[Federal Register Volume 59, Number 61 (Wednesday, March 30, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-20330]
[[Page Unknown]]
[Federal Register: March 30, 1994]
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UNITED STATES 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 March 7, 1994, through March 18, 1994. The
last biweekly notice was published on March 16, 1994.
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 P-223, Phillips Building, 7920 Norfolk Avenue,
Bethesda, 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 April 29, 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: February 18, 1994
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) Figure 3.2-1, ``REACTOR COOLANT
COLD LEG vs CORE POWER LEVEL,'' of TS 3/4.2.6, ``REACTOR COOLANT COLD
LEG TEMPERATURE,'' for Units 1 and 3 to include the cold leg
temperature between 552 deg.F and 562 deg.F at core power levels
between 90 percent and 100 percent within the AREA OF ACCEPTABLE
OPERATION. Also, the proposed amendment would modify TS 3/4.1.1.4,
``MINIMUM TEMPERATURE FOR CRITICALITY,'' and BASES 3/4.1.1.4, ``MINIMUM
TEMPERATURE FOR CRITICALITY,'' for all units to allow the minimum
temperature for criticality to be established at 545 deg.F, rather than
the current value of 552 deg.F, to establish the surveillance
temperature at 552 deg.F, rather than the current 557 deg.F, and to
clarify the BASES for this TS.
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 does not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
analyses performed confirmed that the existing safety analysis for
cycle 5 of all three PVNGS [Palo Verde Nuclear Generating Station]
units remains valid for a 10 deg.F reduction in RCS [reactor coolant
system] temperature.
Standard 2 -- Create the possibility of a new or different kind of
accident from any accident previously evaluated.
This amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated. The
analyses performed demonstrated that the current licensing basis
analyses results remain valid with a 10 deg.F reduction in RCS [reactor
coolant system] temperature, and that the safety system settings remain
unchanged.
Standard 3 -- Involve a significant reduction in a margin of
safety.
This amendment request will not involve a significant reduction in
a margin of safety. There is no reduction in the margin of safety since
the changes apply only to the reactor coolant cold leg temperature and
the minimum temperature for criticality, the safety analyses have been
reevaluated (and reperformed where necessary) using the new
temperature, and the results remain valid. All other safety limits and
safety system settings remain unchanged. Therefore, there is no
reduction in any 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, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: February 4, 1994
Description of amendment request: The proposed amendment revises
the Action Statement of Technical Specification 3.6.5, Vacuum Relief
System, to require that in Modes 1-4 with one vacuum relief system
inoperable the system be restored to operable status within seventy-two
hours or be in at least hot standby within the next six hours.
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 proposed amendment does not physically alter the plant in
any manner. The proposed amendment does not introduce any new
equipment nor does it require any existing equipment or systems to
perform a different type of function than they are currently
designed to perform. The proposed amendment to Technical
Specification 3.6.5 allows additional time to restore an inoperable
containment vacuum relief system to operable status. Changing the
completion time to seventy-two hours remainsquite conservative for
this non-ESF system since a seventy-two hour restoration time is
specified for two-train ESF systems which mitigate Final Safety
Analysis Report (FSAR) Chapter 15 accidents. The CVRS [containment
vacuum relief system] is designed to protect the structural
integrity of containment during an inadvertent actuation of the
containment spray system, which is not an FSAR Chapter 15 accident.
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.
The proposed amendment does not introduce any new equipment nor
does it require any existing equipment or systems to perform a
different type of function than they are currently designed to
perform. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed amendment to Technical Specification 3.6.5 allows
additional time to restore an inoperable containment vacuum relief
system to operable status. Changing the completion time to seventy-
two hours remains conservative since a seventy-two hour restoration
time is specified for two-train ESF systems which mitigate FSAR
Chapter 15 accidents. The CVRS is designed to protect the structural
integrity of containment during an inadvertent actuation of the
containment spray system, which is not an FSAR Chapter 15 accident.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety as defined in the Technical
Specifications of FSAR.
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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: S. Singh Bajwa
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: February 24, 1994
Description of amendment request: The proposed amendments would
provide surveillance requirements for a planned modification to the
Keowee emergency power generators' underground power path breaker
closing logic. The planned modification would provide an automatic
close feature for the underground path breakers under certain specified
conditions. The modification is needed to correct a design deficiency
which resulted in a single failure vulnerability when both Keowee units
are in their normal alignment. The single failure vulnerability is
being prevented by means of administrative controls pending
implementation of a permanent corrective action. The proposed
amendments would add an annual operability test to Technical
Specification 4.6, Emergency Power Periodic Testing, of the automatic
close feature.
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 evaluated:
The Keowee Hydro units provide the main source of emergency
power for the Oconee Nuclear units, but they are not accident
initiators. The FSAR [Final Safety Analysis Report] Loss of Electric
Power Accident assumes two types of events: (1) Loss of load (unit
trip) and (2) Loss of all system and station power. The changes
performed by the modification that added the automatic closure
circuitry do not increase the likelihood of either. Also, the
modifications to the Keowee operating logic will not adversely
affect the ability to mitigate LOOP [Loss of Offsite Power], LOCA
[Loss of Coolant Accident], and LOCA/LOOP accidents as described in
the FSAR. The loss of all station power accident analysis
assumptions are still valid. This modification has no adverse impact
on the ability of the Keowee Units to satisfy their design
requirements to achieving rated speed and voltage within 23 seconds
of receipt of an emergency start signal.
The surveillance change that is included in [the] amendment
request is provided to assure the availability of the electrical
power systems for mitigation of Design Basis Accidents (DBAs). As
described within the technical justification [from the licensee's
application], the Keowee breaker circuitry was modified to allow the
Keowee Unit that is aligned to the overhead power path to
automatically close to the underground power path if the postulated
fault occurs. The surveillance change is an additional restriction
not presently included in the Technical Specifications. [The]
amendment will ensure the operability of the Keowee Unit ACB [Air
Circuit Breaker] automatic close feature and will assure that proper
testing requirements are maintained.
Based on the above and the technical justification provided in
[the amendment application], there is no significant increase in the
probability of a DBA as a result of this change, nor is there a
significant increase in the consequences of a DBA as a result of
this change since the proposed amendment assures the availability of
the electrical power system.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
The proposed change makes physical changes to the plant
configuration. However, the modification simply changes the Keowee
control logic to remove the possibility of a certain postulated
failure from causing a loss of emergency power to the Oconee nuclear
units. The Keowee emergency power systems will remain operable and
available to mitigate accidents. Operation of ONS [Oconee Nuclear
Station] in accordance with [the] Technical Specifications will not
create any failure modes not bounded by previously evaluated
accidents. Consequently, this change will not create the possibility
of a new or different kind of accident from any kind of accident
previously evaluated.
(3) Involve a significant reduction in a margin of safety:
Margins of safety associated with [the] Technical Specifications
have been evaluated. No safety or design limits are adversely
affected, so margins of safety as defined in the bases to any
Technical Specifications are not reduced as a result of the Keowee
modification. The design basis of the auxiliary electrical system is
to supply the required ES [Engineered Safeguards] loads of one Unit
and safe shutdown loads of the other two units. The Technical
Specification amendment includes an additional surveillance
restriction not presently included in the Technical Specifications.
The proposed amendment assures the continued availability of the
electrical power systems; thus preserving the existing margin of
safety. Therefore, there will be no 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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews, Director
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 7, 1993, as supplemented
February 8, 1994
Description of amendment request: The proposed amendment would
revise the Physical Security Plan (PSP).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
The accident mitigation features of the plant are not affected by
the proposed compensatory measures for protecting the site during
periods when security systems are degraded and therefore no decrease
occurs in the effectiveness of the security program to protect against
radiological sabotage or increased risk to the public health and
safety. This is due to continued compliance with existing regulatory
requirements and other commitments within the security plan. These
changes have no impact on the design basis security threat and
accordingly do not create the possibility of a new or different kind of
accident. New systems, modes of equipment operation, failure modes or
other plan situations are not introduced by these changes. The proposed
changes allow flexibility for the use of compensatory measures and do
not change any safety limits, LCOs, or surveillance requirements on
equipment to operate the plant.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 14, 1994
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to reflect changes that have
been made to 10 CFR Part 20 AND 10 CFR 50.36a.
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:
The proposed revisions to the liquid and gaseous concentration
release rate limits will not involve a significant increase in the
probability or consequences of an accident previously evaluated
because there will be no change in the types and amounts of
effluents that will be released, nor will there be an increase in
individual or cumulative occupational radiation exposures.
The administrative changes for definitions, terminology,
paragraph references, and record keeping requirements are necessary
so that the Waterford 3 Technical Specifications will remain
consistent with the revised federal regulations (i.e., 10CFR20 and
10CFR50.36a). Record retention and reporting requirements will
continue to meet NRC regulations. These changes are administrative
in nature and do not affect plant hardware or operation.
Restricting access to high radiation areas via guards rather
than locked doors provides operational flexibility while continuing
to meet the underlying intent of precluding unauthorized access.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
Changes to the liquid and gaseous concentration limits are
necessary to provide adequate operational flexibility. Operational
history at Waterford 3 has demonstrated that the use of
concentration values associated with the old 10CFR20.106
requirements has resulted in calculated maximum individual doses to
a member of the public that are small percentages of the limits of
10CFR50, Appendix I. The proposed revisions will not create the
possibility of a new or different kind of accident from any
previously evaluated because the revisions will not change the types
and amounts of effluent that will be released.
The administrative changes for definitions, terminology,
paragraph references, and record keeping are necessary so that the
Technical Specifications will remain consistent with the revised
federal regulations (i.e., 10CFR20 and 10CFR50.36a). Record
retention and reporting requirements will continue to meet NRC
regulations. These changes are administrative in nature and do not
affect plant hardware or operation.
Restricting access for ALARA [as low as reasonably achievable]
with guards rather than locked doors will continue to meet the
underlying intent of the TS. These changes do not involve plant
hardware or operation.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any previously
evaluated.
The proposed revisions do not involve any changes in the types
or increases in the amounts of effluents released off site. The
methodology used to control radioactive effluents and calculate
effluent monitor setpoints will result in the same effluent release
rate as the current methodology. The basic requirements for TS
concerning effluent releases (10CFR50.36a) indicate that compliance
with TS will keep average annual release to small percentages of
10CFR20 limits. For liquid effluent releases, the annual dose of 500
mrem, that is the bases for the concentrations in the new 10CFR20.
The 50.36a requirements further indicate that operational
flexibility is allowed, compatible with considerations of health and
safety, which may temporarily result in release higher than such
small percentages, but still within the limits specified in the old
10CFR20.106 that references Appendix B maximum permissible
concentrations (MPCs). For gaseous effluent releases, the limits
associated with the gaseous release rate TS will be maintained at
the current instantaneous dose rate limits. Compliance with the
limits of the new 10CFR20.1301 will be demonstrated by operating
within the limits of 10CFR50, Appendix I, and 40CFR190. The revision
will not change the types and amounts of effluent that will be
released.
The administrative changes for definitions, terminology,
paragraph references, and record keeping are necessary so that the
Technical Specifications will remain consistent with the revised
federal regulations (i.e., 10CFR20 and 10CFR50.36a). Record
retention and reporting requirements will continue to meet NRC
regulations. These changes are administrative in nature and do not
affect plant hardware or operation.
Controlling access to high radiation areas for ALARA can be
performed effectively by guards in place of locked doors. These
changes do not involve plant hardware or operation.
Therefore, the proposed changes will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Florida Power and Light Company, et al., Docket No. 50-335, St.
Lucie Plant, Unit No. 1, St. Lucie County, Florida
Date of amendment request: February 22, 1994
Description of amendment request: The proposed amendment modifies
the minimum stored borated water inventory requirements for Operational
Modes 1 through 4 by revising Figure 3.1-1 and Limiting Condition for
Operation (LCO) 3.1.2.8 of the unit Technical Specifications (TS). The
associated bases for TS 3/4.1.2 are also revised to reflect the
bounding borated water makeup volumes, as a function of boric acid
concentration, which define the proposed inventory requirements. The
proposed amendment will significantly improve operational flexibility
with no risk to plant safety and will provide for consistency of
operation between the two St. Lucie units.
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:
Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if operation of the facility in accordance with the
proposed amendment would not: (1) involve a significant increase in
the probability or consequences of an accident previously evaluated;
or (2) create the possibility of a new or different kind of accident
from any accident previously evaluated; or (3) involve a significant
reduction in a margin of safety. Each standard is discussed as
follows:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment will reduce the minimum borated water
inventory required to be stored in the Boric Acid Makeup Tanks
(BAMT) during unit operation in Modes 1 through 4. The reduction in
BAMT inventory will not affect any equipment postulated to
malfunction in the Updated Final Safety Analysis Report (UFSAR) to
initiate an accident nor will it impact the operation of any other
equipment whose malfunction could adversely affect safety-related
structures, systems, or components. Credit is not taken for boron
addition to the Reactor Coolant System from the BAMTs for purposes
of reactivity control in accidents analyzed in the UFSAR. The
minimum required capability to achieve and maintain safe shutdown
for such events has not been altered. Therefore, operation of the
facility in accordance with the proposed amendment will not involve
a significant increase in the probability or 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 reduction in minimum required BAMT inventory does not change
the boration system function, configuration, operation, or design
basis as described in the UFSAR. The proposed change does not alter
the modes of plant operation and does not affect the operation of
safety-related structures, systems, or components. Therefore,
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.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The reduced BAMT minimum inventory requirements are defined by
analyses that utilize an approved plant cooldown scenario and
conservative physics parameters representative of the present and
future planned reactor core designs for St. Lucie Unit 1. The
analytical methodology employed to determine the revised inventory
requirements is the same as that used to establish the existing
inventory requirements. The existing reactivity control Limiting
Conditions for Operation (LCO) related to safe shutdown margins and
redundant boron flow paths have not been altered. Sufficient
quantities of borated water will continue to be stored in the BAMTs
to assure compliance with these LCOs during the prescribed plant
operating modes. Therefore, operation of the facility in accordance
with the proposed amendment would not involve a significant
reduction in a margin of safety.
Based on the discussion presented above and on the supporting
Evaluation of Proposed TS Changes, FPL has concluded that this
proposed license amendment involves no significant hazards
consideration.
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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW, Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: February 18, 1994
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) by deleting
the frequencies specified for audits performed under the cognizance of
the Company Nuclear Review Board (CNRB). The periodicity of the audits
for these activities will be controlled as described in the licensee's
Topical Quality Assurance Report (FPLTQAR), wherein the minimum audit
frequency for any activity is established as biennial unless the audit
is otherwise required to be performed more frequently by the TS, Code
of Federal Regulations, or other licensing commitments. Periodic audits
of selected aspects of operational phase activities are performed with
a frequency commensurate with safety significance. During the interval
between the periodic audits, continuing performance evaluations are
conducted of activities important to plant safety.
In addition, the licensee proposes to revise the TS in accordance
with Generic Letter 93-07. Generic Letter 93-07, ``Modifications of the
Technical Specifications Administrative Control Requirements for
Emergency and Security Plans,'' issued December 28, 1993, provided
guidance for changes to the TS to remove the audit of the emergency and
security plans and implementing procedures from the list of
responsibilities of the company nuclear audit and review group. The
basis of this change is that Parts 50 and 73 of Title 10 of the Code of
Federal Regulations (10 CFR) include provisions that are sufficient to
address these 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) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments relocate the administrative control
criteria for minimum audit frequencies from the facility TS to the
FPL Quality Assurance (QA) Program. The QA Program is described in
the FPL Topical Quality Assurance Report pursuant to 10 CFR 50,
Appendix B. In addition, the proposed amendments in accordance with
Generic Letter 93-07, changes the TS to remove the audit of the
emergency and security plans and implementing procedures from the
list of responsibilities of the Company Nuclear Review Board. The
changes being proposed are administrative in nature and do not
affect assumptions contained in plant safety analyses, the physical
design and/or operation of the plant, nor do they affect the TS that
preserve safety analysis assumptions. Therefore, operation of the
facility in accordance with the proposed amendments would not affect
the probability or consequences of an accident previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The changes being proposed are administrative in nature and will
not change the physical plant or the modes of operation defined in
the Facility License. The change does not involve the addition or
modification of equipment nor does it alter the design or operation
of plant systems. Therefore, operation of the facility in accordance
with the proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The changes being proposed are administrative in nature and do
not alter the bases for assurance that safety-related activities are
performed correctly or the basis for any TS that is related to the
establishment of or maintenance of a safety margin. Therefore,
operation of the facility in accordance with the proposed amendments
would not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
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, Unit 1,
Matagorda County, Texas
Date of amendment request: March 14, 1994
Description of amendment request: The licensee proposes to make a
change to the technical specifications to add a new Limiting Condition
For Operation (LCO), 3.0.6. LCO 3.0.6 will allow equipment removed from
service or declared inoperable to comply with actions to be returned to
service, under administrative controls, solely to perform testing. The
new LCO will permit non-compliance with the applicable Action statement
to perform the post-maintenance and surveillance testing required to
demonstrate the operability of the equipment being returned to service
or the operability of other equipment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The implementation of LCO 3.0.6 will allow the orderly and
judicious return to service of inoperable equipment. This LCO will
permit equipment removed from service to comply with required actions
to be returned to service under administrative controls to verify the
component or system will perform its safety function. The
administrative controls will ensure the time involved will be limited
to only the time required to demonstrate the component or system's
operability. The implementation of this new LCO will provide an
acceptable method of testing technical specification equipment prior to
its return to operable service following required maintenance. These
actions will ensure that the equipment being returned to service is
capable of performing its designed safety function prior to being
declared operable. Therefore, this action will ensure the probability
or consequences of an accident previously evaluated are not
significantly increased.
2. The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
The equipment is only being tested in its designed configuration or
being returned to service to allow testing of another component or
system. Therefore, the use of this new LCO will not result in 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 use of the new LCO will only allow the return to service of
equipment that is expected to operate as designed. The use of the LCO
will be limited to the performance of testing on the equipment being
returned to service or on other equipment that is dependent on the
equipment being returned to service. This testing is limited to post-
maintenance testing and the testing necessary to prove operability.
Since the equipment will be controlled by administrative requirements
that will ensure all necessary actions are taken, this change does not
involve a significant reduction in a margin of safety.
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: 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, NW, Washington, DC 20036
NRC Project Director: Suzanne C. Black
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: February 22, 1994
Description of amendment requests: The proposed amendments would
modify the technical specifications to reduce surveillance requirements
for testing during power operation. This modification was proposed to
licensees in NRC 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:
Criterion 1
Although the surveillance requirements are lessened by these
proposed changes, the changes are consistent with those found
acceptable by the NRC in Generic Letter 93-05. The proposed changes
have been determined to be compatible with our plant operating
experience. Based on these considerations, it is concluded that the
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2
The proposed changes do not involve physical changes to the
plant or changes in plant operating configuration. The changes only
involve frequency of testing required to be performed. The changes
are consistent with those found to be acceptable by the NRC in
Generic Letter 93-05. Thus, it is concluded that the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3
Although the surveillance requirements are lessened by these
proposed changes, the changes are consistent with those found
acceptable by the NRC in Generic Letter 93-05. The proposed changes
have been determined to be compatible with our plant operating
experience. Based on these considerations, it is concluded that the
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Ledyard B. Marsh
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of amendment request: February 15, 1994
Description of amendment request: The proposed amendment would
delete from the Technical Specifications the operational and
surveillance requirements for the turbine overspeed protection system.
The licensee intends to continue testing of the overspeed protection
system as part of plant procedures.
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) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment does not involve a significant increase
in the probability or consequences for an accident previously
evaluated. The proposed deletion of the turbine overspeed protection
T/S [technical specification] will not significantly change the
surveillance tests on the Unit 2 turbine. The surveillance schedule
and tests will be under administrative procedures outside of the TSs
similar to that of Unit 1 and will be in line with operating
experience at Cook Nuclear Plant and applicable industry experience.
The Unit 2 turbine is now operating in its ninth operating cycle
with over 90,000 hours of operation. Turbine overspeed protection
surveillance results have been very good since unit startup in 1978.
In 1983, a wear problem was found with the overspeed plungers.
Replacement plungers were installed. Then in 1988, these plungers
were replaced with parts having stellited (hardened) surfaces. There
have been no subsequent problems. Our expectation is that the
turbine overspeed protection system will remain available to perform
its function of preventing excessive turbine overspeed. Lastly, the
STS [Standard Technical Specifications] developed by the MERITS
program in NUREG-1431 do not include a T/S for turbine overspeed
protection. The omission of an overspeed protection T/S in NUREG-
1431 indicates that a T/S is not needed to ensure an adequate level
of safety for a nuclear facility. This view is supported by WCAP
11618 which uses the NRC's ``Interim Policy Statement Criteria'' to
evaluate the need for a turbine overspeed protection T/S and
concludes that it is not needed. For these reasons, we believe that
deleting the turbine overspeed protection T/S will not significantly
increase the probability or consequences of an accident previously
evaluated.
(b) Create the possibility of a new or different kind of
accident from any previously analyzed.
The proposed amendment does not create the possibility of a new
or different kind of accident from any previously evaluated. This
request to delete the turbine overspeed protection T/S eliminates a
control on the surveillance testing of the Unit 2 turbine. The
design function of the turbine overspeed protection and the
operation of the turbine/generator remain the same. The operating
history of the Unit 2 surveillance results to date and our continued
testing support the view that the turbine overspeed protection will
remain available. For these reasons, we believe that the proposed
changes will not create the possibility of a new or different kind
of accident from any previously analyzed.
(c) Involve a significant reduction in a margin of safety.
The proposed amendment does not involve a significant reduction
in the margin of safety. Turbine overspeed protection surveillance
results have been excellent since 1983. The years of operating data
well within acceptance criteria on Unit 2 turbine overspeed
protection provide ample evidence that there is no significant
degradation of the system to perform its function. The reliability
of the overspeed protection was improved by the replacement of the
plungers with parts having stellited surfaces. The surveillance
schedule and tests will be based on operating experience at Cook
Nuclear Plant and applicable industry experience. Surveillance
testing will continue under an administrative program outside of
TSs. Thus the turbine overspeed protection is expected to remain
available. Also by eliminating this T/S we will be reducing the
potential for shutting down the unit because of difficulties
performing this T/S surveillance unrelated to the functionality of
the valves and overspeed trip protection. Lastly, the STS developed
by the MERITS program in NUREG-1431 do not include a T/S for turbine
overspeed protection. The omission of an overspeed protection T/S in
NUREG-1431 indicates that a T/S is not needed to ensure an adequate
level of safety for a nuclear facility.
This view is supported by WCAP 11618 which uses the NRC's
``Interim Policy Statement Criteria'' to evaluate the need for a
turbine overspeed protection T/S and concludes that it is not
needed. For these reasons, we believe that the turbine overspeed
protection system will remain operable and so this proposed
amendment 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: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Ledyard B. Marsh
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of amendment request: February 22, 1994
Description of amendment request: The proposed amendment would
revise the reactor coolant system heatup and cooldown curves in the
Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to the P-T [pressure-temperature] curves
are being updated as a result of the Unit 2 Capsule U analysis,
WCAP-13515. The analysis was required per the removal schedule
established in Table 4.4-5 of the Cook Nuclear Plant Technical
Specifications. The analysis was performed based on guidance from R/
G 1.99 [Regulatory Guide 1.99, ``Radiation Embrittlement of Reactor
Vessel Materials''], Revision 2. The change only involves a revised
time frame for material qualification from 12 EFPY [effective full-
power years] to 15 EFPY as supported by the aforementioned
Westinghouse analysis. Therefore, we conclude that the changes will
not involve a significant increase in the probability or
consequences of a previously evaluated accident, nor will the
changes involve a significant reduction in a margin of safety.
(2) Create the possibility of a new or different kind of
accident from any previously analyzed.
The proposed changes do not involve any physical modifications
to the plant. Therefore, the changes should not create the
possibility of a new or different kind of accident from any
previously analyzed or evaluated.
(3) Involve a significant reduction in a margin of safety.
See the response to (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: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Ledyard B. Marsh
Long Island Power Authority, Docket No. 50-322, Shoreham Nuclear
Power Station, Unit 1 (SNPS), Wading River, New York
Date of application for amendment: Amendment No. 11, November 4,
1993 (Reference LSNRC-2115)
Brief description of amendment: This license amendment request
(LSNRC-2115) proposes to delete from the Possession-Only License (POL)
the requirements associated with the safe storage and handling of
irradiated fuel, the accompanying Appendix A of SNPS Technical
Specifications, and Appendix B of SNPS Environmental Protection Plan
(non-radiological). This proposed amendment will update the SNPS POL to
reflect the status of the facility after irradiated fuel removal from
the site. SNPS License Condition No. 3 prohibits this amendment from
being implemented until all the fuel has been removed from SNPS, and
the licensee has certified to the NRC that all the fuel has been
removed.
Basis for the proposed no significant hazards consideration
determination: In accordance with the requirements and standards in 10
CFR 50.92(c), the licensee has provided an analysis of the issues
related to the no significant hazards consideration.
The licensee's analysis of the issues related to no significant
hazards consideration are presented below:
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes will become effective after the fuel and
its related hazards are removed from the site.
Therefore, the proposed changes will update the SNPS license to
reflect the facility status after the removal of irradiated fuel.
This action will not increase the probability or consequences of any
accident previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes will update the license by deleting
requirements which will no longer apply to SNPS and will not have an
adverse impact on the operation of the remaining plant systems and
components.
Therefore, the proposed change does not create the possibility
for an accident or malfunction different from any previously
analyzed.
c. Does the change involve a significant reduction in a margin
of safety?
The proposed change will update the license to reflect the
status of the facility after the removal of irradiated fuel from the
site.
Therefore, the proposed changes will not reduce the margins of
safety for the remaining plant systems and components.
The NRC staff has reviewed the licensee's analysis and based on
this review the three standards of 50.92(c) are satisfied. The NRC
staff agrees with the licensee's analysis and has determined that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Shoreham Wading River Public
Library, Shoreham Wading River High School, Route 25A, Shoreham, NY
11792
Attorney for licensee: Mr. W. Taylor Reveley, III, Hunton and
Williams, Riverfront Plaza, East Tower, 951 East Byrd Street, Richmond
VA 23219-4074
NRC Branch Chief: John H. Austin
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 amendment requests: February 14, 1994
Description of amendment requests: The proposed amendments would
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). This would make the specifications
the same for both units since the configuration for the two units will
become the same during the 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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
SBO/ESU [Station Blackout/Electrical Safeguards Upgrade] Project
modifications as reflected in the proposed Technical Specifications
changes were evaluated to determine their impact, if any, on
potential transients and accidents as described in the Prairie
Island USAR [Updated Safety Analysis Report]. Each transient and
accident was evaluated in terms of the mitigating actions described
or assumed in the USAR analysis. The role of the modified systems in
mitigating the event was analyzed in order to evaluate whether the
modification:
(1) changed, degraded or prevented actions described or assumed
in the USAR analysis;
(2) altered any assumptions made in evaluating the radiological
consequences of the accident;
(3) played a direct part in mitigating the radiological
consequences of the accident; or
(4) affected any fission product barrier.
The evaluation demonstrated that the USAR transient and accident
analyses remain valid and bounding.
As part of the evaluation, the revised emergency diesel
generator load sequence was analyzed and found to be bounded by the
existing analyses.
In particular, the USAR analyses of the loss of offsite power
(LOOP) event and the large break loss of coolant accident (LBLOCA)
remain valid and bounding. In addition, the current USAR analysis
for the radiological consequences of a LBLOCA remains valid.
Further, the plant response to a loss of AC power event is not
degraded as a result of these changes but, in fact, is significantly
improved.
In order to determine the effect of the modifications upon the
probability and consequences of an accident, the following items
were specifically evaluated:
(1) the applicable design, material and construction standards;
(2) instrumentation accuracies and response times;
(3) the equipment operating and design limits, including
electrical bus loading, emergency diesel generator loading and
battery loading;
(4) the system interfaces;
(5) voltage margins; and
(6) coordination of protective devices.
Structures, systems and components involved in the modifications
were evaluated as follows:
(1) The design specifications for the new structures, systems
and components were considered for the following requirements:
- seismic;
- separation including control/power circuit interaction,
redundancy/separation of systems, and isolation between safety and
non-safety circuits;
- environmental parameters;
- severe meteorological events;
- missiles; and
- fire protection.
All structures, systems and components meet the appropriate
design requirements for their respective classifications.
(2) Structures, systems and components were additionally
evaluated for the following:
- Structural loads were determined for new cable runs in the
existing plant and for new cable penetrations in the existing
structures.
- New electrical loads requirements were determined.
- System/equipment protection features have been maintained in
the modification.
- Support system performance was specified to maintain the
safety function of the equipment.
- System/equipment redundancy and independence is maintained.
- The frequency of operation of existing equipment was evaluated
and determined not to be affected.
- The testing requirements imposed on new structures, systems
and components are in accordance with their safety classification.
Failures of systems and components involved in the modifications
were analyzed, and it was determined that all safety functions were
maintained.
Required engineered safeguards features loads are accommodated
with the improved auxiliary electrical systems configuration; and,
as demonstrated by the performance of a failure modes and affects
analysis, no single failure will prevent the modified plant from
performing its required safety function in the event of an accident
on either unit.
For the reasons discussed above, the proposed amendment does not
significantly increase the probability or consequences of an
accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The SBO/ESU Project modifications as reflected in the proposed
Technical Specifications changes were evaluated to determine if they
could create the possibility of a new or different kind of accident
from any accident previously evaluated.
The modifications were evaluated to determine the types of
accidents which could result from malfunction of the new/modified
structures, systems and components. It was determined that no new or
different kinds of accidents from those previously evaluated are
created. USAR analyses remain bounding.
For these reasons, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The new Unit 1 480V safeguards configuration provides additional
circuit breakers for improved motor control center (MCC) feeder
circuit coordination by eliminating subfed 480V MCCs from safeguards
480V buses. The proposed Technical Specification changes identify
the new 480V buses and require the operability of both of the buses
per train rather than the one bus per train of the current
configuration and current Technical Specification requirements.
Since the operability requirements are not decreased nor are the
allowed out-of-service times increased by the proposed changes, the
margin of safety is maintained.
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 requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Ledyard B. Marsh
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: February 16, 1994 (Reference LAR 94-03)
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to revise TS 4.6.1.2, ``Containment
Integrity.'' The specific TS changes proposed are as follows:
(1) The requirement to conduct three Type A tests specifically at
40 plus or minus 10 month intervals during each 10-year service period
would be replaced with a requirement to conduct three Type A tests at
approximately equal intervals during each 10-year service period.
(2) The requirement to conduct the third Type A test of each set
during the shutdown for the 10-year plant inservice inspection would be
deleted.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes do not affect the initiation of any
accident, nor do the proposed changes involve modifications to any
plant equipment.
The proposed change to the schedule provides flexibility in
meeting the current requirement for 3 tests in 10 years and is
consistent with the intent of the 10 CFR 50, Appendix J requirement
to perform Type A tests at approximately equal intervals. The test
type and test method used for Type A testing would not be changed.
The Type A test acceptance criteria would not be changed, and
containment leakage will continue to be maintained within the
required limits.
Elimination of the requirement to perform the third Type A test
during the shutdown for the 10-year plant ISI does not involve any
modification to plant equipment or affect the operation or design
basis of the containment. These surveillances are independent of
each other and provide assurance of different plant characteristics.
The Type A tests assure the required leak-tightness of the
containment to demonstrate compliance with the guidelines of 10 CFR
100. The 10-year ISI program provides assurance of the integrity of
plant structures, systems, and components and verifies the
operational readiness of pumps and valves in accordance with 10 CFR
50.55a.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not involve modifications to any
existing equipment or affect the operation or design basis of the
containment. The proposed changes do not affect the response of the
containment during a design basis accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
c. Does the change involve a significant reduction in a margin
of safety?
The proposed changes to the schedule provide flexibility in
meeting the Type A testing schedule requirements. These proposed
changes do not affect or change any limiting conditions for
operation (LCO) or any other surveillance requirements in the TS and
the Bases for the surveillance requirement remains unchanged. The
testing method, acceptance criteria, and bases are not changed and
still provide assurance that the containment will perform its
intended function.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: Theodore R. Quay
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: January 10, 1994
Description of amendment request: The proposed amendment would
relocate the seismic monitoring instrumentation Limiting Condition for
Operation, Surveillance Requirements, and associated tables and Bases
contained in TS sections 3.3.7.2 and 4.3.7.2 to the Updated Final
Safety Analysis Report (UFSAR).
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) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The function of the seismic monitoring instrumentation system is
to monitor the magnitude and effect of a seismic event only, and
cannot initiate or mitigate an accident previously evaluated.
Furthermore, the proposed TS changes to relocate the seismic
monitoring instrumentation requirements from TS to the UFSAR are in
accordance with the criteria for determining those requirements that
should remain in the TS as defined by the NRC in its final policy
statement, ``Final Policy Statement on Technical Specifications
Improvements for Nuclear Power Reactors,'' dated July 22, 1993. The
seismic monitoring instrumentation LCO, SRs, and associated tables
and Bases proposed for relocation from TS to the LGS UFSAR will
continue to be implemented by administrative controls that will
satisfy the applicable requirements of TS section 6 ``Administrative
Controls.'' Those requirements include a review of changes to plant
systems and equipment and to the applicable administrative controls
in accordance with the provisions of 10CFR50.59.
Criterion 2 of the July 22, 1993 NRC final policy statement
states, ``A process variable, design feature, or operating
restriction that is an initial condition of a Design Basis Accident
or Transient analysis that either assumes the failure of or presents
a challenge to the integrity of a fission product barrier.'' The
seismic monitoring instrumentation system is not a system that
monitors a process variable that is an initial condition for
accident or transient analyses. The seismic monitoring
instrumentation is also not a design feature or an operating
restriction that is an initial condition of Design Basis Accident or
transient analyses since it only provides information regarding the
magnitude of and the plant equipment response to a Design Basis
earthquake. Therefore, the current LGS seismic monitoring
instrumentation TS requirements do not meet Criterion 2 of the July
22, 1993 NRC final policy statement.
Criterion 3 of the July 22, 1993 NRC final policy statement
states, ``A structure, system, or component that is part of the
primary success path and which functions or actuates to mitigate a
Design Basis Accident or Transient that either assumes the failure
of or presents a challenge to the integrity of a fission product
barrier.'' The LGS seismic monitoring instrumentation system does
not provide a function or actuate in order to mitigate the
consequences of a Design Basis Accident or transient. Therefore, the
current LGS seismic monitoring instrumentation TS requirements do
not meet Criterion 3 of the July 22, 1993 NRC final policy
statement.
Criterion 4 of the July 22, 1993 NRC final policy statement
states, ``A structure, system or component which operating
experience or probabilistic safety assessment has shown to be
significant to public health and safety.'' Operating experience has
shown that the LGS seismic monitoring instrumentation system has no
impact on public health and safety as defined by the NRC final
policy statement. Furthermore, LGS specific probabilistic risk
assessment (PRA) does not credit the seismic monitoring
instrumentation system as a significant factor in the plant response
to an accident. Therefore, the current LGS seismic monitoring
instrumentation TS requirements do not meet Criterion 4 of the July
22, 1993 NRC final policy statement for determining those
requirements that should remain in TS. This conclusion is consistent
with the function of the seismic monitoring instrumentation system
stated above.
These proposed TS changes will maintain the current operation,
maintenance, testing, and system operability controls of the seismic
monitoring instrumentation system. Furthermore, any future changes
to the seismic monitoring instrumentation system will be evaluated
for the effect of those changes on system reliability as required by
10CFR50.59. The seismic monitoring instrumentation system
performance will not decrease due to these proposed TS changes and
the system will continue to be administratively controlled in
accordance with TS Section 6, including the requirements of
10CFR50.59, thereby precluding a future decrease in its performance.
In accordance with the current TS Section 3.3.7.2, with the
seismic monitoring instrumentation inoperable, the plant would not
be required to shutdown and the provisions of TS Section 3.0.3
(i.e., plant shutdown) would not be applicable. Therefore, the
inoperability of this system and therefore the consequences of an
accident while this system is inoperable, was previously evaluated
as not significant enough to require a change to the plant operating
condition.
Since the seismic monitoring instrumentation system does not
monitor a process variable that is an initial condition for an
accident or transient analyses, or actuates any accident mitigation
feature, and since the operation, maintenance, testing, and
modification of the seismic monitoring instrumentation system will
continue to be administratively controlled, including the
requirements of 10CFR50.59; therefore, maintaining the reliability
of the system, the proposed TS changes will not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The function of the seismic monitoring instrumentation system is
to monitor the magnitude and effect of a seismic event only. The
proposed TS changes to relocate the seismic monitoring instruments
requirements from TS to the UFSAR are in accordance with the
criteria for determining those requirements that should remain in
the TS as defined by the NRC in its final policy statement, dated
July 22, 1993. The seismic monitoring instrumentation system does
not monitor a process variable that is an initial condition for an
accident or transient analyses. The seismic monitoring
instrumentation is also not a design feature or an operating
restriction that is an initial condition of a Design Basis Accident
or transient analyses since it only provides information regarding
the magnitude of and the plant equipment response to a Design Basis
earthquake.
These proposed TS changes to relocate the TS requirements to the
UFSAR will not alter the operation of the plant, or the manner in
which the seismic monitoring instrumentation system will perform its
function, and any future changes will continue to be
administratively controlled in accordance with TS Section 6,
including the requirements of 10CFR50.59.
These proposed TS changes will not impose new conditions nor
result in new types of equipment which will result in different
types of malfunctions of equipment important to safety than any type
previously evaluated.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
These proposed TS changes to relocate the seismic monitoring
instrumentation requirements from TS to the UFSAR are in accordance
with the criteria for determining those requirements that should
remain in the TS as defined by the NRC in final policy statement,
dated July 22, 1993.
Criterion 1 of the NRC final policy statement states,
``Installed instrumentation that is used to detect, and indicate in
the control room, a significant abnormal degradation of the reactor
coolant pressure boundary.'' The NRC final policy statement explains
that ''...This criterion is intended to ensure that Technical
Specifications control those instruments specifically installed to
detect excessive reactor coolant leakage. This criterion should not,
however, be interpreted to include instrumentation to detect
precursors to reactor coolant pressure boundary leakage or
instrumentation to identify the source of actual leakage (e.g.,
loose parts monitor, seismic instrumentation, valve position
indicators).'' Based on the above NRC guidance, the LGS UFSAR, and
TS Bases 3.3.7.2, the seismic monitoring instrumentation does not
detect, and indicate in the control room, a significant abnormal
degradation of the reactor coolant pressure boundary. Therefore, the
current LGS seismic monitoring instrumentation TS requirements do
not meet Criterion 1. Furthermore, operating experience has shown
that the LGS seismic instrumentation system has no impact on public
health and safety as defined by the NRC final policy statement. In
addition, the LGS specific PRA does not credit the seismic
monitoring instrumentation system as a significant factor in the
plant response to accidents.
The seismic monitoring instrumentation LCO, SRs, and associated
tables and Bases proposed for relocation to the LGS UFSAR will
continue to be implemented by administrative controls that will
satisfy the applicable requirements of TS section 6 ``Administrative
Controls.'' Those requirements include a review of future changes to
the system and applicable administrative controls in accordance with
the provisions of 10CFR50.59.
Accordingly, based on the above discussion of NRC specific
guidance, operating experience, and continued imposition of
administrative controls, the proposed TS changes do 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
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 amendment request: February 3, 1994
Description of amendment request: The licensee commenced operating
on a 24-month fuel cycle, instead of the previous 18-month fuel cycle,
with fuel cycle 9. Fuel cycle 9 started in August 1992; however, the
facility shut down in February 1993 for a ``Performance Improvement
Plan'' outage and a restart date has not yet been established. In order
to accommodate operation on a 24-month cycle after the facility
restarts, the licensee requested an amendment to the Technical
Specifications (TSs) to incorporate the changes listed in items 1-7
below:
(1) The licensee proposed changing the calibration frequency for
the reactor coolant temperature instrument channels (specified in TS
Table 4.1-1) to accommodate operation on a 24-month cycle.
(2) The licensee proposed changing the calibration frequency for
the steam generator level instrument channels (specified in TS Table
4.1-1) to accommodate operation on a 24-month cycle.
(3) The licensee proposed changing the calibration frequency for
the containment pressure instrument channels (specified in TS Table
4.1-1) to accommodate operation on a 24-month cycle.
(4) The licensee proposed changing the calibration frequency for
the steam line pressure instrument channels (specified in TS Table 4.1-
1) to accommodate operation on a 24-month cycle.
(5) The licensee proposed changing the calibration frequency for
the turbine first stage pressure instrument channels (specified in TS
Table 4.1-1) to accommodate operation on a 24-month cycle.
(6) The licensee proposed changing the calibration frequency for
the turbine trip low auto stop oil pressure instrument channels
(specified in TS Table 4.1-1) to accommodate operation on a 24-month
cycle.
(7) The licensee proposed changing the calibration frequency for
the 480V bus undervoltage and alarm relays (specified in TS Table 4.1-
1) to accommodate operation on a 24-month cycle.
These proposed changes follow the guidance provided in Generic
Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle,'' as applicable.
The licensee also requested the following additional changes:
(1) The addition to TS Table 3.5-5 of limiting conditions for
operation (LCO) requirements for a wide range containment pressure
variable.
(2) The addition of a quarterly functional test surveillance
requirement to Item 4 of TS Table 4.1-1 for the low average temperature
actuation circuits of the reactor coolant temperature channels.
(3) The addition of a second line to Item 14 of TS Table 4.1-1 to
specify surveillance requirements for the wide range containment
pressure instrumentation.
(4) The revision of Item 20 to TS Table 4.1-1 to clarify that both
the reactor trip and the engineered safety features (ESF) actuation
relay logic channels are functionally tested.
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:
Consistent with the criteria of 10 CFR 50.92, the enclosed
application is judged to involve no signicant hazards based on the
following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of any accident
previously evaluated?
Response:
The proposed changes do not involve a significant increase in
the probability or consequences of any accident previously
evaluated. The proposed changes extend the calibration intervals
(given in [TS] Table 4.1-1) for the reactor coolant loop temperature
instrumentation used for Engineered Safety Features Actuation
Systems (ESFAS) and Post Accident Monitoring (PAM) functions, the
steam generator (SG) level instrumentation used for ESFAS and PAM
functions, the containment pressure instrumentation used for ESFAS
and PAM functions, the steam line pressure instrumentation used for
ESFAS and PAM functions, the turbine first stage pressure
instrumentation used for ESFAS functions, the 480V bus undervoltage
and alarm relays used for ESFAS functions, and the turbine trip low
auto stop oil pressure instrumentation. These changes are being made
to accommodate a 24 month operating cycle. Other changes include: 1)
the addition to [TS] Table 3.5-5 of limiting conditions for
operation (LCO) requirements for the wide range containment pressure
channels; 2) the addition of a quarterly functional test
surveillance requirement to Item 4 of [TS] Table 4.1-1 for the low
Tavg [average temperature] actuation circuits of the reactor
coolant temperature channels; 3) the addition of a second line to
Item 14 of [TS] Table 4.1-1 to specify the surveillance requirements
for the wide range containment pressure channels; and 4) the
revision of Item 20 to [TS] Table 4.1-1.
Extension of the calibration intervals in question were
evaluated and the results documented in the ESFAS and Indicating
Instrument Surveillance Test Extension reports (References 7 and 8
[Engineered Safety Features Actuation Systems Surveillance Test
Extensions, NYPA document IP3-RPT-ESS-00400, dated May 10, 1993 and
Indicating Instruments Surveillance Test Extensions, NYPA document
IP3-RPT-MULTI-00424, dated May 5, 1993]). ESFAS and indicating
instrument drift analyses were performed to evaluate actual past and
projected future instrument drift. Revised safety system loop
accuracy/setpoint calculations, which include any additional
instrument uncertainties resulting from the proposed calibration
interval extensions, show that sufficient margin exists between the
analytical and field trip settings for the low Tavg, the SG
low-low level, the high and high-high containment pressure, the high
differential steam line pressure, the low steam line pressure, the
high steam flow (dependent upon turbine first stage pressure), the
turbine trip low auto stop oil pressure, and the 480V bus
undervoltage trip functions. Safety analyses are not affected.
Additionally, postulated uncertainties associated with the extended
calibration intervals for the wide range reactor coolant loop
temperature, the narrow and wide range SG level, and the steam line
pressure instrumentation will be accomodated by changes to the
Emergency Operating Procedure (EOP) settings. Extension of the
calibration interval for the narrow range containment pressure
instrumentation channels does not affect EOP settings. Safety
analyses are not affected by the EOP setting changes.
The results of the changes to: 1) add to [TS] Table 3.5-5 LCO
requirements for the wide range containment pressure channels, 2)
add a quarterly functional test surveillance requirement to Item 4
of [TS] Table 4.1-1 for the low Tavg actuation circuits of the
reactor coolant temperature channels, and 3) add a monthly channel
check surveillance requirement to [TS] Table 4.1-1 for the wide
range containment pressure channels are consistent with Westinghouse
Standard Technical Specifications (W STS - Reference 12 [NUREG-1431,
Revision O, ``Standard Technical Specifications - Westinghouse
Plant,'' dated September 28, 1992]). The addition of LCO
requirements to [TS] Table 3.5-5 for the wide range containment
pressure instrumentation, the addition of a quarterly functional
test requirement to Item 4 of [TS] Table 4.1-1 for the low Tavg
actuation circuits, and the separation of surveillance requirements
for the narrow and wide range containment pressure instrumentation
into two lines on [TS] Table 4.1-1 consitute additional technical
specification controls. Changes which consitute additional technical
specification limitations and controls are classified by Federal
Register dated April 6, 1983 (48 FR 14870, April 6, 1983) as not
likely to involve significant hazards considerations. The change to
[TS] Table 3.5-5 ensures conistency with the Authority's commitment
to Regulatory Guide (RG) 1.97 [``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions
During and Following an Accident''] for the containment pressure
variable.
The current surveillance requirement specified by Item 20 has
been interpreted by Indian Point 3 as including on-line testing of
both the reactor trip and engineered safety features (ESF) actuation
logic channels, but since the wording may be confusing, this
application proposes to change the wording to clarify that both the
reactor trip and the ESF actuation logic channels are functionally
tested at least every two months on a staggered basis (i.e., one
train per month). The change is consistent with W STS and only
involves a wording change which strengthens the Technical
Specification requirement. The change does not involve hardware,
procedural, or operational changes, and, therefore, does not affect
safety analyses.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any previously
evaluated?
Response:
The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated. Extension
of the calibration intervals in question were evaluated and the
results documented in the ESFAS and Indication Instrument
Surveillance Test Extension reports. ESFAS and indicating instrument
drift analyses were performed to evaluate actual past and projected
future instrument drift. Revised safety system loop accuracy/
setpoint calculations and EOP setting calculations show that,
although some EOP setting changes will be made to accommodate
postulated drift associated with the extended calibration intervals,
safety analyses are not affected.
The changes to 1) specify LCO and surveillance requirements for
the wide range containment pressure instrumentation channels, 2) add
a quarterly functional test surveillance requirement to Item 4 of
[TS] Table 4.1-1 for the low Tavg actuation circuits of the
reactor coolant temperature channels, and 3) clarify that both
reactor trip and ESF actuation logic channels are functionally
tested constitute additional technical specification limitation and
controls. Additionally, these changes are consistent with W STS.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed changes do not involve significant reductions in
margins of safety. Loop accuracy/setpoint calculations show that
sufficient margin exists between the analytical and field trip
settings for the low Tavg, the SG low-low level, the high and
high-high containment pressure, the high differential steam line
pressure, the low steam line pressure, the high steam flow
(dependent upon turbine first stage pressure), the turbine trip low
auto stop oil pressure, and the 480V bus undervoltage trip functions
to accommodate postulated uncertainties associated with the extended
calibration intervals. And, although changes to EOP settings will be
made to accommodate the postulated uncertainties associated with the
extended calibration intervals for the wide range reactor coolant
loop temperature, the narrow and wide range SG level, and the steam
line pressure instrumentation, the EOP setting changes do not in any
way adversely affect the analytical limits established by safety
analyses.
Extension of the calibration intervals in question do not affect
safety analyses. The other changes being made in this application
involve additional technical specification limitations and controls
and are consistent with W STS. None of the changes involve
significant reductions 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 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: Robert A. Capra
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 4, 1994
Description of amendment request: The proposed change would modify
Sections 5.3 and 5.6 of the Technical Specifications (TSs) to allow the
use of Westinghouse Vantage+ fuel with ZIRLO cladding. The present TSs
require the fuel rod cladding to be Zircaloy-4, which is used in the
Westinghouse Standard and Vantage 5H fuel designs.
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:
The proposed changes to Technical Specifications 5.3.1 and 5.6.1
for Salem Generating Station (SGS) Unit Nos. 1 and 2:
1. do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The fuel cladding design criteria for SGS would remain the same
for ZIRLO clad fuel as it is for Zircaloy-4 clad fuel. All fuel
design and performance criteria will continue to be met using NRC-
approved methods and no new single failure mechanisms will be
introduced. The use of ZIRLO clad fuel does not introduce any
changes to plant equipment or operation that would adversely affect
accident initiators or precursors. The proposed changes would not
result in any changes to compliance with licensing basis safety
limits.
2. do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed changes would require that NRC[-]approved methods
be used in fuel assembly design. No new operating configurations
potentially resulting in the occurrence of a previously unanalyzed
event would be allowed by the proposed change.
3. do not involve a significant reduction in a margin of safety.
The proposed change would continue to require that NRC[-
]approved methods are used to ensure compliance with the fuel design
and safety limits which ensure that an acceptable margin of safety
is maintained relative to fuel assembly design.
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
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 amendment request: March 1, 1993
Description of amendment request: The proposed amendment would
clarify Technical Specification 3.6.1.2, Primary Containment Leakage,
and revise the ``as-found'' value of the overall integrated primary
containment leakage rate which is used when determining the test
schedule for future Type A tests within Surveillance Requirement
4.6.1.2.b. This amendment also requests an exemption from the
requirements of 10 CFR 50 Appendix J, Primary Reactor Containment
Leakage Testing for Water-Cooled Power Reactors.
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.
These proposed changes clarify Technical Specification 3.6.1.2
by providing a more definitive action to take if the leakage rate
limit(s) specified in the LCO are not being met. The current Action
is not clear on what actions are necessary if the leakage rate
limits (e.g., Type B and C limits) are known to be exceeded while
the reactor coolant system (RCS) temperature is above 200 deg. F,
which has caused compliance difficulties. The revised Action is
modeled after the one in the Primary Containment Integrity
Specification, which (through the definition of Primary Containment
Integrity) includes a provision that the containment leakage rates
be in compliance with the requirements of Specification 3.6.1.2.
Surveillance Requirement 4.6.1.2.b has been revised to reflect
the actual plant design basis leakage rate of La as the value
against which the ``as-found'' Type A test results are compared when
determining the test schedule for future Type A tests. The
probability of exceeding the maximum allowable leakage rate,
La, is not significantly increased since the ``as-left''
leakage rate requirement of 0.75 La (which must be met during
startup from any outage in which a Type A test has been performed)
is still imposed through LCO 3.6.1.2.a, Action 3.6.1.2.a and
Surveillance Requirement 4.6.1.2.a. The Applicability of
Specification 3.6.1.2 has been modified to resolve an existing
conflict with the current Action, which requires that a reactor
coolant system temperature of 200 deg. F not be exceeded with a
leakage rate greater than 0.75 La (during startups from outages
in which a Type A ILRT has been performed). With the modified
Applicability and the retained LCO requirement for the ``as-left''
leakage rate to be less than or equal to 0.75 La, the
requirement of the current Action (not to exceed to 200 deg. F) is
implicitly maintained, due to the provisions contained within
specification 3.0.4. This maintains the same margin for degradation
between performances of the periodic Type A tests as is provided in
the current specification. Since the analysis leakage limit of
La has not changed, the offsite radiological consequences of an
accident assumed in the safety analyses have not been affected.
The deletion of the current link between Specifications 3.6.1.2
and 3.10.1 is an administrative change only, made because the two
Specifications no longer overlap and the link is therefore
unnecessary.
In summary, there is no change in the probability or
consequences of any accident since the clarifications of the
existing LCO, Applicability, Actions, Surveillance Requirements and
the revised ``as-found'' acceptance criterion do not change the
design of the plant, nor the operational characteristics of any
plant system, nor the procedures by which the Operators run the
plant.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed Action to address situations when the leakage rate
limit(s) cannot be met in Operational Conditions 1, 2 and 3, with
the reactor coolant system temperature greater than 200 deg. F, does
not create the possibility of a new or different kind of event - it
only provides the measures to be taken following determination of
increased containment leakage. The clarification to the existing
Applicability simply resolves an existing conflict between the
Applicability and the Action, and ensures that the same requirements
that were contained within the former Action are maintained
following implementation of the change, by preventing plant startup
above a RCS temperature of 200 deg. F (following an outage in which
a Type A test has been performed), unless the leakage rate is below
the 0.75 La test acceptance criterion. Additional changes are
being made to clarify the application of Appendix J requirements.
Revising the ``as-found'' value of La does not create the
possibility of a new or different kind of event - since the analysis
limit value, La, has not been increased and no new mode of
operation has been introduced.
In summary, the proposed changes do not create the possibility
of a new or different kind of accident, since no design changes are
being made that would create a new type of accident or malfunction,
and the method and manner of plant operation remains unchanged.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed Action simply imposes a more definitive action to
take when a leakage rate limit(s) is exceeded, consistent with the
Primary Containment Integrity Specification. The changes to the
Surveillance Requirements to reflect the ``as-found'' value of
La are consistent with the intent of the requirements specified
in Appendix J, and similar requirements have been provided for other
plants. The current requirement for ``as-left'' leakage rates to be
less than or equal to 0.75 La before increasing the reactor
coolant system temperature above 200 deg. F from outages in which a
Type A ILRT has been performed has been retained since the proposed
Action now includes a shutdown requirement, and in accordance with
Technical Specification 3.0.4, ``Entry into an OPERATIONAL CONDITION
or other specified condition shall not be made when the conditions
for the LCO are not met and the associated ACTION requires a
shutdown if they are not met within a specified time interval.''
Since the new Action includes a shutdown provision and the LCO
retains the current limit of 0.75 La, a change into the new
Applicability of Specification 3.6.1.2 cannot occur if 0.75 La
is exceeded. This ensures that the same margin as currently exists
today is maintained for possible degradation between performance of
the periodic Type A tests. The other changes are clarifications and
are administrative in nature. Therefore, the proposed changes do not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037
NRC Project Director: John N. Hannon
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: February 10, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification Table 2.2-1 and Bases Section 2.2.1. The
Functional Unit 14 of Table 2.2-1 would be revised to correct the Total
Allowance, reflecting the undervoltage relay span and to correct the
Allowable Value, reflecting the rack measurement and test equipment
(M&TE) uncertainty. The Bases would be revised to clarify the
relationship between the Trip Setpoint and Allowable Value, expressed
in voltage, and the Total Allowance, Z and S values, expressed in
percent of the undervoltage relay span (calibrated span of 70-100
volts).
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:
The proposed changes do not involve a significant hazards
consideration because operation of the Callaway Plant with these
changes would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Overall protection system performance will remain within the
bounds of the accident analyses documented in FSAR Chapter 15, WCAP-
10961-P, and WCAP-11883 since no hardware changes are proposed.
The RCP undervoltage reactor trip function is a primary trip
function and is credited in FSAR Section 15.3.2, Complete Loss of
Forced Reactor Coolant Flow. The trip setpoint is designed to ensure
plant operation within the DNB design basis. There will be no effect
on this analysis, or any other accident since the safety analysis
limit and trip response time are unaffected and remain the same as
discussed in FSAR Section 15.0.6 and FSAR Table 15.0.4.
The RCP undervoltage reactor trip will continue to function in a
manner consistent with the above analysis assumptions and the plant
design basis. As such, there will be no degradation in the
performance of nor an increase in the number of challenges to
equipment assumed to function during an accident situation.
These Technical Specification revisions do not involve any
hardware changes nor do they affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, ESF actuation setpoints, accident mitigation
capabilities, accident analysis assumptions or inputs. Therefore,
these changes will not increase the probability of an accident
previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any previously evaluated.
As discussed above, there are no hardware changes associated
with these Technical Specification revisions nor are there any
changes in the method by which any safety-related plant system
performs its safety function.
Changes to the Total Allowance and Allowable Value terms in
Technical Specification Table 2.2-1 will require only minor changes
to the acceptance criteria sections of a few surveillance
procedures. The normal manner of plant operation is unaffected. If
an undervoltage relay setpoint is found to be below the nominal trip
setpoint in Table 2.2-1, entry into Action Statements a or b of
Specification 2.2.1 will be affected insofar as the Allowable Value
is being lowered and the Total Allowance value contained in Equation
2.2-1 is being raised. However, the nominal trip setpoint is
unchanged and the required plant condition for exiting the Action
Statements, i.e. adjusting the trip setpoint consistent with the
Table 2.2-1 value, is likewise unchanged. The revisions to the Total
Allowance and Allowable Value correct errors in their derivation and
were calculated using the previously approved Westinghouse setpoint
methodology. The setpoint equations cited in that methodology are
unchanged; however, inputs to those equations have been revised to
reflect the undervoltage relay span and the rack M&TE uncertainty.
No new accident scenarios, transient precursors, failure
mechanism, or limiting single failures are introduced as a result of
these changes. There will be no adverse effect or challenges imposed
on any safety-related system as a result of these changes.
Therefore, the possibility of a new or different type of accident is
not created.
(3) Involve a significant reduction in a margin of safety.
There will be no change to the DNBR Correlation Limit, the
design DNBR limits, or the safety analysis DNBR limits discussed in
Bases Section 2.1.1.
As discussed previously, the response time of the RCP
undervoltage reactor trip function will remain within the
assumptions used in the accident analyses. The analysis of the
complete loss of flow accident will remain as presented in FSAR
Section 15.3.2.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on DNBR limits,
FQ, F-delta-H, LOCA PCT, peak local power density, or any other
margin of safety. The safety analysis limit, 9384 Vac at the RCP
motor, and the nominal trip setpoint, 10,584 Vac, remain the same as
before.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Project Director: John N. Hannon
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: February 17, 1994
Description of amendment request: The proposed amendment would
revise Technical Specifications 3/4.5.1 and Bases Section 3/4.5.1. A
new Action Statement a. would be added to Specification 3.5.1 to
provide a 72 hour allowed outage time (AOT) for one accumulator
inoperable due to its boron concentration not meeting the 2300-2500 ppm
band. The AOT for Action Statement b. would be changed to 24 hours in
lieu of the current AOT of 1 hour. Surveillances 4.5.1.1.a.1) and
4.5.1.1.b would be revised and Surveillance 4.5.1.2 would be deleted
per the guidance of NRC Generic Letter 93-05. Bases Section 3/4.5.1
would be revised to discuss the 72 hour and 24 hour AOTs for Action
Statements a. and b. above.
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:
The proposed changes to the Technical Specifications do not
involve a significant hazards consideration because operation of
Callaway Plant in accordance with these changes would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Overall protection system performance will remain within the
bounds of the accident analyses documented in FSAR Chapter 15, WCAP-
10961-P, and WCAP-11883 since no hardware changes are proposed.
The safety injection (SI) accumulators are credited in FSAR
Section 15.6.5 for large and small break LOCA. There will be no
effect on these analyses, or any other accident analysis, since the
analysis assumptions are unaffected and remain the same as discussed
in FSAR Section 15.6.5. Design basis accidents are not assumed to
occur during allowed outage times covered by the Technical
Specifications. As such, the ECCS Evaluation Model equipment
availability assumptions made in FSAR Section 15.6.5 remain valid.
The SI accumulators will continue to function in a manner
consistent with the above analysis assumptions and the plant design
basis. As such, there will be no degradation in the performance of
nor an increase in the number of challenges to equipment assumed to
function during an accident situation.
These Technical Specification revisions do not involve any
hardware changes nor do they affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, ESF actuation setpoints, accident mitigation
capabilities, accident analysis assumptions or inputs. The effect on
the Callaway core damage frequency has been quantified as
insignificant. Therefore, these 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 kind of
accident from any previously evaluated.
As discussed above, there are no hardware changes associated
with these Technical Specification revisions nor are there any
changes in the method by which any safety-related plant system
performs its safety function. The normal manner of plant operation
is unaffected.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the possibility of a new or different type of accident is
not created.
(3) Involve a significant reduction in a margin of safety.
There will be no change to the DNBR Correlation Limit, the
design DNBR limits, or the safety analysis DNBR limits discussed in
Bases Section 2.1.1.
As discussed previously, the performance of the SI accumulators
will remain within the assumptions used in the large and small break
LOCA analyses, as presented in FSAR Section 15.6.5.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on DNBR limits,
FQ, F-delta-H, LOCA PCT, peak local power density, or any other
margin of safety.
Based upon the preceding information, it has been determined
that the proposed changes to the Technical Specifications do not
involve a significant increase in the probability or consequences of
an accident previously evaluated, create the possibility of a new or
different kind of accident from any accident previously evaluated,
or involve a significant reduction in a margin of safety. Therefore,
it is concluded that the proposed changes meet the requirements of
10CFR50.92(c) and do not involve a significant hazards
consideration.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Project Director: John N. Hannon
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: March 1, 1994
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2). Specifically, the change would
eliminate certain surveillance requirements for the emergency diesel
generators which have been determined to be unnecessary.
The NRC has completed a comprehensive examination of surveillance
requirements in TS that require testing at power. The evaluation is
documented in NUREG-1366, ``Improvements to Technical Specification
Surveillance Requirements,'' dated December 1992. The NRC staff found,
that while the majority of testing at power is important, safety can be
improved, equipment degradation decreased, and an unnecessary burden on
personnel resources eliminated by reducing the amount of testing at
power that is required by TS. Based on the results of the evaluations
documented in NUREG-1366, the NRC issued Generic Letter 93-05, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation,'' dated September 27,
1993.
The safety function of the Emergency Diesel Generators (EDGs) is to
supply AC electrical power to plant safety systems whenever the
preferred AC power supply is unavailable. Consistent with Generic
Letter 93-05, Item 10.1 and NUREG-1366, the licensee is requesting a
change to the testing requirements of an operable EDG when the
alternate safety buses' EDG is inoperable or an offsite circuit is
inoperable, the separation of the hot restart test of an EDG from the
24 hour loaded run, and the elimination of fast loading of EDGs except
for the 18 month surveillance test of the Loss of Offsite Power (LOOP)
capability.
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:
Specifically, operation of North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
(1) Involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated.
Modifying the operability testing requirements for an inoperable
EDG or inoperable offsite AC source(s), gradual loading of EDGs
during surveillance testing, and separating the hot restart test of
an EDG from the 24 hour load run test of EDGs does not affect the
probability of occurrence or consequences of any previously
evaluated accidents. Surveillance testing of the EDG in accordance
with Revision 2 of Regulatory Guide 1.9 (December 1979) will
continue to ensure that the EDGs will be capable of performing their
intended safety functions. Therefore, modifying the operability
testing requirements for an inoperable EDG or inoperable offsite AC
source(s), gradual loading of EDGs during surveillance testing, and
separating the hot restart test of an EDG from the 24 hour load run
test of EDGs does not affect the probability or consequences of any
previously analyzed accident.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
Modifying the operability testing requirements for an inoperable
EDG or inoperable offsite AC source(s), gradual loading of EDGs
during surveillance testing, and separating the hot restart test of
an EDG from the 24 hours load run test of EDGs does not involve any
physical modifications of the plant or result in a change in a
method of operation. Surveillance testing of the EDG in accordance
with Revision 2 of Regulatory Guide 1.9 (December 1979) will
continue to ensure that the EDGs will be capable of performing their
intended safety functions. Therefore, a new or different type of
accident is not made possible.
(3) Involve a significant reduction in a margin of safety.
Modifying the operability testing requirements for an inoperable
EDG or inoperable offsite AC source(s), gradual loading of EDGs
during surveillance testing, and separating the hot restart test of
an EDG from the 24 hour load run test of EDGs does not affect any
safety limits or limiting safety systems settings. System operating
parameters are unaffected. The availability of equipment required to
mitigate or assess the consequences of an accident is not reduced.
Surveillance testing of the EDG in accordance with Revision 2 of
Regulatory Guide 1.9 (December 1979) will continue to assure that
the EDGs will be capable of performing their intended safety
functions. Safety margins are, therefore, not decreased.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Herbert N. Berkow
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: May 10, 1993
Description of amendment request: The amendment proposes to modify
the Technical Specifications (TS) to incorporate new power to flow
limits based on core power stability calculations performed for Cycle
9. In addition, the proposed amendment would clarify the maximum
measured decay ration permitted during 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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's evaluation of
the licensee's analysis is presented below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change related to the instability regions on the power
to flow map is based on new calculations using a new code (STAIF) while
maintaining a decay ration of 0.9 or less as required in IEB 88-07. The
closer the operators come to a decay ration of 1.0, the closer the core
comes to potential core power instabilities. By ensuring that the decay
ratio is maintained below 0.9, the operators reduce the likelihood of
core power instabilities. The result of the revised calculations using
this new code is that the restricted regions are expanded over the
regions contained in the current TS. This increase in restricted
regions results in plant operation further from potential core power
instabilities compared to the restricted regions in the current TS,
resulting in a decreased probability of core power oscillations. The
power to flow map regions are operating restrictions that, for the core
power oscillation restricted regions, are intended to reduce the
likelihood of the onset of oscillations. The core power oscillation
restricted regions on the power to flow map do not contribute to any
mitigative actions or plant response after a power oscillation occurs,
thus the proposed change does not change the consequences of any
accidents previously evaluated.
The proposed amendment would also change the wording of the
technical specifications to clarify that action must be taken to reduce
the measured decay ration if any two neutron signals of ``greater than
or equal to 0.75,'' as opposed to the current wording ``greater than
.75,'' are measured. This would not have any measurable effect on the
implementation of the affected TS, and would, if anything, result in
action being taken at a maximum lower value than the current TS. This
proposed amendment would not, therefore, involve a change in the
probability or consequences of an accident previously evaluated.
2. Does the amendment create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change; modifies existing restrictions on the power to
flow map, and does not involve any modifications to plant systems or
components or the manner in which they are operated.
Changing the wording of the TS to require that action be taken to
reduce the measured decay ration if any two neutron signals of
``greater than or equal to 0.75,'' as opposed to the current wording
``greater than .75,'' are measured, does not involve any modifications
to plant systems or components or the manner in which they are
operated.
Based on these considerations, this does not create or increase the
possibility of a new or different kind of accident.
3. Does the amendment involve a significant reduction in a margin
of safety?
The margin of safety related to the proposed TS change is the core
power vs. core flow restrictions on the power to flow map. These
restrictions are currently based on maintaining a decay ratio less than
0.9, which provides a margin of at least a decay ration of 0.1 from
what is defined as a decay ration (1.0) that would result in an
unstable core. Since the revised curves are based on ensuring decay
rations of less than 0.9 are maintained, the existing margin of safety
is maintained.
Changing the wording of the TS to require that action be taken to
reduce the measured decay ratio if any two neutron signals of ``greater
than or equal to 0.75,'' as opposed to the current wording ``greater
than .75,'' are measured, would not have any measurable effect on the
implementation of the affected TS, and would, if anything, result in
action being taken at a maximum lower value than the current TS. This
would not have any significant impact on how close the plant was
allowed to operate to potential core power instability, and would not,
therefore, have a significant effect ont he margin of safety related to
the proposed TS.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: Nicholas S. Reynolds, Esq., Winston &
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: Theodore R. Quay
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: July 29, 1993, with supplemental
information provided March 11, 1994 and March 17, 1994
Description of amendment request: The amendment proposes to modify
the Technical Specifications (TS) to reflect a new refueling platform.
Specifically, the amendment would add new values for protective
features in the TS to reflect the new refueling platform. Values for
the old refueling platform are retained 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. The staff's evaluation of the licensee's analysis is
presented below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
The only accident evaluation affected by the proposed changes are
those associated with the Fuel Handling Accident (FHA) analyses
presented in WNP-2 Final Safety Analysis Report (FSAR) section 15.7.4.
As discussed therein, the fuel handling accident event that produces
the largest number of failed spent fuel rods is the drop of a spent
fuel bundle into the reactor core when the reactor vessel head is off.
The probability of dropping a spent fuel assembly onto other fuel
assemblies in the reactor vessel does not increase with the new design.
The NF500 mast functions identically to the old mast when grappling,
lifting, or moving a fuel assembly. It does not degrade platform design
features such as grapple fail-safe on loss of air, dual lifting cables,
backup cable reel brake, and the grapple engaged loaded interlock, all
of which serve to protect against a fuel drop event. The new mast is
more rigid than the previous mast design and, therefore, is less prone
to mast bowing. The consequences of dropping a fuel assembly are also
unaffected because the weight of the mast is not considered in existing
FHA analysis. The number of postulated fuel pins which fail as a result
of the FHA is unaffected since the energy imparted by the dropped
assembly is independent of the mast design, and mitigating systems will
function as previously analyzed. Further, analysis by GE of a
postulated accident in which the exposed portion of the NF500 mast is
struck by a missile and severed while lifting a fuel bundle with both
falling onto the top of the core has been conducted, showing that the
consequences of the increased weight of the mast and bundle are bounded
by the current WNP-2 FSAR analysis for the fuel bundle only FHA.
Retaining the ability to use the old mast does not introduce any
changes to the current TS that reflect the analysis of the old mast.
The proposed change would not, therefore, significantly increase the
probability or consequences of a previously analyzed accident.
2. Does the amendment create the possibility of a new or different
kind of accident from any accident previously evaluated?
No new failure modes are introduced as a result of the proposed
changes. The NF500 mast in intended as an exact replacement for the
currently installed mast, and is designed to match or exceed the
strength and performance of the NF400 mast in all areas. No new fuel
handling methods or surveillance procedures will be necessary as a
result of installation of the new mast. The proposed change does not
affect the manner in which protective interlocks operate. Limits on
fuel travel in all directions are unchanged. Retaining the ability to
use the NF400 mast presents no new accident possibilities since no
changes in fuel mast operation would result from use of the existing
mast. The proposed change would not, therefore, create the possibility
of a new or different kind of accident from any previously analyzed.
3. Does the amendment involve a significant reduction in a margin
of safety?
The changed refueling mast cutoff and interlock values account for
the increase weight of the mast, or a portion thereof, and do not
affect the margins related to the fuel bundle drop analyses. The new
mast has the same single failure protection as the old mast. The
proposed change would not, therefore, involve a significant reduction
in a margin of safety.
The NRC staff has determined that 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: Nicholas S. Reynolds, Esq., Winston &
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: Theodore R. Quay
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: December 20, 1993
Description of amendment request: The amendment proposes to modify
the Technical Specifications (TS) to address new containment purge and
vent valves to be installed in the 1994 refueling outage. The TS are
being modified to remove the requirement to ensure the remaining
existing-design valves' position remains at less than or equal to
70 deg. because the valves have a permanently installed mechanical stop
to limit the open position to ensure adequate closure times. In
addition, this modification is being requested because the current TS
are too limiting for the new valves, which are designed to close from a
90 deg. open position. The TS are also being modified to change the
containment leak testing requirements for the new valves from 6 months
to 2 years, to reflect the improved seat design of the replacement
valves. Additional administrative changes are proposed to delete an
out-of-date note, and to relocate an action statement requirement from
surveillance section of the TS to the action statements section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
Regarding the removal of the requirement to ensure the remaining
existing-design valves' position remains at less than or equal to
70 deg.: The maximum open position of the containment purge and vent
valves is not one of the initiating events for any previously evaluated
accident in the WNP-2 FSAR. Thus the proposed change will not affect
the probability of an accident previously evaluated. The containment
purge and vent valves' position is considered in the accident analyses,
and could affect the analyzed consequences of events. The current
limiting condition for operation (LCO) and Action Statement requires,
and the surveillance verifies, that the permanently installed 70 deg.
block is in place and effective. If the existing valves were open
further than 70 deg., the valves may not close in time. The valves have
a welded mechanical stop installed that limits the position to no more
than 70 deg. open, which is a fixed condition that can only be changed
by plant modification requiring evaluation against the requirements of
10 CFR 50.59. The licensee considers the mechanical stop as sufficient
to ensure the existing valves will remain within existing analysis
bounds for a design basis loss of coolant accident (LOCA). In addition,
the new valves are qualified to close within the 5 seconds assumed in
the design basis LOCA. The licensee considers, therefore, that the
existing and new valves will operate as required for accident
mitigation with the proposed change, and that the proposed change will
not affect the consequences of accidents previously evaluated.
Regarding the modification of the containment leak testing
requirements to reflect the new design valves: The containment purge
and vent valves are not one of the initiating events for any previously
evaluated accident in the WNP-2 FSAR. Thus the proposed change will not
affect the probability of an accident previously evaluated. The metal
to metal seat valves meet the Appendix J criteria necessary to be
tested as type C valves. Type C valves can be tested every 2 years,
compared to every 6 months for the current valves. The testing
frequency is based on the performance of the valve types to ensure that
they are capable of maintaining the necessary leak tightness over the
test interval. The new valves' design has been certified to provide the
same leak tightness over 2 years that the current valves provide over 6
months, thus the consequences of analyzed events remains unaffected by
the proposed change.
Regarding the administrative changes: The proposed change would (1)
delete a note that was applicable only through April 10, 1988, and (2)
move an action that is currently stated in the SURVEILLANCE
REQUIREMENTS section of a TS to the ACTIONS section of the same TS.
These changes do not affect the design or operation of the plant or the
implementation of the affected TS, and as such would not affect the
probability or consequences of previously analyzed events.
2. Does the amendment create the possibility of a new or different
kind of accident from any accident previously evaluated?
Regarding the removal of the requirement to ensure the remaining
existing-design valves' position remains at less than or equal to
70 deg.: No aspect of the design or plant operation is affected by
deletion of the surveillance or removal of the reference to the block
from the LCO and Action Statement, no new modes of plant operation are
introduced, and the proposed change does not require physical
modification of the plant. The valves not being replaced will continue
to be limited from opening greater than 70 deg. by the welded and non-
adjustable blocking feature. The capability of these valves to close
within 5 seconds to meet the limiting design basis accident (LOCA) will
remain unchanged. The replacement valves will be capable of closing
within the same 5 seconds from a full-open position of 90 deg.. Since
the proposed change does not introduce any new component, system, or
plant operating conditions, the change does not create the possibility
of a new or different kind of accident from any previously analyzed.
Regarding the modification of the containment leak testing
requirements to reflect the new design valves: The proposed change in
surveillance frequency for the replacement valves does not introduce
any new mode of plant operation, nor does it involve plant
modifications. The new valves operate in the same manner as the old
valves, only the seating surfaces are different. This does not affect
the way the valves operate to perform their function. The proposed
change would not, therefore, involve any new or different kinds of
accidents from any previously evaluated.
Regarding the administrative changes: The proposed changes do not
introduce any new modes of plant or equipment operation, nor do they
involve physical modification of the plant. The proposed change would
not, therefore, create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the amendment involve a significant reduction in a margin
of safety?
Regarding the removal of the requirement to ensure the remaining
existing-design valves' position remains at less than or equal to
70 deg.: The margin of safety of concern with the proposed change is
the need for the containment purge and vent valves to close in 5
seconds, which will ensure that part 100 limits for design basis events
are not exceeded. The proposed change does not affect the maximum open
position of the existing valves, thus the valves will still close
within 5 seconds. In addition, the new valves, with their maximum full
open position of 90 deg., are a new design that will still close within
the five seconds from the full open position, thereby preserving the
existing margin of safety.
Regarding the modification of the containment leak testing
requirements to reflect the new design valves: The margin of safety
involved in the proposed TS change is the amount of leakage that may
occur due to plant degradation that may affect the design basis
accident assumptions for leakage. The new design valves have been
certified to provide the same leak tightness over 2 years that the
current valves provide over 6 months, thus the leakage assumptions for
design basis events is unaffected. The proposed change would not,
therefore, affect the margin of safety provided by the TS.
Regarding the administrative changes: There are no margins of
safety affected by the administrative changes.
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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: Theodore R. Quay
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: January 6, 1994
Description of amendment request: The amendment proposes to modify
the Technical Specifications (TS) to remove the requirements for the
Seismic Monitoring Instrumentation from the TS and relocate them to the
FSAR and plant procedures. The requirements described in the
specifications will be maintained in the FSAR and plant procedures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The staff's evaluation of the licensee's analysis is
presented below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
The seismic monitors only provide monitoring and recording of
seismic events that might occur in the vicinity of WNP-2. The
instrumentation are not relied upon in current accident analyses for
any automatic or manual initiation of safety systems in response to a
seismic event. The proposed change would not, therefore, significantly
increase the probability or consequences of a previously analyzed
accident.
2. Does the amendment create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not affect the manner in which the plant
is operated, maintained, or tested. The proposed change would not,
therefore, create the possibility of a new or different kind of
accident from any previously analyzed.
3. Does the amendment involve a significant reduction in a margin
of safety?
The seismic monitors provide monitoring and recording functions
only, and are not relied upon in accident analyses for automatic or
manual initiation of any safety system. Thus the results of analyzed
events, and the associated margins of safety, are unaffected by the
administrative removal of the seismic monitors from the TS. The
proposed change would not, therefore, involve a significant reduction
in a margin of safety.
The NRC staff has determined that 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: Nicholas S. Reynolds, Esq., Winston &
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: Theodore R. Quay
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: February 17, 1994
Description of amendment request: The amendment proposes to modify
the Technical Specifications (TS) to support hydrostatic testing of the
reactor coolant system. Specifically, the proposed amendment would: (1)
add a Special Test Exception that would allow Mode 4 (Cold Shutdown)
operation up to 212 deg.F, compared to the current limit of 200 deg.F,
without shutdown cooling in operation, to conduct hydrostatic testing,
and (2) add a new reactor metal temperature vs reactor vessel pressure
(P/T) limit curve that is applicable up to 8 effective full power years
(EFPY), for use during hydrostatic testing and non-nuclear plant
heatup.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
Regarding the proposed Special Test Exception: The proposed change
would allow performance of hydrostatic testing in OPERATIONAL CONDITION
4 at temperatures greater than 200 deg.F but less than or equal to
212 deg.F. Operating in this condition is only allowed if specified
OPERATIONAL CONDITION 3 secondary containment requirements are met. The
operating condition is not considered as an initiator for any event
analyzed in the FSAR, therefore the proposed change would not affect
the probability of an accident previously evaluated.
The specified OPERATIONAL CONDITION 3 requirements compensate for
the allowed temperature increase and assure that the consequences of a
potential leak will be conservatively bounded by the existing FSAR
accident analyses, as discussed below.
The hydrostatic test is conducted near water solid, all rods in,
and temperature less than or equal to 212 deg.F. The stored energy in
the core will be very low (approximately 43 days of shutdown conditions
and partial core replacement during refueling) and the potential for
failed fuel and a subsequent increase in coolant activity above
Technical Specification limits is minimal. In addition, secondary
containment will be OPERABLE and capable of handling airbone
radioactivity from leaks that could occur during the performance of the
testing. Maintaining the temperature less than or equal to 212 deg.F
will ensure that any leak will not flash to steam, thereby ensuring the
potential for airborne activity remains low. Requiring the standby gas
treatment system (SGTS) to be OPERABLE will conservatively ensure that
any airborne radiation from leaks will be processed by the SGTS thereby
limiting releases to the environment. Existing pipe breaks analyzed in
Chapter 15 of the FSAR are bounding for the proposed condition. In the
event of a large break, the reactor would rapidly depressurize,
allowing the low pressure ECCS subsystems to operate. The capability of
the subsystems required for OPERATIONAL CONDITION 4 would be adequate
to keep the core flooded under this condition. Small system leaks would
be detected by leakage inspections before significant inventory loss
occurred. Thus the consequences of previously analyzed accidents are
not increased by the proposed amendment.
Regarding the proposed P/T limit curve: The proposed change would
modify the P/T limit curves that are based on prevention of brittle
fracture of the reactor vessel. The proposed change would result in
plant operation closer to the actual brittle fracture condition of the
reactor vessel, potentially making a brittle fracture more likely. This
condition is offset by the slow heatup conducted using only pump heat,
which would result in lower stresses in the reactor vessel than are
assumed in the brittle fracture analyses. The resulting P/T limit curve
based on 8 EFPY would have sufficient conservatism from the actual
vessel brittle fracture condition to make vessel failure as unlikely as
the original 32 EFPY curve.
The potential reactor vessel failure mechanisms are not affected by
the proposed change, therefore the consequences of previously analyzed
accidents are unaffected by the proposed change.
2. Does the amendment create the possibility of a new or different
kind of accident from any accident previously evaluated?
Regarding both the proposed Special Test Exception and the proposed
P/T limit curve: The proposed change introduces no new failure modes,
involves no physical modification to the plant or change in system
configurations, nor does it involve changes in plant, system, or
component operation. The proposed change, therefore, does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the amendment involve a significant reduction in a margin
of safety?
Regarding the proposed Special Test Exception: The hydrostatic test
is conducted with low stored energy in the reactor, which is bounded by
the assumed decay heat in current safety analyses. In the unlikely
event that a leak from the reactor coolant system were to occur, the
RPV would depressurize and the low pressure systems would be available
to keep the core flooded. This would ensure that the fuel peak clad
temperature would not exceed 2200 deg.F, which is the design basis that
provides the margin of safety for the reactor itself. In addition,
secondary containment will be maintained during the hydrostatic test,
which would ensure that any potential airborne activity that might
occur would be filtered through the SGTS. This would ensure that the
current margins to the 10 CFR Part 100 limits remain bounded by current
analyses. The proposed change would, therefore, not involve a
significant reduction in the margins of safety.
Regarding the proposed P/T limit curve: The proposed new curves
would allow plant operation closer to the actual brittle fracture
condition of the reactor vessel during hydrostatic test conditions
only. This would result in a reduced margin in the protection afforded
by the P/T curve. The new curves would, however, allow a lower
temperature for conduct of the hydrostatic test, which would increase
the heat sink available in the RCS, and increase the margin to decay
heat loads assumed in accident analyses. This would result in reduced
potential for extensive flow from any break, reduce the time for
initiation of low pressure ECCS systems, and reduce the available
radioactive decay products that are available for release during any
postulated accident condition. The overall impact of the conditions
increases the margin to 10 CFR Part 100 limits that are the design
margin of safety for postulated loss of coolant accidents. The overall
effect of the proposed change would not involve a significant reduction
in the overall margins of safety.
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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: Theodore R. Quay
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: February 23, 1994
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
6.8.c by removing the requirement to conduct a biennial review of plant
procedures in accordance with American National Standards Institute
(ANSI) N18.7-1976. The licensee proposes using alternate programs, that
are already in place, to ensure that procedures are periodically
reviewed and maintained current. A biennial review of the Integrated
Plant Emergency Operating Procedures (IPEOPs), however, would continue.
The requirements for these alternate programs and for the IPEOP review
would be added to the Operational Quality Assurance Program Description
(OQAPD).
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:
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1) 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 eliminating the periodic reviews of routine
administrative and technical procedures. Sufficient controls are
established to ensure that procedures impacting safety-related
structures, systems, and components are maintained current,
accurate, and usable. This TS change will therefore not impact the
function or method of operation of plant equipment. Thus, a
significant increase in the probability of a previously analyzed
accident does not result due to this change. No systems, equipment,
or components are affected by the proposed changes. Thus, the
consequences of a malfunction of equipment important to safety
previously evaluated in the Updated Safety Analysis Report (USAR)
are not increased by this change. The proposed changes do not affect
equipment or its operation, and, thus, do not affect the
probabilities or consequences of an accident. Therefore, WPSC
concludes that this change does not significantly increase the
probability or consequences of an accident.
2) 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 periodic procedure reviews do not
contribute to accident initiation, a change related to such an
activity does not produce a new accident scenario or produce a new
type of equipment malfunction. Also, this change does not alter any
existing accident scenarios. The proposed changes do not affect
equipment or its operation, and thus, do not increase the
possibility of a new or different kind of accident.
3) involve a significant reduction in the margin of safety.
The proposed changes do not affect equipment or its operation,
and thus, do not involve any reduction in the margin of safety.
Therefore, use of the proposed Technical Specification 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: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: John N. Hannon
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.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendment: February 17, 1994
Brief description of amendment request: The proposed amendments
would revise the combined Technical Specifications (TS) for the Diablo
Canyon Power Plant Unit Nos. 1 and 2 to revise TS 3/4.3.2, ``Engineered
Safety Feature Actuation System Instrumentation,'' as follows: (1)
Table 3.3-3, functional unit 6.c.2), channels to trip, would be changed
from 2/steam generator in one steam generator to 2/steam generator in
any 2 steam generators to correct an administrative error. (2) Table
3.3-4 would be changed as follows: a. functional unit 4.6., Negative
Steam Pressure Rate - High, trip setpoint and allowable value, would be
changed from -100 psi/sec and -105.4 psi/sec to 100 psi and 105.4 psi,
respectively; b. a note would be added stating that the time constants
utilized in the rate-lag controller for Negative Steam Pressure Rate -
High, are equal to 50 seconds.
Date of individual notice in Federal Register: March 1, 1994 (59 FR
9789)
Expiration date of individual notice: March 31, 1994
Local Public Document Room location: California Polytechnical State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: March 4, 1994
Brief description of amendment request: The proposed amendment
would add a new Section 3/4.10.8, ``Inservice Leak and Hydrostatic
Testing,'' and the Bases. The new section would allow Hope Creek to
remain in OPERATIONAL CONDITION 4 with reactor coolant temperatures up
to 212 *F to facilitate inservice leak and hydrostatic testing.
Date of publication of individual notice in Federal Register: March
16, 1994 (59 FR 12384)
Expiration date of individual notice: April 15, 1994
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
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: April 28, August 12, and November 17,
1993, and February 2, 1994
Brief description of amendment request: The proposed changes
increase the spent fuel pool capacities for Salem 1 and 2 from the
current 1170 fuel assemblies to 1632 fuel assemblies. Also, the decay
time for refueling operations is being extended from 100 hours to 168
hours.
Date of publication of individual notice in Federal Register: March
4, 1994 (59 FR 10440)
Expiration date of individual notice: April 4, 1994
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079.
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.
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of application for amendment: September 3, 1993, as
supplemented February 1, 1994
Brief description of amendment: The amendment revises the heatup
and cooldown curves and the low-temperature overpressure protection
(LTOP) controls. The changes to the LTOP controls support proposed
modifications to allow a variable-setpoint (VLTOP) protection system.
The VLTOP system will increase the allowable operating pressure band in
the LTOP region and increase the flexibility in the use of the reactor
coolant pumps.
Date of issuance: March 15, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 185
Facility Operating License No. DPR-53: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR 50963) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 15, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
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: September 17, 1993, as
supplemented on January 4, 1994
Brief description of amendments: The amendments implement the
recommendations provided in Generic Letter 88-16, ``Removal of Cycle-
Specific Parameter Limits From Technical Specifications,'' by removing
cycle specific values from the Technical Specifications (TSs) and
incorporating them in a separate document. The amendments also include
two other changes. One is the removal of outdated references to power
operation with less than four reactor coolant pumps in operation and
the other includes administrative changes to clarify the existing TSs,
but do not alter the existing requirements.
Date of issuance: March 17, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 186 and 163
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57844) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated March 17, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: August 27, 1993, as
supplemented February 21, 1994
Brief description of amendments: The amendments revise the
requirements for snubber visual inspection intervals and corrective
actions in accordance with Generic Letter 90-09. The amendments also
remove two of the options for determining the sample size to be used
for snubber functional testing.
Date of issuance: March 11, 1994
Effective date: March 11, 1994
Amendment Nos.: 60, 60, 48, and 48
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4935) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 11, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: June 1, 1992
Brief description of amendments: The amendments update the leakage
test requirements of the drywell airlock to the standards of 10 CFR
Part 50, Appendix J, Section III.D.2.
Date of issuance: March 11, 1994
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 125, 119, 145, and 141
Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 28, 1992 (57 FR
48818) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 11, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: For Dresden, The Morris Public
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,
The Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: February 25, 1993
Brief description of amendments: The amendments change Technical
Specification 4.8.1.1.2.c to update the diesel fuel oil testing
requirements to the standards of ASTM D4057-88 (new fuel oil test);
ASTM D975-88 (water and sediment content testing); and ASTM D2276-89
(impurity levels). The updated standards will be referenced in the
Technical Specification Bases.
Date of issuance: March 10, 1994
Effective date: March 10, 1994
Amendment Nos.: 97 and 81
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36431) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 10, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: October 28, 1993, supplemented
by letter dated January 21, 1994.
Brief description of amendments: The amendments revise the ECCS
injection valve stroke times and ECCS response times to allow the
licensee to perform Motor Operated Valve modifications that slow down
injection valve stroke times. As part of this change, a limited break
spectrum Loss-Of-Coolant Accident analysis was performed to evaluate
the impact of the slower response on the Peak Cladding Temperatures and
to update the plants licensing bases.
Date of issuance: March 9, 1994
Effective date: March 9, 1994
Amendment Nos.: 96 and 80
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4937) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 9, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: October 21, 1993
Brief description of amendments: The amendments delete the
requirements for demonstrating the operability of redundant equipment
when emergency core cooling system equipment is found to be inoperable,
or made inoperable for maintenance. The changes are consistent with the
guidance provided by the NRC staff in Generic Letter 93-05, dated
September 27, 1993.
Date of issuance: March 8, 1994
Effective date: March 8, 1994
Amendment Nos.: 144 and 140
Facility Operating License Nos. DPR-29 and DPR-30. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59747) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 8, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi
Date of application for amendment: April 21, 1993
Brief description of amendment: This amendment deleted License
Condition 2.C(36), Attachment 1, Item (c)(4) which implemented the
requirements of Regulatory Guide 1.97, ``Instrumentation For Light-
Water-Cooled Nuclear Power Plants to Assess Plant and Environs
Conditions During and Following an Accident,'' for the Grand Gulf
Nuclear Station because analysis shows that these requirements are
being met by alternative methods.
Date of issuance: March 7, 1994
Effective date: March 7, 1994
Amendment No: 112
Facility Operating License No. NPF-29. Amendment revises the
license.
Date of initial notice in Federal Register: May 12, 1993 (58 FR
28056) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 7, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, Post Office Box 1406, S. Commerce at Washington, Natchez,
Mississippi 39120.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 8, 1991, as supplemented by letters
dated March 6, 1992, and January 28, 1993
Brief description of amendment: The amendment revised the Technical
Specifications by revising the fuel oil amounts in the feed and storage
tanks for the emergency diesel generators, clarifying the testing for
the interconnecting piping, and revising the specific gravity of the
fuel oil.
Date of issuance: March 16, 1994
Effective date: March 16, 1994
Amendment No.: 92
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 26, 1991 (56 FR
29274), as revised April 14, 1993 (58 FR 19478) The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated March 16, 1994. No significant hazards consideration comments
received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: June 21, 1993
Brief description of amendments: These amendments will change
Technical Specifications Section 6.0, ``Administrative Controls,'' by
(a) revising unit staff titles to those of the current FPL Nuclear
Division organization, (b) revising the composition of the Facility
Review Group (FRG) to broaden the scope of available expertise, and (c)
making minor editorial corrections.
Date of issuance: March 2, 1994
Effective date: March 2, 1994
Amendment Nos.: 126, 65
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 21, 1993 (58 FR
39050) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 2, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
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: August 17, 1993, as
supplemented January 14, 1994.
Brief description of amendments: These amendments relocate fire
protection requirements from the Technical Specifications to the Final
Safety Analysis Report in accordance with Generic Letter 86-10,
``Implementation of Fire Protection Requirements,'' and amend the
license conditions accordingly.
Date of issuance: February 25, 1994
Effective date: February 25, 1994
Amendment Nos. 159 and 153
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Licenses and Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR 50967) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 25, 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. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: December 16, 1993
Brief description of amendment: The amendment revises the Technical
Specifications to clarify the requirements for maintaining secondary
containment integrity when one or more Reactor Building Ventilation
supply and exhaust valves are declared inoperable. The Technical
Specifications add a new Limiting Condition for Operation, Basis
Statement and Surveillance Requirements for these isolation valves.
Date of issuance: March 7, 1994
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 168
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4938) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated March 7, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, New Jersey
08753.
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: August 26, 1993
Brief description of amendment: The amendment revises the plant
Technical Specifications (TSs) to accommodate limited fuel
reconstitution based on NRC Generic Letter (GL) 90-02, Supplement 1.
Such reconstitution may be appropriate in the event of a leaking fuel
rod, in which case the fuel rod would be replaced with a stainless
steel or zirconium alloy filler rod.
Date of issuance: March 15, 1994
Effective date: As of its date of issuance, to be implemented
within 30 days of issuance.
Amendment No.: 183
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59751). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 15, 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.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: April 16, 1991, as supplemented
January 6, 1993.
Brief description of amendments: The amendments revise the
technical specifications to incorporate recommendations from NRC
Generic Letter 90-06 for power-operated relief valve and block valve
reliability and low-temperature overpressure protection.
Date of issuance: March 9, 1994
Effective date: March 9, 1994
Amendment Nos.: 176 & 161
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 3, 1993 (58 FR
12261) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 9, 1994
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: January 25, 1993, as
supplemented by letters dated November 3 and 23, and December 9, 1993,
and January 5 and 24, 1994.
Brief description of amendment: This amendment increases the
maximum number of spent fuel assemblies that can be stored in the Maine
Yankee fuel pool to 2019 from 1476. The increase in fuel storage
capacity is required so that storage space is available for spent fuel
through the duration of the current operating license, including the
final full core offload.
Date of issuance: March 15, 1994
Effective date: March 15, 1994
Amendment No.: 144
Facility Operating License No. DPR-36: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 26, 1993 (58 FR
16423) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 15, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, Maine 04578.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: December 22, 1993
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.4.4.e (Emergency Ventilation System) to permit
fuel handling operations to continue during refueling beyond 7 days
with one circuit of the emergency ventilation system inoperable,
provided the remaining emergency ventilation system circuit is operable
and in operation. The change to TS 3.4.4.e is consistent with the NRC's
Improved Standard Technical Specifications, NUREG-1433.
Date of issuance: March 8, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 146
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4940) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 8, 1994. No significant hazards
consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: December 27, 1993
Brief description of amendment: The amendment relocates TS Tables
3.2.7, ``Reactor Coolant Isolation Valves,'' and 3.3.4, ``Primary
Containment Isolation Valves,'' from TSs 3.2.7/4.2.7 and 3.3.4/4.3.4,
respectively, to a plant procedure which governs lists removed from TSs
per Generic Letter (GL) 91-08, ``Removal of Component Lists from
Technical Specifications.'' The plant procedure would be subject to the
requirements specified in the Administrative Controls section of the
NMP-1 TSs. The proposed amendment would also make conforming changes to
the TS Bases. These lists of valves will continue to be included in the
NMP-1 Updated Final Safety Analysis Report. Relocation of these valve
lists from the NMP-1 TSs to the plant procedure is consistent with NRC
staff guidance issued in GL 91-08.
Date of issuance: March 7, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 145
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4941) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 1994. No significant hazards
consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
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: June 11, 1993, supplemented by
letter dated November 15, 1993.
Brief description of amendment: The amendment revises the pressure/
temperature (P/T) limits for the reactor vessel. Specifically, Figure
3.4-2, ``Millstone Unit 2 Reactor Coolant System Pressure-Temperature
Limitations for 12 Full Power Years,'' on page 3/4 4-19, is revised to
reflect the change in the curves and the title change to ``Millstone
Unit 2 Reactor Coolant System Pressure-Temperature Limitations for 20
EFPY.''
Date of issuance: January 27, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 170
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 21, 1993 (58 FR
39054) The November 15, 1993, submittal provided information that did
not change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 27, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: May 7, 1993
Description of amendment request: The amendment changes Technical
Specification (TS) 3/4 6.1 relating to primary containment integrity.
Limiting Condition for Operation (LCO) 3.6.1.7 is changed to delete the
requirements applicable to the 36-inch containment shutdown purge
supply and exhaust isolation valves in the containment air purge (CAP)
system. Surveillance Requirement (SR) 4.6.1.7.1 and associated footnote
and SR 4.6.1.7.2 are deleted also. To maintain document consistency,
certain other editorial changes were made.
Date of issuance: March 7, 1994
Effective date: Not effective until operational MODE 5 is entered
when commencing the third refueling outage, and is to be implemented
prior to reentering operational MODE 4.
Amendment No.: 29
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34083). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 7, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library, 47
Front Street, Exeter, New Hampshire 03833.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: July 7, 1993
Brief description of amendments: The amendments revise the combined
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit
Nos. 1 and 2 to change TS 5.1.3, ``Map Defining Unrestricted Areas and
Site Boundary for Radioactive Gaseous and Liquid Effluents,'' to be
consistent with a recent interpretation of the restricted area
definition in 10 CFR 20.
Date of issuance: March 3, 1994
Effective date: March 3, 1994
Amendment Nos.: 90 & 89
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43930) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 3, 1994, and an
environmental assessment was published in the Federal Register on
February 25, 1994, (59 FR 9252). No significant hazards consideration
comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: September 8, 1993 (Reference
LAR 93-06)
Brief description of amendments: The amendments revise the combined
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit
Nos. 1 and 2. Specifically, TS 1.44, ``Radiological Monitoring and
Controls Program,'' 3/4.11, ``Radioactive Effluents,'' and 6.14,
``Radiological Monitoring and Controls Program (RMCP), Offsite Dose
Calculation Procedure (ODCP) and Environmental Radiological Monitoring
Procedure (ERMP),'' are revised to change the Semiannual Radioactive
Effluent Release Report to Annual Radioactive Effluent Release Report.
The amendment also revises TS 6.2.3, ``Onsite Safety Review Group
(OSRG),'' 6.5.2, ``Plant Staff Review Committee,'' and 6.5.3.7,
``Nuclear Safety Oversight Committee Review,'' to implement
organizational changes.
Date of issuance: March 7, 1994
Effective date: March 7, 1994
Amendment Nos.: 91 and 90
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57855) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 7, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: October 8, 1993
Brief description of amendments: The amendments revised the
existing definition of CHANNEL CALIBRATION in Technical Specification
1.4 to allow in-place qualitative methods to be used to verify
resistance temperature detector or thermocouple sensor behavior.
Date of issuance: March 8, 1994
Effective date: March 8, 1994
Amendment Nos.: 133 and 102
Facility Operating License Nos. NPF-14 and NPF-22. These amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59754) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 8, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Philadelphia Electric Company, Public Service Electric and Gas
Company Delmarva Power and Light Company, and Atlantic City
Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: November 1, 1993 as
supplemented January 26, 1994 and February 18, 1994.
Brief description of amendments: These amendments concern the
Radiation Monitoring Systems - Isolation and Initiation Functions of
the Technical Specifications and are necessary to support modification
5281. This modification replaces the obsolete control room ventilation
radiation monitoring equipment.
Date of issuance: March 15, 1994
Effective date: March 15, 1994
Amendments Nos.: 184 and 189
Facility Operating License Nos. DPR-44 and DPR-56: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64614) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 15, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: July 15, 1993
Brief description of amendment: The amendment revises the Technical
Specifications to eliminate the reactor scram and Main Steam Line
Isolation Valve closure requirements associated with the Main Steam
Line Radiation Monitors. The changes are consistent with Licensing
Topical Report NEDO-31400, ``Safety Evaluation for Eliminating the
Boiling Water Reactor Main Steam Isolation Valve Closure Function and
Scram Function of the Main Steam Line Radiation Monitor,'' dated May
1987.
Date of issuance: March 9, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 207
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41513) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 9, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: October 16, 1992
Brief description of amendments: These amendments revise TS 3/
4.3.4, ``Turbine Overspeed Protection,'' to allow one surveillance
every 31 days for verification of turbine overspeed protection system
operability. Currently, the surveillance tests are performed at power
every 7 days and again every 31 days. The 31-day test is performed by
an operator with an observer at the valve.
Date of issuance: March 9, 1994
Effective date: March 9, 1994
Amendment Nos.: 111 and 100
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 17, 1993 (58
FR 8783) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 9, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama
Date of application for amendments: January 10, 1992 (TS304)
Brief description of amendments: The amendments address emergency
diesel generator availability for the plant shared systems of Standby
Gas Treatment and Control Room Emergency Ventilation.
Date of issuance: March 9, 1994
Effective date: March 9, 1994
Amendment Nos.: 203, 222 and 176
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 1992
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 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: January 8, 1993; which was
supplemented by submittals dated April 1, May 3, and August 18, 1993;
and February 22, 1994.
Brief description of amendments: The amendments remove the
surveillance requirement to perform reactor vessel nozzle inspections
at the end of each 10-year inspection interval.
Date of issuance: March 15, 1994
Effective date: March 15, 1994
Amendment Nos.: 177 and 168
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: February 3, 1993 (58 FR
7007) The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated March 15, 1994. No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
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: July 2, 1993, as supplemented
December 10, 1993
Brief description of amendments: These amendments modify the
Technical Specifications having cycle-specific parameters limits by
replacing the values of those limits with a reference to a Core
Operating Limits Report for the values of those limits.
Date of issuance: March 2, 1994
Effective date: March 2, 1994
Amendment Nos. 189 and 189
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41519) The December 10, 1993, submittal did not expand the scope of the
original application and did not change the proposed no significant
hazards determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated March 2, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Virginia Electric and Power Company, Docket Nos. 50-280, 50-281,
50-338, and 50-339, Surry Power Station, Unit Nos. 1 and 2, Surry
County, Virginia, and North Anna Power Station, Unit Nos. 1 and 2,
Louisa County, Virginia.
Date of application for amendments: July 20, 1993
Brief description of amendments: These amendments delete the
Technical Specifications requirement for Station Nuclear Safety and
Operating Committee review of the Emergency and Security Plans. This
requirement remains in the respective plans. The audit frequencies are
also being deleted from the TS.
Date of issuance: March 1, 1994
Effective date: March 1, 1994
Amendment Nos. 188, 188, (Surry 1&2) 180, 161 (North Anna 1&2)
Facility Operating License Nos. DPR-32, DPR-37, NPF-4 and NPF-7:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46242) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 1, 1994. No significant
hazards consideration comments received: No
Local Public Document Room locations: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185, and 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: December 10, 1993
Brief description of amendments: These amendments modify the
surveillance requirements for the Auxiliary Feedwater System pumps and
valves, define ``staggered test basis,'' and make administrative
changes to the Technical Specifications.
Date of issuance: March 7, 1994
Effective date: March 7, 1994
Amendment Nos. 190 and 190
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2873) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 1994. 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 April 29, 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).
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: February 25, 1994 as
supplemented on March 11, 1994
Brief description of amendment: The amendment revised the Technical
Specifications by adding a footnote to Specification 3/4.4.3.1,
``Reactor Coolant System Leakage - Leakage Detection Systems,'' to
permit continued plant operations with inoperable drywell floor drain
sump flow monitoring instrumentation until the first time the plant is
required to be brought to COLD SHUTDOWN after March 15, 1994.
Date of issuance: March 14, 1994
Effective date: March 14, 1994
Amendment No.: 89
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazardsconsideration: No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated March
14, 1994.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook,
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of application for amendment: December 15, 1993, as
supplemented February 15 and 24, 1994 (December 15, 1993, application
supersedes the licensee's March 10, 1993 application.)
Brief description of amendment: The amendment revises the Technical
Specifications to allow the continuance of voltage-based steam
generator tube plugging criteria for outside-diameter stress corrosion
cracking at tube support plate elevations. The amendment allows the use
of a 2.0 volt interim repair criterion for Cycle 14 operation.
Date of issuance: March 15, 1994
Effective date: March 15, 1994
Amendment No.: 178
Facility Operating License No. DPR-58. Amendment revises the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration. Yes. The December 15, 1993,
application was noticed in the Federal Register on January 5, 1994 (59
FR 621). The NRC also 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 finding
be provided to the staff by the close of business on March 7, 1994. The
notice was published in the South Haven Tribune on March 1, 1994, and
in the Herald-Palladium on March 2, 1994. No comments have been
received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of Michigan, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated March 15, 1994.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
NRC Project Director: Ledyard B. Marsh
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook,
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of application for amendment: February 15, 1994
Brief description of amendment: The amendment revises the Technical
Specifications
Date of issuance: March 14, 1994
Effective date: March 14, 1994
Amendment No.: 177
Facility Operating License No. DPR-58. Amendment revises the
Technical Specifications. Public comments requested as to proposed no
significant 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 finding be provided to the staff by the close of business
on March 7, 1994. The notice was published in the South Haven Tribune
on March 1, 1994, and in the Herald-Palladium on March 2, 1994. No
comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultantion with the State of Michigan, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated March 14, 1994
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
NRC Project Director: Ledyard B. Marsh
Dated at Rockville, Maryland, this 23rd day March 1994.
For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[Doc. 94-7331 Filed 3-29-94; 8:45 am]
BILLING CODE 7590-01-F
