by these
changes.
The new system meets or exceeds the design standards of the
original system. Additionally, the decrease in warmup time will
increase the availability and usefulness of the analyzers to
mitigate the consequences of an accident. Therefore, the proposed
changes will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. Create the possibility of a new or different type of accident
from any accident previously evaluated; or,
The proposed TS changes do not involve the introduction of any
new accident initiators. The new containment monitoring system will
enhance the ability of [the] CAD system to mitigate the consequences
of an accident and prevent the introduction of a new or different
type of accident previously evaluated. The new system meets or
exceeds existing design standards and will be tested to ensure its
reliability. The new containment monitoring system is a monitoring
system and will not introduce new accident initiators. Therefore,
the proposed changes will not create the possibility of a new or
different type of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Although the number of analyzers is being reduced, the proposed
modification and TS changes will enhance the ability of the
containment monitoring system to support the operation of the CAC
and CAD systems [through] the use of improved equipment that meets
or exceeds the design standards of the original system. Therefore,
the proposed changes will not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Charles L. Miller
Power Authority of The State of New York, Docket No. 50-286, Indian
PointNuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: May 10, 1994
Description of amendment request: The proposed amendment would
revise Section 3.1.C.3 and Table 4.1-1 of Appendix A of the Operating
License. This change would require that the reactor coolant average
temperature (Tavg) be restored to greater than or equal to
540 deg. F within a 15-minute period or be in hot shutdown within the
following 15 minutes. The proposed change in Table 4.1-1 entitled,
``Minimum Frequencies for Checks, Calibrations and Tests,'' will add
the requirement for Tavg instrument check frequency to be reduced
to 30 minutes when the Tavg deviation and low Tavg alarms are
not reset and the control banks are above zero steps. This application
also proposes revision to the Bases to reflect that the minimum
temperature for criticality provides assurance that the reactor is
operated within the bounds of the safety analyses. In addition, the
proposed application also includes an administrative change to correct
some typographical errors on page 3.1-25 of the Technical
Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The probability or the consequences of an
accident previously evaluated will not be affected because the
proposed changes will make the Minimum Temperature for Criticality
Specification (540 deg. F) more restrictive than the current
specification which allows reactor criticality at a temperature as
low as 450 deg. F. The proposed changes will also make the minimum
temperature for criticality consistent with the licensing basis
safety analyses. In addition, critical operation at Tavg less
than 540 deg. F will require operator response to restore Tavg
to [greater than or equal to] 540 deg. F within 15 minutes or be in
hot shutdown within the following 15 minutes. [***] [T]he minimum
temperature for criticality when defined to be within 7 deg. F below
the no-load Tavg value of 547 deg. F does not adversely affect
pressurizer operability, reactor vessel nil-ductility temperature,
the reactor protection system operability, nor the plant design
basis analyses and is supported by the current licensing basis
safety analyses. The presence of two separate alarms, each
annunciating on a 1-out-of-4 Tavg signal, will provide
assurance that constant Tavg monitoring is available during
approaches to criticality. The proposed change also increases the
surveillance frequency for Tavg instrument check when the
Tavg deviation and low Tavg alarms are not reset and the
control banks are above zero steps. Therefore, the proposed changes
have no effect on the probability or consequences of an accident
previously evaluated.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
The proposed license amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed changes do not involve the addition of any
new or different type of equipment, nor do they involve the
operation of equipment required for safe operation of the facility
in a manner different from those addressed in the Final Safety
Analysis Report. The safety analyses, which assume a critical
temperature of 547 deg. F, are applicable for critical temperatures
as low as 540 deg. F. The proposed changes will ensure that the
plant parameters are within their analyzed ranges and will increase
the surveillance frequency for the Tavg instrument check when
the Tavg deviation and low Tavg alarms are not reset and
the control banks are above zero steps. Therefore, the proposed
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
:Response:
The proposed license amendment does not involve a significant
reduction in a margin of safety. The proposed changes do not affect
any safety related system or component operation or operability,
instrument operation, or safety system setpoints and do not result
in increased severity of any of the accidents considered in the
safety analyses. Operator response to a drop in temperature after
reaching criticality for a specified period of time will place the
reactor in the hot shutdown condition where the LCO [limiting
condition for operation] does not apply. The proposed changes are
being made to make the Technical Specifications consistent with the
licensing basis safety analyses and increase the surveillance
frequency. These changes have no effect on any margin of safety and,
therefore, do not create 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: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Michael L. Boyle, Acting
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: March 31, 1994
Description of amendment request: This amendment request would
revise the diesel fuel oil storage operability requirements and the 5
minute diesel hot restart test conditions. In addition, the amendment
request also revises the Surveillance Requirements to allow the 24-hour
diesel generator endurance test to be conducted during any operational
condition.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The addition of a 48 hour period to complete restoration of the
required fuel oil level prior to declaring the diesel generator
inoperable does not significantly increase the probability or
consequences of an accident previously evaluated. PSE&G believes
that the attendant risk of maintaining the diesel generator OPERABLE
status under temporary conditions where fuel oil supply is below
48,800 gallons (but greater than 44,709 gallons) is less than the
attendant risk of initiation and completion of shutdown actions
currently required by Technical Specifications under these
conditions. Since a minimum 44,709 gallon [6 day] supply of oil will
be maintained for these 48 hours, and procedures are implemented to
obtain replenishment fuel oil when the level falls below 48,800
gallons of fuel, and the probability of an event requiring the
onsite power sources during this brief period are low (as stated in
NUREG-1433), PSE&G concludes that this change does not increase the
likelihood of accidents occurring nor significantly affect the
performance of any system involved in the occurrence or mitigation
of the accidents.
The proposed amendment to allow the 24 hour diesel generator
endurance run to be conducted during any mode of operation does not
significantly increase the probability or consequences of an
accident previously evaluated in Chapter 15 of the UFSAR since the
capability to safely shutdown the plant following a Loss of Offsite
Power (LOP), LOCA or LOCA/LOP coincident with a single failure is
maintained throughout the surveillance test. The 24 hour endurance
test does not disable any of the automatic actuations and interlocks
of the diesel generator control functions, nor prevent the
satisfactory completion of the LOP or LOCA/LOP loading sequence if a
LOP or LOCA signal is received at any time during the test. Required
Class-1E onsite power OPERABILITY during normal operation, shutdown
cooling, loss of off-site power, and accident conditions will be the
same.
In addition, the performance of proposed Surveillance
Requirement 4.8.1.1.2.k.1 during Operational Conditions 1 or 2 will
not significantly increase the consequences of perturbations to any
of the electrical distribution systems that could result in a
challenge to steady state operation or to plant safety systems.
Performance of proposed Surveillance Requirement 4.8.1.1.2.k.1
during Operational Conditions 1 or 2, or failure of the
surveillance, will not cause, or result in, an anticipated
operational occurrence with attendant challenges to plant safety
systems that has not been previously analyzed for the existing
monthly surveillances.
Therefore, PSE&G concludes that this above change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The establishment of the new Surveillance Requirement
4.8.1.1.2.k.2, allows the five minute diesel hot restart test to be
performed at times other than after the 24 hour endurance run test,
but does not alter nor modify the test requirements currently
required by Surveillance Requirement 4.8.1.1.2.h.4.b. The proposed
Surveillance Requirement 4.8.1.1.2.k.2 maintains the test conditions
required by current Specifications, and satisfies the intent of
Regulatory Guide 1.9, Rev. 3, paragraph 2.2.10. Therefore, PSE&G
believes that this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed amendment does not involve any physical changes to
plant structures, systems or components, or change the manner in
which the plant is operated. Therefore, the proposed changes will
not increase the probability of accidents of a different type, nor
will they create malfunctions of a different type than any
previously evaluated in the SAR.
3. Will not involve a significant reduction in a margin of
safety. The basis for this statement is outlined in Item 1 above.
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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: March 31, 1994
Description of amendment request: The amendment request would
establish an allowed out-of-service time (AOT) of 72 hours for any one
low pressure Emergency Core Cooling System injection subsystem
inoperable in addition to an inoperable High Pressure Coolant Injection
system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below
:1. Will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change establishes a 72 hour Allowed Out-of-Service
Time (AOT) for coincidental High Pressure Coolant Injection system
(HPCI) and one low pressure injection/spray subsystem inoperability,
in accordance with NUREG-1433. In addition, PSE&G will justify
retaining the existing HCGS AOTs for one, two, and three Low
Pressure Coolant Injection (LPCI) subsystems inoperable.
The HCGS electrical distribution system that supplies the ECCS
(Emergency Core Cooling System) equipment with power contains four
(4) Emergency Diesel Generator (EDGs) while the electrical
distribution system of the NUREG-1433 plant contains only three (3)
EDGs. This additional, redundant system adds reliability to the
overall electrical distribution system. In NUREG-1433, the loss of
one EDG (Emergency Diesel Generator) results in the inoperability of
two (2) Low Pressure Injection (LPCI) pumps or one (1) LPCI and one
(1) 100% capacity CS subsystem, whereas the loss of one EDG at HCGS
would only result in the inoperability of one LPCI pump and one-half
of a CS subsystem (50% capacity). Another design feature of the HCGS
LPCI system that makes it more reliable than the plant discussed in
NUREG-1433 is that the LPCI system injects directly into the reactor
vessel via water boxes mounted inside the core shroud. The LPCI
system of the NUREG-1433 plant injects into the reactor vessel via
the recirculation loops, 2 LPCI pumps inject into each recirculation
loop. Therefore, following a rupture of a
recirculation line, a Design Base Accident Loss of Coolant
Accident, the NUREG-1433 plant would lose 2 LPCI subsystems, while HCGS
would maintain all 4 LPCI subsystems available for injection. The end
result is that, following assumed loss of a single EDG electrical
system, the HCGS plant is left with much more injection capability than
the NUREG-1433 plant. For these reasons:
*the HCGS AOT for one LPCI subsystem inoperable, given at least
one CS subsystem is operable, is 30 days as opposed to 7 days as
specified in NUREG-1433, and
*the HCGS AOT for two LPCI subsystems inoperable, given at least
one CS subsystem is operable, is 7 days. This AOT is the same as the
AOT in NUREG-1433 for one (1) LPCI subsystem inoperable. However,
this AOT is less restrictive than the action required in NUREG-1433
for one (1) LPCI subsystem inoperable with one (1) CS subsystem
inoperable, which is immediate entry into Technical Specification
3.0.3., and
*the HCGS AOT for three LPCI subsystems inoperable, given both
CS subsystems are operable, is 72 hours as opposed to immediate
entry into Technical Specification 3.0.3 as specified in NUREG-1433.
NUREG-1433 would be overly conservative for HCGS, since HCGS has
four (4) EDGs and four (4) physically separate channels of
electrical power, while the NUREG-1433 plant only has 3 EDGs. HCGS
has greater capability to supply power to its ECCS in an emergency
than the NUREG-1433 plant. Similarly, the incorporation of an AOT of 72
hours for HPCI and one low pressure injection/spray subsystem as
proposed in NUREG-1433 would also be conservative for HCGS.
In the HCGS safety analysis, the small break LOCA is the design
basis event for the HPCI system. The most limiting single failure
coincident with a large or small break LOCA is the failure of the DC
source (Channel A) common to the HPCI system, one CS subsystem, and
one LPCI subsystem. The change made by this submittal would
establish a 72 hour AOT for either one LPCI subsystem or one CS
subsystem to be inoperable coincident with HPCI being inoperable. As
discussed above, if an initiating event occurs while in the proposed
AOT for the ECCS, the resulting scenario would be within the design
basis of the HCGS. Even with a failure of one CS subsystem while
HPCI and LPCI are in a 72 hour LCO or a failure of one LPCI
subsystem while HPCI and CS are in a 72 hour LCO, HCGS would still
be within its design basis if an initiating event occurred. Adequate
core cooling is ensured by the operability of the ADS and the
remaining low pressure injection/spray subsystems.
The loss of feedwater flow transient analyzed in Chapter 15 of
the HCGS UFSAR assumes both the HPCI and RCIC systems to be
operable. The existing TS for HPCI allows the system to be
inoperable for up to 14 days. Sufficient injection capability is
assured because the RCIC system will be required to be operable if
the HPCI is inoperable. For this transient, the proposed 72 hour AOT
for the HPCI and one low pressure injection/spray subsystem
inoperable presents no challenge greater than the existing 14 day
HPCI AOT.
For the steam line breaks outside of containment, the HPCI
system is assumed unavailable. For feed line breaks outside of
containment, either the HPCI or RCIC systems are capable of
providing adequate cooling to the vessel to prevent cladding damage.
The RCIC system will be required to be operable during the proposed
72 hour AOT.
The Core Damage Frequency (CDF) at HCGS is calculated based on
Probabilistic Risk Assessment (PRA) models using present Technical
Specifications that do not allow simultaneous outages of either HPCI
and one LPCI subsystem or HPCI and one CS subsystem. The proposed
action statement, which would allow 72 hours of plant operation with
the HPCI system inoperable coincident with one low pressure ECCS
injection/spray subsystem (LPCI or CS) inoperable, was determined to
have a potential impact on 28 of the core damage sequences. These 28
core damage sequences were requantified with a new model (based on
the proposed action statement that would allow the simultaneous
outage), and none of the frequencies of the 28 sequences were
significantly affected.
In addition, the probability of an accident is not affected
because no physical modifications are being made to the plant.
Finally, implementation of this proposed change would reduce
unnecessary plant shutdowns without commensurate effects on safety
and minimize associated challenges to safety systems.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Establishing an AOT of 72 hours for the HPCI system inoperable
in addition to any one (1) low pressure injection/spray subsystem
does not alter the function of the equipment nor involve any type of
plant modification. Additionally, no new modes of plant operation
are involved with these changes. The proposed change therefore will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Will not involve a significant reduction in a margin of
safety.
The basis for this statement is outlined in item 1 above.
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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: April 25, 1994
Description of amendment request: The proposed amendment would
change the minimum level in the Emergency Diesel Generator (EDG) Fuel
Oil Day Tank from 200 gallons to 360 gallons. The change is necessary
to support a required run time of 55-60 minutes from the day tank. The
Bases would also be changed to include reference to Amendment 59, which
authorized a change to the EDG surveillance 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 operation of the Hope Creek Generating Station (HCGS) in
accordance with the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment to the Technical Specifications will
bring the minimum allowable EDG fuel oil day tank levels into
agreement with the levels corresponding to EDG run times reviewed
and approved by the NRC in Amendment 59 to the Hope Creek Generating
Station Facility Operating License. Increasing the minimum level
requirement is conservative and when compared to the current level
requirement, will act to reduce the consequences of any accident or
malfunction of equipment important to safety previously evaluated in
the Updated Final Safety Analysis Report (FSAR). The proposed change
will have no impact on the probability of any accident.
2. The operation of the Hope Creek Generating Station (HCGS) in
accordance with the proposed change will not create the possibility
of a new or different kind of accident from any previously
evaluated.
The physical changes to the plant and to the manner in which the
plant is operated in the proposed revisions add conservatism to the
current requirements in the Technical Specifications. No new or
different kind of accident is created by the proposed change.
3. The operation of the Hope Creek Generating Station (HCGS) in
accordance with the proposed change does not involve a significant
reduction in a margin of safety.
The proposed revision, by maintaining a higher (more
conservative) Emergency Diesel Generator Day Tank level, will result
in an increase in margins 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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: April 25, 1994
Description of amendment request: The proposed amendment would
eliminate the requirement from the Hope Creek technical specifications
to perform Type C leak rate tests in accordance with 10 CFR 50,
Appendix J, of identified containment isolation valves that penetrate
the primary containment and terminate below the minimum water level in
the suppression chamber (torus). The valves would still be subject to
testing in accordance with the American Society of Mechanical Engineers
Boiler and Pressure Vessel Code and Type A leak rate tests (Integrated
Leak Rate Test) of the primary containment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The Containment Isolation Valves (CIVs) for which Appendix J,
Type C leak rate testing will no longer be performed are all on
lines which penetrate the Torus and terminate below the Torus
minimum water level. Since the Torus is designed and operated to be
filled with water during and following any postulated Design Basis
Accident, the CIVs will remain water sealed during these conditions.
Type C testing of individual CIVs per the requirements of
10CFR50 Appendix J is not necessary since no potential containment
bypass leakage path exists due to the water seal and closed system
piping. The CIVs will, however, continue to be tested pursuant to
the applicable requirements of Section XI of the American Society of
Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code.
In addition, the elimination of 10CFR50 Appendix J, Type C
testing for the subject CIVs will not affect the overall leak-tight
integrity of the Primary Reactor Containment, which will be
demonstrated during 10CFR50 Appendix J Type A Integrated Leak Rate
Testing as described in the Hope Creek Generating Station Updated
Final Safety Analysis Report (HCGS UFSAR) Section 6.2.6.1.
Consequently, radiological releases and their consequences due
to leakage of the subject CIVs will be minimized and within the
existing plant licensing basis.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
This proposal does not involve any hardware or logic changes,
nor does it alter the way in which any plant systems operate. Post-
accident Containment isolation features, boundaries and system
interfaces are not affected by the changes. Therefore, there are no
new possibilities or types of accidents considered.
3. Will not involve a significant reduction in a margin of
safety.
The proposed elimination of 10CFR50 Appendix J, Type C testing
for certain CIVs in lines which penetrate the Torus and terminate
below the minimum water level will not adversely affect the margins
of safety associated with the plant's licensing bases.
The water seal provided by the Torus, in conjunction with the
closed system piping, precludes post-accident bypass leakage.
Individual CIVs will be tested in accordance with the ASME B&PV Code
Section XI - Division 1, Article IWV-3000, as required. 10CFR50
Appendix J, Type A testing will ensure that the overall Containment
leakage rate is consistent with the plant's licensing bases. In
addition, any leakage associated with the subject CIVs will have
little radiological significance since it will involve liquid
releases (i.e., minimum fission products). CIV seat leakage will be
confined within the closed system piping downstream of the CIVs.
Existing Containment isolation features, boundaries and system
interfaces are not affected by the changes.
Since the CIVs and the systems they serve are all located in the
Reactor Building, any leakage (e.g., packing gland leakage) which
escapes the confines of the closed system piping, will be contained
within the Reactor Building. The Reactor Building is a
radiologically controlled area which is served by the Filtration,
Recirculation, and Ventilation System. This assures that all
radioactive releases to the environment are within the existing
plant licensing bases. The elimination of Type C testing for the
subject isolation valves will not affect the existing radiological
release evaluations currently described in the HCGS-UFSAR.
The proposed Amendment will not affect the functional capability
of any plant safety-related structures, systems or components nor
will it result in any relaxation of existing plant licensing bases.
In addition, the implementation of the proposed Amendment will
result in a reduction of radiological exposure to plant personnel.
Therefore, the proposed revision will not reduce 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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: March 28, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.3.2, Table 3.3-4, Item 7b, Sustained
Degraded Voltage for Salem Generating Station Units 1 and 2. The
amendment proposes to change the Sustained Degraded Voltage Trip
Setpoint from greater than or equal to 91.6% of bus voltage for less
than or equal to 13 seconds to greater than or equal to 94.6% of bus
voltage for less than or equal to 13 seconds, and the Allowable Value
for Sustained Degraded Voltage from greater than or equal to 91% of bus
voltage for less than or equal to 15 seconds to greater than or equal
to 94% of bus voltage for less than or equal to 15 seconds.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident or malfunction of equipment important to
safety previously evaluated.
The proposed amendment will increase the minimum voltage
available at the vital buses and maintain vital loads within their
voltage ratings under degraded voltage conditions.
The change to the Sustained Degraded Voltage Trip Setpoint also
considered the minimum bus recovery voltage following a transient
that would reset the undervoltage relay to prevent unnecessary
transfers to the emergency diesel generators. Therefore, the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated in
the UFSAR.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed amendment to the Sustained Degraded Voltage Trip
Setpoint ensures a higher minimum voltage is available to all vital
loads during any electrical transient, and is sufficient to operate
these loads within their voltage rating. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any previously evaluated.
3. Will not involve a significant reduction in a margin of
safety.
The proposed amendment will increase the minimum voltage
available at the vital buses and maintain vital loads within their
voltage ratings. This ensures that the minimum voltage for any vital
load will continue to be available including under degraded voltage
conditions. Calculations have determined that a Sustained Degraded
Voltage Trip Setpoint of greater than or equal to 94.6% of bus
voltage for less than or equal to 13 seconds is the optimum setpoint
with an Allowable Value of greater than or equal to 94% of bus
voltage for less than or equal to 15 seconds to meet the current
design requirements. These values provide sufficient margin based on
the minimum voltage of 93.2% required to ensure that vital loads
will operate within their designed voltage range. Therefore, the
proposed amendment does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this 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: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, and
Docket No. 50-354, Hope Creek Generating Station, Salem County, New
Jersey
Date of amendment request: April 13, 1994
Description of amendment request: The proposed changes revise the
Quality Assurance audit frequencies in the Hope Creek and Salem Unit
Nos. 1 and 2 Technical Specifications. These revisions will permit a
biennial audit frequency with a 25% extension to the frequency and
transfer subsequent control over the audit program to the Updated Final
Safety Analysis Report. Basis for proposed no significant hazards
consideration determination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The likelihood that an accident will occur is neither increased
or decreased by the proposed Technical Specification changes which
affect review and audit frequencies. The proposed changes will not
impact the function or method of operation of plant equipment. Thus,
there is not a significant increase in the probability of a
previously analyzed accident due to the changes. Also, the
consequences of a malfunction of equipment important to safety
previously evaluated in the UFSAR is not increased by the changes.
The proposed changes affect review and audit frequencies. As
such, the proposed changes have no impact on accident initiators or
plant equipment, and thus, do not affect the probabilities or
consequences of an accident.
The audit program verifies that functions and methods of plant
operation have not been altered or degraded. This verification
process will continue, but on a more flexible basis. The proposed
changes will not alter the function or diminish the quality of the
audits. Frequency may only be decreased for those programs which
demonstrate acceptable performance. The changes provide added
flexibility in scheduling audits and facilitate resource allocation
to areas with perceived weaknesses. In this regard, the proposed
changes offer an opportunity for possible decreases in the
likelihood that an accident would occur.
Therefore, we conclude that the proposed changes do not
significantly increase the probabilities or consequences of an
accident.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed changes do not involve changes to the physical
plant or operations. Since program audits do not contribute to
accident initiation, changes related to audit functions cannot
produce a new accident scenario or produce a new type of equipment
malfunction. Also, the [changes do] not alter any existing accident
scenarios. The proposed changes do not affect equipment or its
operation, and, thus, do not create the possibility of a new or
different kind of accident.
The proposed changes revise the audit frequency to biennial and
permit a 25% extension to this frequency. However, audit frequency
may only be decreased for those programs exhibiting acceptable
performance. The 25% extension is a provision to provide added
flexibility in scheduling audits and is not applicable to the audits
of the Facility Emergency and Security Plans.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident.
3. Will not involve a significant reduction in a margin of
safety.
The proposed changes concerning conduct of reviews and audits do
not directly affect plant equipment or operation. Safety limits and
limiting safety system settings are not affected.
The proposed changes will not alter the function or diminish the
quality of the reviews and audits. For the audit program, the
changes propose a biennial audit frequency and a subsequent transfer
of control over the audit program to the Updated Final Safety
Analysis Report. This will not result in a loss of regulatory
control. 10 CFR 50.54(a)(3) requires that changes to the quality
assurance description report which reduce commitments be submitted
to the NRC prior to implementation. In addition, the audit frequency
may only be decreased for those programs which demonstrate
acceptable performance. The changes provide added flexibility in
scheduling audits and facilitate resource allocation to areas with
perceived weaknesses. The proposed changes should result in a more
effective audit program that will contribute to an improvement in
plant safety.
Therefore, use of the proposed Technical Specification changes
would not involve any 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: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079 (Salem Generating Station, Units
1 and 2) and Pennsville Public library, 190 S. Broadway, Pennsville,
New Jersey 08070 (Hope Creek Generating Station)
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: May 3, 1994
Description of amendment request: The proposed amendments would
modify the Technical Specification (TS) for Combustible Gas Control (3/
4.6.4.1) by changing the surveillance frequency for performing the
channel functional test to once-per-quarter and the channel calibration
to once-per-refueling. Also, the TS for the Auxiliary Feedwater System
(3/4.7.1.2) would be changed to reduce the surveillance frequency for
performing pump operability tests to once-per-quarter on a staggered
test basis. These changes are consistent with the provisions of Generic
Letter 93-05, ``Line-Item Technical Specifications Improvements to
Reduce Surveillance Requirements For Testing During Power Operation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed changes to the surveillance requirements for the
hydrogen analyzers and the Auxiliary Feedwater pumps are consistent
with the intent of Generic Letter 93-05, Line-Item Technical
Specification Improvements to Reduce Surveillance Requirements for
Testing During Power Operation, and NUREG-1366, Improvements to
Technical Specification Surveillance Requirements. The proposed
changes will modify surveillance frequency for both the hydrogen
analyzers and the Auxiliary Feedwater Pumps. Changing the
surveillance frequency for the hydrogen analyzers and the Auxiliary
Feedwater pumps does not affect the probability of occurrence or the
consequences of accidents identified in the UFSAR. No accident
precursors are being generated by the proposed increase in
surveillance frequency. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
a previously analyzed accident.
2. Create the possibility of a new or different kind of
accident.
The proposed changes to the surveillance requirements for the
hydrogen analyzers and the Auxiliary Feedwater pumps are consistent
with Generic Letter 93-05 and NUREG-1366. There are no modifications
or changes in operating conditions associated with the proposed
changes. Therefore, the proposed changes will not increase the
possibility of a new or different kind of accident from any accident
previously identified.
3. Involve a significant reduction in a margin of safety.
The Technical Specification operability requirements for the
hydrogen analyzers and the Auxiliary Feedwater pumps are not being
changed. Surveillance testing will still be performed on a routine
frequency. The proposed frequency will be capable of performing its
intended function and ensuring a consistent degree of reliability.
Therefore, the changes to the surveillance frequencies do not
involve a significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: September 29, 1993 (TS 333)
Description of amendment request: The proposed amendment would
increase the amount of Boron-10 required to be available in the Standby
Liquid Control System (SLCS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The Standby Liquid Control System (SLCS) is designed to make the
reactor subcritical from rated power to a cold shutdown at any time
in core life with the control rods remaining withdrawn in the rated
power pattern. The increase in the amount of Boron-10 that is
required to be stored in the SLCS Solution Tank does not affect the
precursors for any accident or transient analyzed in Chapter 14 of
the BFN Final Safety Analysis Report (FSAR). Since there is no
change to an accident precursor, there is no increase in the
probability of any accident previously evaluated.
The increase in the amount of Boron-10 required to be stored in
the SLCS Solution Tank restores the ability of the SLCS to maintain
the Boron concentration required to ensure cold shutdown for future
anticipated core configurations. The minimum concentration of
natural Boron in the reactor vessel required to achieve cold
shutdown for the currently anticipated future core configurations is
660 ppm. The 660 ppm concentration was recommended by General
Electric in Service Information Letter (SIL) No. 325, June 1980.
The increase in the amount of Boron-10 required to be stored in
the SLCS Solution Tank restores the ability of the SLCS to maintain
the Boron concentration required to ensure cold shutdown for future
anticipated core configurations. The increased Boron-10 storage
requirements does not represent a change to the designed Boron
concentration capability of the SLCS. Since this change will ensure
the ability of the SLCS to mitigate the consequences of an accident
for future anticipated core designs, the change does not involve a
significant increase in the consequences of any accident previously
evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The increase in the amount of Boron-10 required to be stored in
the SLCS Solution Tank does not affect the function or operation of
any other system. The proposed change does not introduce any new
modes of operation or modify existing equipment design. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The SLCS shutdown margin is determined for each core
configuration by using the BWR simulator code to calculate the core
multiplication for the cold, xenon-free, all-rods-out condition at
the exposure point of maximum cold reactivity. The resulting k-
effective is subtracted from the critical k-effective of 1.0 to
obtain the SLCS shutdown margin. Increasing the amount of Boron-10
stored in the SLCS Solution Tank increases the SLCS shutdown margin.
Therefore, the proposed amendment results in an increased 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: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: September 30, 1993 (TS 312)
Description of amendment request: The proposed amendment removes
Technical Specification requirements for scram discharge volume air
header pressure trip switches in the Reactor Protection System (RPS)
for the Browns Ferry Nuclear Plant (BFN) Unit 2. The amendment also
clarifies the description of the RPS scram discharge volume high water
level bypass in the Technical Specifications for BFN Units 1, 2, and 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The Unit 2 scram pilot air header pressure trip switches perform
the same function as the high water level switches in the scram
discharge instrument volume. They automatically initiate control rod
insertion (scram) in the event that degraded air conditions are
detected in the BWR control air supply system. Since the scram pilot
air header pressure trip function is to ensure that the Control Rod
Drive (CRD) System is available to mitigate the consequence of an
accident or transient, and the removal of the scram pilot air header
pressure e trip switches scram function does not effect the
precursors for any accident or transient analyzed in Chapter 14 of
the BFN Final Safety Analysis Report (FSAR), there is no increase in
the probability of any accident previously evaluated.
The scram system has been analyzed, based on a plant-specific
maximum inleakage, and the removal of the scram pilot air header
pressure trip switches scram function would still result in a
successful scram, provided that the CRD leakage rate and the water
level instrumentation response characteristics remain within the
success criteria region. Administrative controls and periodic
measurements of the CRD and scram discharge system performance
ensure acceptable response time delays and stall flow rates. Since
the scram function would be successfully performed, the removal of
the scram pilot air header pressure trip switches scram function
does not involve a significant increase in the consequences of any
accident previously evaluated.
The clarification of the description of the scram discharge
volume high water level bypass in the Reactor Protection System does
not reflect a modification to plant equipment, maintenance
activities, or operating instructions. The revised description does
not effect the precursors for any accident or transient analyzed in
Chapter 14 of the BFN FSAR or equipment used in the mitigation of
these accidents or transients. Therefore, there is no increase in
the probability of any accident previously evaluated nor an increase
in the consequences of any accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The design criteria for the scram system is contained in the
generic Safety Evaluation Report (SER), which was transmitted by NRC
letter to All BWR Licensees, dated December 9, 1980, BWR Scram
Discharge System. Section 4.2.4.1 of the generic SER states that the
CRD system shall be analyzed based on a plant-specific maximum
inleakage to ensure that the system function is not lost prior to
initiation of automatic scram. This analysis has been performed and
it was concluded that the removal of the scram pilot air header
pressure trip switches scram function would still allow a successful
scram, provided that the CRD leakage rate and the water level
instrumentation response characteristics remain within the success
criteria region. Administrative controls and periodic measurements
of the CRD and scram discharge system performance ensure acceptable
response time delays and stall flow rates.
The overall scram system design is in conformance with the
generic SER. No new system failure modes are created as a result of
removing the scram pilot air header pressure trip switches scram
function since, assuming the maximum CRD inleakage due to a degraded
air supply system, redundant and diverse level switches will
initiate a successful scram prior to this function becoming
disabled. The removal of the scram pilot air header pressure trip
switches scram function does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The clarification of the description of the scram discharge
volume high water level bypass in the Reactor Protection System does
not reflect a modification to plant equipment, maintenance
activities, or operating instructions. No new external threats,
system interactions, release pathways, or equipment failure modes
are created. Therefore, the clarification of this description does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The water level in the scram discharge instrument volume is
monitored by redundant Magnetrol and heated reference resistive
temperature devices (RTD) level switches. Since the level switches
are redundant and diverse, redundancy and diversity in the
instrumentation that initiates the scram signal is maintained even
with the removal of the scram pilot air header pressure trip
switches scram function.
The scram system has been analyzed, based on a plant-specific
maximum inleakage, and the removal of the scram pilot air header
pressure trip switches scram function would still result in a
successful scram, provided that the CRD leakage rate and the water
level instrumentation response characteristics remain within the
success criteria region. Since the scram system would successfully
operate to mitigate the consequences of accidents and transients
previously analyzed, the proposed amendment does not involve a
significant reduction in the margin of safety.
The clarification of the description of the scram discharge
volume high water level bypass in the Reactor Protection System does
not reflect a modification to plant equipment, maintenance
activities, or operating instructions. There is no change to the
licensing or design basis of the Reactor Protection System.
Therefore, the revised description does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and
50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone
County, Alabama
Date of amendment request: April 4, 1994 (TS322, Revision 1)
Description of amendment request: The proposed amendment supersedes
an amendment request dated March 25, 1993. Similar to the March 25,
1993, proposal, the revised amendment request for the Browns Ferry
Nuclear Plant (BFN), Units 1, 2, and 3, continues to seek elimination
of Technical Specifications (TS) for the reactor scram and main steam
line isolation functions associated with the main steam line radiation
monitors (MSLRMs). Additionally the revised amendment request proposes
elimination of TS for the remaining isolation functions associated with
the MSLRMs as well. The remaining isolation functions include main
steam line drain valve closure, reactor recirculation loop sample valve
closure, and main condenser mechanical vacuum pump (MVP) trip and
isolation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The probability of occurrence of previously evaluated accidents
is based on initial conditions and assumptions which are not
dependent directly or indirectly on the functions of the Main Steam
Line Radiation Monitors (MSLRMs). Elimination of the MSLRM scram and
isolation functions does not affect the operation of the other
Reactor Protection System or Primary Containment Isolation System
functions required to mitigate a Rod Drop Accident (RDA). Also,
eliminating the MSLRM functions does not affect the Control Rod
Drive System and, hence, cannot increase the probability of RDA. The
proposed change does not involve any increase in the probability of
an [any] previously evaluated accident.
There is no significant increase in the consequences of any
previously evaluated accident. Elimination of the MSLRM reactor
scram and isolation functions could potentially increase the amount
of radioactivity released following an RDA. However, the potential
increase in release from those previously determined is not
significant. Both the NEDO-31400 analysis and the additional BFN
dose calculations determined that the resulting doses (See Table B
[of this request]) remain well below the limits of Standard Review
Plan (SRP) 15.4.9 and 10 CFR 100.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This proposed change deletes the reactor scram and isolation
functions of the MSLRMs. The sole purpose of these functions is to
assist in mitigating the consequences of an RDA, a previously
analyzed event. This event is terminated by a high flux scram.
The NEDO-31400 RDA analysis without the MSLRM scram and MSIV
[Main Steam Isolation Valve] closure functions has been reviewed and
accepted by the NRC. An RDA without the MSLRM MVP isolation function
has been previously reviewed by the NRC as documented in Section
14.6.2 of the Updated Final Safety Analysis Report (UFSAR). An RDA
without the MSLRM recirculation sample line isolation coupled with a
sample line break is no different than that if the normal RWCU
[Reactor Water Cleanup] [system] sample path is in service (see
Section III.C) [of this request]. However, TVA has evaluated the
consequences of this event and determined the MSLRM sample line
isolation function is not required to maintain radioactive releases
within the acceptable limits.
The NEDO-31400 RDA analysis and the additional BFN offsite dose
calculations show that the elimination of the MSLRM functions does
not create the possibility of a new or different kind of
accident from any accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
A reliability assessment of the elimination of the MSLRM scram
function on reactivity control failure frequency and core damage
frequency was performed as part of the NEDO-31400 analysis. The
results of the NEDO-31400 analysis indicated a negligible increase
in reactivity control failure frequency associated with deletion of
the MSLRM scram function. However, this increase is offset by a
reduction in the frequency of transient initiating events
(inadvertent scrams). This reduction in transient initiating events
represents a net reduction in core damage frequency of 0.3 percent.
BFN TS Bases 3.2 states that the MSLRMs are provided to detect
gross fuel failures as in an RDA and provide main steam isolation
valve closure to maintain radiological releases below the 10 CFR 100
limits. The BFN offsite dose calculations and the NRC's Safety
Evaluation Report approving the NEDO-31400 RDA analysis document
that the potential radiological release consequences following an
RDA without the MSLRMs reactor scram and isolation functions are
still well within the 10 CFR 100 limits. Thus, this 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
thisreview, 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: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 16, 1994 (TS 93-18)
Description of amendment request: The proposed change would revise
Surveillance Requirement 4.8.1.1.1.b to reflect installation of common
station service transformers with auto load tap changers and the
resulting change to the alignment of this portion of the distribution
system. The proposed amendment would replace the terms ``normal'' and
``alternate'' with ``unit generator supported'' and ``preferred power
(GDC 17),'' respectively, circuits that must be demonstrated operable.
Basis for proposed no significant hazards consideration
determination: Asrequired by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
TVA has evaluated the proposed technical specification (TS)
changeand has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The description clarification for the circuits in TS 4.8.1.1.1.b
does not change any TS requirement and will not alter plant
functions or components as a result of normal alignment to the
common station service transformers (CSSTs). By implementing normal
CSST alignment to unit power, SQN has reduced the required transfers
to obtain alignment to the preferred power circuit under accident
conditions. The verifications provided by TS Surveillance
4.8.1.1.1.b will continue to ensure the capability of this transfer
during occasional use of the unit station service transformer (USST)
for maintenance or repair activities. The accidents described in
Chapter 15 of the SQN Updated Final Safety Analysis Report assume
the unit power supply to be from the USST circuit that requires
transfer to the CSST after turbine trip. This analysis is bounding
for normal alignment to the CSST that does not require a transfer.
These changes minimize the impact of malfunctions because transfers
will not be required as often. The consequences of an accident are
therefore not increased by this change. The proposed unit power
alignment to the CSSTs has always been used for unit start-up; and
since no plant functions are changed as a result of normal alignment
to the CSSTs, other than a lower potential for transfers, the
probability of an accident is not increased.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The TS 4.8.1.1.1.b clarification does not change plant functions
and only provides a more appropriate description of the circuits
involved in the transfer function. The transfer from the USST
circuit to the CSST circuit will still be available when SQN is
using the USST circuit, but the need for these transfers will be
reduced by normal alignment of unit power to the preferred CSST
circuit. These changes will not create a new or different kind of
accident because the plant functions remain unchanged and only the
CSST alignment for normal unit power and circuit descriptions is
altered. This CSST alignment has always been used for unit start-up
and is being expanded to apply to normal power operation in addition
to start-up and accident conditions.
3. Involve a significant reduction in a margin of safety.
The clarification for the unit power circuit descriptions does
not alter the intent of requirements for this surveillance. Plant
functions and setpoints remain unchanged for automatic transfers
from the USST circuit to the CSST circuit. This circuit description
change provides a clearer explanation of how the transfer functions.
The margin of safety has not been altered for unit power alignment
to the USST or CSST, and this change will support SQN's use of the
CSST for the normal unit power circuit. Therefore, these changes do
not involve a significant reduction in a margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, 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: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: December 6, 1993
Description of amendment request: The proposed amendment would
remove the requirement to perform jet pump integrity and operability
surveillances in the idle loop during operation with one recirculation
pump. It also provides consistency between the Bases and Section 4.6.F
on jet pump surveillance requirements, and includes some administrative
changes within the Bases.
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:
Removing the surveillance requirement to examine the jet pump
flow patterns within the idle loop during single loop operation does
not increase the probability or consequences of an accident
previously evaluated. The probability of loss of the ability of such
a jet pump to maintain a refloodable volume of two thirds core
height is not increased. The stresses undergone without forced
recirculation flow are very small and therefore a jet pump within an
idle recirc loop should not physically degrade during this time. The
ability of the core standby cooling systems to provide their
intended function is not degraded by removal of this surveillance
requirement; therefore the consequences of the recirculation line
break accident are not increased.
This change will remove a surveillance requirement that has been
determined to be unnecessary based upon the operating conditions
during single recirculation loop operation. Other surveillance
requirements within the technical specifications adequately
determine jet pump operability when stresses exist which could cause
a jet pump failure to occur. The change will not affect any plant
hardware, plant design, safety limit settings, or plant system
operation, and therefore does not modify or add any initiating
parameters that would significantly increase the probability or
consequences of any previously analyzed accident.
As discussed above, there is no increase in the probability of a
jet pump failing and being undetectable with the proposed technical
specification surveillances. Therefore, there is no increase in the
possibility of not being able to maintain a floodable volume of two
thirds core height during the design basis LOCA [loss-of-coolant
accident]. As a result, there is no 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: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301
Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,
One International Place, Boston, Massachusetts 02110-2624
NRC Project Director: Walter R. Butler
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.
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of application for amendment: January 14, 1994
Brief description of amendment request: The proposed amendment
would revise the technical specifications (TSs) by removing component
lists from the TSs in accordance with NRC Generic Letter (GL) 91-08 and
by removing the schedule for withdrawal of reactor vessel material
specimen capsules from the TSs in accordance with GL 91-01.
Date of individual notice in Federal Register: May 23, 1994 (59 FR
26675)
Expiration date of individual notice: June 22, 1994
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2 Benton County, Washington
Date of application for amendment: December 6, 1993, supplemented
by letter May 6, 1994
Brief description of amendment request: The proposed amendment
would revise Technical Specifications (TS) 3/4.4.2 and 3/4.5.1 to
require main steam system and automatic depressurization system safety/
relief valve (SRV) surveillance testing within 12 hours after steam
pressure and flow have been found to be adequate to do the testing.
Also, TS Table 4.3.7.5-1 would be revised to require SRV position
indicator surveillance testing within 12 hours after steam pressure and
flow have been found to be adequate to do the testing.
Date of individual notice in Federal Register: May 13, 1994 (59 FR
25131)
Expiration date of individual notice: June 13, 1994
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2 Benton County, Washington
Date of application for amendment: May 5, 1994
Brief description of amendment request: The proposed amendment
would change the plant operating license to rename three primary
containment isolation check valves listed in the technical
specifications. The licensee is making an administrative change to
rename valve PI-EFC-X29d to make its number consistent with other
similar valves in the technical specifications. The licensee is
renaming excess flow check valves PI-EFCX-72f and PI-EFCX-73e because
they are replacing them with swing check valves that have a different
numbering nomenclature.
Date of individual notice in Federal Register: May 12, 1994 (59 FR
24762)
Expiration date of individual notice: June 13, 1994
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
rooms for the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: January 20, 1994
Brief description of amendments: The amendments increase the
departure from nucleate boiling ratio limit from 1.24 to 1.30 to
accommodate the uncertainties in core inlet flow distribution. The
amendments also add the analytical method supplement entitled ``System
80TM Inlet Flow Distribution'' to the list of methods used to
determine core operating limits.
Date of issuance: May 26, 1994
Effective date: May 26, 1994
Amendment Nos.: 76, 62 and 48
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12358) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 26, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: April 1, 1993
Brief description of amendments: The amendments make
administrative, editorial, and format changes to the Operating
Licenses. These changes include the deletion or incorporation, as
appropriate, of all handwritten or ``pasted-up'' changes and the
removal of all previous license conditions that have been completed to
the satisfaction of the Commission. The changes and reformatting result
in the Operating Licenses containing only those license conditions that
are currently applicable.
Date of issuance: May 20, 1993
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 189 and 166
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: May 12, 1993 (58 FR
28052) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated May 20, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: February 9, 1994
Brief description of amendment: This amendment changes the
Technical Specifications and the License. These changes consist of
revised wording for the License, clarifying wording to aid operators in
selecting the correct pressure/temperature curve during startup and
shutdown operations, and removal of certain obsolete mechanical snubber
acceptance criteria.
Date of issuance: May 16, 1994
Effective date: May 16, 1994
Amendment No.: 153
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12359) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 16, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
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 No. 1, Lake County, Ohio
Date of application for amendment: March 1, 1993
Brief description of amendment: The amendment revises the Technical
Specifications to provide specific actions to take if primary
containment leakage limits are exceeded and cannot be restored when
reactor coolant temperature is greater than 200 deg. F.
Date of issuance: May 26, 1994
Effective date: May 26, 1994
Amendment No. 60
Facility Operating License No. NPF-58. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14896) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 26, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of application for amendment: January 17, 1994
Brief description of amendment: The amendment removes Technical
Specification 3/4.4.12, ``Failed Fuel Rods'' and its associated Bases
Section 3/4.4.12.
Date of Issuance: May 17, 1994Effective date: As of the date of
issuance to be implemented within 30 days.
Amendment No.: 171
Facility Operating License No. DPR-61. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 16, 1994 (59
FR 7687) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated May 17, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: March 19, 1993
Brief description of amendment: The amendment changed Technical
Specification 4.18.6 and Table 4.18-2 to make the requirements for C-3
reports on steam generator tube inspections consistent with the Babcock
& Wilcox Standard Technical Specifications.Date of issuance: May 19,
1994Effective date: May 19, 1994
Amendment No.: 172
Facility Operating License No. DPR-51. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 21, 1993 (58 FR
39049) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 19, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: September 3, 1993
Brief description of amendments: These amendments eliminate the
reliance on the five ``cranking'' diesel generators by replacing the
motor driver on one of the standby steam generator feedwater pumps with
a dedicated diesel driver.
Date of issuance: May 20, 1994
Effective date: May 20, 1994
Amendment Nos. 164 and 158Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 13, 1993 (58 FR
52985) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 20, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: November 30, 1993
Brief description of amendment: The amendment revises the plant
Technical Specifications (TS) to remove the protective and maximum
allowable setpoint limits for axial power imbalance and the trip
setpoint for nuclear overpower based on reactor coolant system (RCS)
flow (flux-to-flow) from the TS and relocate them to the existing TMI-1
Core Operating Limits Report (COLR).
Date of Issuance: May 23, 1994
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 184
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2867). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 23, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: March 3, 1994
Brief description of amendment: The amendment revises the technical
specifications in accordance with the guidance provided by Generic
Letter 93-08, ``Relocation of Technical Specification Tables of
Instrument Response Time Limits.''
Date of issuance: May 19, 1994
Effective date: May 19, 1994
Amendment No.: 73
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12380) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 19, 1994.No significant
hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of application for amendment: January 21, 1994
Brief description of amendment: The amendment revised the Technical
Specifications by changing the name of one operator and one of the
owners of the Duane Arnold Energy Center from Iowa Electric Light and
Power Company to IES Utilities Inc. The title of the position
responsible for the management of the Nuclear Division has also been
changed to Vice President, Nuclear from Manager-Nuclear Division. One
blank page was deleted and several other editorial changes were made.
Pages 1 through 4 of the Facility Operating License were revised to
reflect the corporate name change and a spelling error.
Date of issuance: May 13, 1994
Effective date: May 13, 1994
Amendment No.: 198
Facility Operating License No. DPR-49. Amendment revised the
license and Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10008) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 13, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of application for amendment: December 22, 1993
Brief description of amendment: The amendment revised the Technical
Specifications by correcting several typographical and administrative
errors.
Date of issuance: May 18, 1994
Effective date: May 18, 1994
Amendment No.: 199
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10008) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 18, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401.
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: April 14, 1994, as
supplementedApril 20, 1994.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to change the laboratory testing protocol for the
charcoal absorbers for the Control Room Emergency Ventilation System
(TS 3.7.6.1), the Enclosure Building Filtration System (TS 3.6.5.1) and
the Storage Pool Ventilation System (TS 3.9.15).
Date of issuance: May 23, 1994
Effective date: As of the date of issuance to be implemented
within30 days.
Amendment No.: 175
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes (59 FR 23085, May 4, 1994) That
notice provided an opportunity to submit comments on the Commission's
proposed no significant hazards consideration determination. No
comments have been received. The notice also provided for an
opportunity to request a hearing by June 3, 1994, but indicated that if
the Commission makes a final no significant hazards consideration
determination any such hearing would take place after issuance of the
amendment.The Commission's related evaluation of the amendment, finding
of exigent circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated May
23, 1994.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County,Minnesota
Date of application for amendments: February 14, 1994
Brief description of amendments: The amendments revise Technical
Specifications to reflect the new configuration for the Unit 1 480V
safeguards bus arrangement (two 480V safeguards buses fed by each 4160V
safeguards bus). These changes make the specifications the same for
both units since the configuration for the two units will become the
same at the completion of the current outage.
Date of issuance: May 17, 1994
Effective date: May 17, 1994
Amendment Nos.: 110 & 103
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14892) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 17, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: January 6, 1993
Brief description of amendment: The amendment revised the Technical
Specifications (TS) to incorporate the following changes:
(1) The residual heat removal (RHR) pump flow calibration frequency
(specified in TS Table 4.1-1) was changed to accommodate operation on a
24-month cycle.
(2)The RHR loop isolation valve automatic isolation and interlock
testing frequency (specified in TS Table 4.1-3) was changed to
accommodate operation on a 24-month cycle.
(3)The RHR system leakage testing frequency (specified in TS
Section 4.4.I.4) was changed to accommodate operation on a 24-month
cycle.
(4)The recirculation pump testing frequency (specified in TS
Section 4.5.B.1.a) was changed to accommodate operation on a 24-month
cycle.
(5)The accumulator check valve operability testing frequency
(specified in TS Section 4.5.B.2.b) was changed to accommodate
operation on a 24-month cycle.
(6)The safety injection (SI)/RHR check valve gross leakage testing
frequency (specified in TS Section 4.5.B.2.c) was changed to
accommodate operation on a 24-month cycle.
These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to
Accommodate a 24-Month Fuel Cycle,'' as applicable.
In addition, the gross leakage surveillance requirements for
certain SI/RHR system check valves (specified in TS Section 4.5.B.2.d)
was changed to implement requirements as set forth in NRC generic
letter, dated February 23, 1980, regarding testing of low pressure
injection (LPI)/RHR check valves. Therefore, Item A.5 of the February
11, 1980, Confirmatory Order is considered rescinded.
Date of issuance: May 20, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 148
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 17, 1993 (58
FR 8777) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 20, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: October 18, 1993, and suplement
dated March 7, 1994
Brief description of amendment: This amendment revised TS 3/4.3.2
and associated Bases to increase surveillance test intervals and add
out-of-service times for isolation actuation instrumentation. The
changes are in accordance with General Electric Company's Licensing
Topical Reports which have been previously reviewed and approved by the
NRC staff.
Date of issuance: May 25, 1994
Effective date: As of date of issuance and shall be implemented
within 60 days of the date of issuance.
Amendment No.: 70
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64615) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 25, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: October 26, 1993 and
supplemented by letter dated December 14, 1993.
Brief description of amendment: The amendment request revised the
MINIMUM CHANNELS OPERABLE requirement for suppression pool water
temperature instruments, Accident Monitoring ACTION STATEMENTS, and
removes ACTIONS and Surveillance Requirements for suppression chamber
temperature and level instruments.
Date of issuance: May 25, 1994
Effective date: May 25, 1994
Amendment No.: 71
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67861) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 25, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama.
Date of amendments request: October 14, 1993
Brief description of amendments: The requested changes incorporate
changes allowing longer surveillance test intervals (STIs) and allowed
outage times (AOTs) for the reactor trip system (RTS) and engineered
safety features actuation system (ESFAS) instrumentation into the
Technical Specifications. The proposed changes also revise certain RTS/
ESFAS functions, minimum channels operable, channel calibration, and
channel functional test requirements to ensure they are in concert with
the Westinghouse Standard Technical Specifications and WCAP-10271,
``Evaluation of Surveillance Frequencies and Out-of-Service Times for
Reactor Protection Instrumentation Systems.''
Date of issuance: May 16, 1994
Effective date: May 16, 1994
Amendment Nos.: 107 and 99
Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (57 FR
17605) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 16, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns
Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama
Dates of application for amendments: April 6, 1992 (TS 308), and
September 28, 1992 (TS 326)
Brief description of amendments: The amendments change the
Technical Specifications (TS) to add Automatic Depressurization System
(ADS) high drywell pressure bypass timer requirements, revise the ADS
timer trip level setting, increase the number of ADS valves required to
be operable for startup, and revise the limiting conditions for
operation with inoperable ADS valves. The ADS bases have also been
revised for consistency with these TS changes.
Date of issuance: May 19, 1994
Effective date: May 19, 1994
Amendment Nos.: 205 and 178
Facility Operating License Nos. DPR-33 and DPR-68: Amendments
revised the Technical Specifications.
Dates of initial notice in Federal Register: May 27, 1992 (57 FR
22269) and November 25, 1992 (57 FR 55593).The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
May 19, 1994.No significant hazards consideration comments received:
None
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: February 9, 1994; supplemented
April 13, 1994 (TS 93-21)
Brief description of amendments: The amendments change Technical
Specification Table 3.3-11 to reflect the addition of ionization fire
detectors to Fire Zones 184, 185, 186, and 187.
Date of issuance: May 23, 1994
Effective date: May 23, 1994
Amendment Nos.: 181 and 173
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12369) The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated May 23, 1994. No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: October 1, 1993; supplemented
March 29, 1994 (TS 93-09)
Brief description of amendments: The amendments revise the
setpoints and time delays for the Auxiliary Feedwater loss of power and
6.9 kv shutdown board loss-of-voltage and degraded-voltage
instrumentation. In addition, the description, total number of
channels, channels to trip, minimum channels operable, actions, trip
setpoints, allowable values, channel checks, and channel functional
test requirements for loss-of-power instrumentation have been revised.
Date of issuance: May 24, 1994
Effective date: May 24, 1994
Amendment Nos.: 182 and 174
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4947) The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated May 24, 1994.No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: March 30, 1994
Brief description of amendments: The proposed amendments revise the
technical specifications by adding an alternative method for verifying
that the emergency diesel generator fuel oil meets requirements.
Date of issuance: May 13, 1994
Effective date: May 13, 1994, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 1 - Amendment No. 24; Unit 2 - Amendment No.
10
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17606) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 13, 1994 No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: February 25, 1994
Brief description of amendments: The amendments revise the
surveillance frequency of the nozzles in the Quench Spray and
Recirculation Spray Systems from 5 to 10 years. The change is in
accordance with NRC Generic Letter 93-05, ``Line-Item Technical
Specifications Improvements to Reduce Surveillance Requirements at
Power Operation,'' dated September 27, 1993.
Date of issuance: May 16, 1994
Effective date: May 16, 1994
Amendment Nos.: 182 and 163
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17608) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 16, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: October 4, 1993
Brief description of amendments: These amendments revise the NA-1&2
Technical Specifications requirements to allow the use of ZIRLO
material for fuel cladding.
Date of issuance: May 26, 1994
Effective date: May 26, 1994
Amendment Nos.: 183 and 164
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57859) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 26, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: February 25, 1994
Brief description of amendments: The amendments revise the
surveillance frequency of the nozzles in the containment spray and
recirculation spray systems from 5 to 10 years.
Date of issuance: May 20, 1994
Effective date: May 20, 1994
Amendment Nos. 191 and 191
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17608) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 20, 1991.No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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 20555,
and at the local public document room for the particular facility
involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By July 8, 1994, 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 20555 and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
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, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Commonwealth Edison Company, Docket No. STN 50-457, Braidwood
Station, Unit No. 2, Will County, Illinois
Date of application for amendment: April 21, 1994
Brief description of amendment: The amendment effects a one-time
only change to Technical Specification (TS) Surveillance Requirement
4.7.1.1 by adding a note which relieves Braidwood, Unit 2, from
compliance with the provisions of TS 4.0.4 until initial entry into
Mode 2. This will permit Braidwood, Unit 2, to reach Mode 3 to reset
Main Steam Safety Valves (MSSVs) and proceed with a startup. The
amendment is applicable only until entry into Mode 2 following forced
outage A2F27.
Date of issuance: May 16, 1994
Effective date: May 16, 1994
Amendment No.: 51
Facility Operating License No. NPF-77. This amendment revised the
Technical Specifications. Public comments requested as to proposed no
siginificant hazards consideration: Yes. The NRC published a public
notice of the proposed amendment, issued a proposed finding of no
significant hazards consideration and requested that any comments on
the proposed no significant hazards consideration be provided to the
staff by the close of business on May 12, 1994. The notice was
published in the Joliet News Herald and the Morris Daily Herald on May
9, 1994. No comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the state of Illinois and
final no significant hazards consideration determination are contained
in a Safety Evaluation dated May 16, 1994.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
Local Public Document Room location: Wilmington Township Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481
NRC Project Director: James E. Dyer
Dated at Rockville, Maryland, this 1st day of June 1994.
For the Nuclear Regulatory Commission.
Gus C. Lainas,
Acting Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation.
[FR Doc. 94-13765 Filed 6-7-94; 8:45 am]
Billing Code 7590-01-F