[Federal Register Volume 59, Number 51 (Wednesday, March 16, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10316]
[[Page Unknown]]
[Federal Register: March 16, 1994]
=======================================================================
-----------------------------------------------------------------------
UNITED STATES NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating LicensesInvolving
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 February 18, 1994, through March 4, 1994.
The last biweekly notice was published on March 2, 1994 (59 FR 9999).
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 15, 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: January 20, 1994
Description of amendment requests: The proposed amendment would
change the departure from nucleate boiling ratio (DNBR) in Safety
Limits, Section 2.1.1.1, and the associated Bases, as well as the DNBR
- Low Trip Setpoint in Table 2.2-1, and the associated Bases, from a
value of 1.24 to 1.30. In addition, the amendment would add a
methodology supplement entitled,System 80GT1TMInlet
Flow Distribution,'' to the list of methods used to determine the core
operating limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1 -- Involve a significant increase in the probability
or consequences of an accident previously evaluated.
The purpose of the proposed TS amendment is to provide a revised
DNBR Safety Limit and Low DNBR Trip Setpoint to ensure that no
anticipated operational occurrence or postulated accident will
result in core conditions exceeding DNBR Safety Limit.
The change in the DNBR Safety Limit from 1.24 to 1.30 can be
accommodated directly by increasing the limit (including the DNBR
Trip Setpoint) or by an increase in the DNBR overall uncertainty
factors for core operating limit supervisory system (COLSS) (EPOL2
and EPOL4) and core protection calculator (CPC) (BERR1). Using the
1.24 DNBR Safety Limit will result in larger uncertainty factors,
and conversely using the increased DNBR Safety Limit of 1.30 will
result in lower uncertainty factors. Therefore, plant operation for
COLSS and CPC are not significantly affected by the choice of the
DNBR Safety Limit and the Trip Setpoint as long as the corresponding
overall uncertainty factors are calculated and implemented. PVNGS
will implement the 1.30 DNBR Safety Limit and its corresponding
overall uncertainty factors in the reload safety analysis for Unit 3
Cycle 5 and in subsequent reload safety analyses for Units 1 and 2.
The proposed amendment changes only the DNBR Safety Limit and
associated Trip Setpoint, and does not in any way impact the
operation of the plant. Safety and setpoint analyses will be
performed consistent with the increased DNBR limit of 1.30. The core
power distribution during all phases of normal and anticipated
operational occurrences will remain bounded by the initial
conditions assumed in Chapter 15 of the PVNGS Updated Safety
Analysis Report (UFSAR). Furthermore, the UFSAR Chapter 15 analysis
remains bounding because the margins of safety will be maintained.
Therefore, the proposed change to Sections 2.1.1.1 and 2.2.1 (Table
2.2-1) will not significantly increase the probability or
consequences of an accident previously evaluated.
The proposed change to Section 6.9.1.10 does not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The proposed change is administrative
in nature and does not involve any change to the configuration or
method of operation of any plant equipment that is used to mitigate
the consequences of an accident. Also, the proposed change does not
alter the conditions or assumptions in any of the UFSAR accident
analyses. Since the FSAR accident analyses remain bounding, the
radiological consequences previously evaluated are not adversely
affected by the proposed change. Therefore, it can be concluded that
the proposed change to Section 6.9.1.10 will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Standard 2 -- Create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed amendment is limited to changing the DNBR Safety
Limit and Low DNBR Trip Setpoint and does not involve any physical
change to plant systems or to the COLSS and the CPC algorithms.
These changes will not affect any safety-related equipment used in
the mitigation of anticipated operational occurrences or design
basis accidents. Therefore, this change to Section 2.1.1.1 and 2.2.1
(Table 2.2-1) will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed change to Section 6.9.1.10 does not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed change is administrative in
nature and does not involve any change to the configuration or
method of operation of any plant equipment that is used to mitigate
the consequences of an accident. Accordingly, no new failure modes
have been defined for any plant system or component important to
safety nor has any new limiting failure been identified as a result
of the proposed change. Therefore, it can be concluded that the
proposed change to Section 6.9.1.10 will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Standard 3 -- Involve a significant reduction in a margin of
safety.
The DNBR Safety Limit specified in TS 2.1.1.1 and the Low DNBR
Trip Setpoint specified in TS 2.2.1 (Table 2.2-1) ensure that
operation of the reactor is prevented from exceeding the DNBR Safety
Limit during normal operation and design basis anticipated
operational occurrences. Therefore, operating within the increased
DNBR Safety Limit will ensure that no anticipated operational
occurrence or postulated accident will result in core conditions
exceeding the specified DNBR Safety Limit. The UFSAR Chapter 15
analysis remains bounding because the margins of safety will be
maintained. Additionally, the COLSS and the CPC overall uncertainty
factors will be calculated and implemented consistent with the
increased DNBR Safety Limit of 1.30. Therefore, this change to
Section 2.1.1.1 and 2.2.1 (Table 2.2-1) will not result in a
significant reduction in a margin of safety.
The proposed change to Section 6.9.1.10 does not involve a
significant reduction in a margin of safety. The proposed change is
administrative in nature and does not adversely impact the plant's
ability to meet applicable regulatory requirements. Therefore, it
can be concluded that the proposed change to Section 6.9.1.10 does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensees' analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004Attorney 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
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: January 9, 1994
Description of amendment request: The proposed amendment would make
changes to the Technical Specifications and License. These changes
consist of revised wording for the license, clarify wording to aid
operators in selecting the correct pressure/temperature curve during
startup and shutdown operations, and removal of certain obsolete
mechanical snubber acceptance criteria.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The first proposed change will modify License DPR-35 to
eliminate the need to issue a new page 3 to identify the latest
amendment number. The second change will provide the correction of
an error of omitting the reference to the subcritical mode of
operation, in relation to the pressure/temperature curves. The third
change will remove the unnecessary mechanical snubber functional
test acceptance criterion to determine if drag force has increased
more than 50% since the last functional test.
Modification of License DPR-35 for Pilgrim Nuclear Power Station
to remove the need to update page 3 whenever a new amendment is
approved will reduce an administrative burden. This license change
also precludes a possible administrative error if the correct
reference is somehow missed. This change does not affect plant
operation or design and is considered an administrative change and
as such does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The second change corrects an error of omission made in an
earlier amendment by inserting a reference to the subcritical
reactor operation phase. This proposal will enhance the procedure
changes and training already accomplished as short term corrective
actions. This change does not affect plant operation or design and
is considered an administrative change and therefore does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The third change removes an acceptance criterion for mechanical
snubber testing not required by the ASME Boiler and Pressure Vessel
Code, Section XI, Subsection IWF nor recommended by the vendor for
mechanical snubbers in use at Pilgrim.
This change will not result in any physical modification to
Pilgrim. The mechanical snubbers will continue to be tested in
accordance with existing plant procedures which reference the ASME
Code Section XI, Subsection IWF. Therefore, this is considered an
administrative change and as such, operation of Pilgrim will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident than previously evaluated because they
are administrative in nature and require no physical alterations of
plant configuration or changes to setpoints or operating parameters.
3. The operation of Pilgrim in accordance with the proposed
amendment will not involve a significant reduction in a margin of
safety.
Because these changes do not alter plant operation or design and
are considered administrative in nature, they 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 request involves no significant hazards consideration.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Walter R. Butler
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of amendments request: April 13, 1993
Description of amendments request: The proposed amendments would
revise design features information pertaining to the elevation at which
the spent fuel pool is designed to prevent inadvertent draining. The
proposed amendment would revise this elevation from 116 feet 4 inches
to ll5 feet 11 inches based on the actual spent fuel pool design.
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 amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The current value in Specification 5.6.2 is incorrect. No basis
can be determined for including this value in Specification 5.6.2
other than incorrectly characterizing the normal fuel pool water
level as the design level to be maintained to prevent inadvertent
draining of the fuel pool. This value was incorrectly incorporated
into the initial standard Brunswick Technical Specifications, Water
Level- Spent Fuel Storage Pool.
The 115' 11'' elevation is equal to 20' 10-7/8'' above the top
of the spent fuel rods seated in the storage racks. This level is
still in excess of the minimum level required (20' 6'') by Technical
Specification 3.9.9.
The accident discussed in UFSAR [Updated Final Safety Analysis
Report] Section 9.1.2.3.2.4.2, Loss of Spent Fuel Pool Cooling, is
not impacted by this change since the spent fuel pool safety
functions are not impacted and Technical Specification minimum fuel
pool levels (Specification 3.9.9) are not changed. As such, the
proposed amendments do not involve a significant increase in the
probability of an accident previously evaluated.
The radiological consequences of this accident are discussed in
UFSAR Section 9.1.2.3.2.5. This analysis assumes spent fuel pool
boiling. In addition, the facilities description of the spent fuel
storage pool (Section 9.1.2.2.1), states that the surface of the
water will be maintained at Elevation 116.3 ft, which is the normal
water level of the pool. Therefore, the (lower) designed level to
prevent inadvertent pool draining is not relevant within this
analysis. As such, the proposed amendments do not involve a
significant increase in the consequences of an accident previously
evaluated.
2. The proposed amendments do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This amendment request corrects a mischaracterization of the
design features and does not involve a change in fuel pool
operations. Specification 5.6.2 states that the fuel pool is
designed to prevent inadvertent draining of the pool below elevation
116'4''. However, it is possible that the pool could drain below
this level by draining through piping connected to the pool coupled
with no flow into the pool. It is not possible, however, with the
fuel pool gates installed, that the fuel pool could be inadvertently
drained below the bottom of the pool overflows to the skimmer surge
tanks. The elevation at the bottom of the overflows to the skimmer
surge tanks is 115'11''. Therefore, this is the correct value to
cite in the design features section of Technical Specifications.
The proposed 115'11'' elevation will not result in new drain
pathways, nor will the minimum fuel pool water level required by the
Technical Specifications be impacted by this change. Therefore, the
proposed amendments do not create the possibility of a new or
different type of accident from any accident previously evaluated.
3. The proposed amendments do not involve a significant
reduction in the margin of safety.
The proposed amendments do not change safety limits, setpoints,
or plant operations. The plant is actually designed to prevent
inadvertent draining of the fuel pool below elevation 115'11'' as
discussed above. This change is not an actual design change; it is a
design clarification correcting the level at which inadvertent
draining of the spent fuel pool is prevented. As such, the proposed
amendments do not involve a significant reduction in the margin of
safety at Brunswick.
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 North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
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
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) in response to Generic Letter
93-08 issued by the NRC and dated December 29, 1993, by relocating the
reactor trip system (RTS) and engineered safety feature activation
system (EFAS) response time limits to the updated Final Safety Analysis
Report (FSAR).
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 are administrative in nature and do not
involve any change to the configuration or method of operation of
any plant equipment used to mitigate the consequences of an
accident.
The proposed changes do not alter the conditions or assumptions
in any accident previously evaluated.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The proposed changes are administrative in nature and do not
involve any change to the configuration or method of operation of
any plant equipment used to mitigate the consequences of an
accident. No new
accident initiators or failure modes are created by relocating
the RTS and ESFAS instrumentation response time limits.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any previously
evaluated.
The proposed changes are administrative in nature and will in no
way affect the TS adequacy in ensuring the response times for the
RTS and ESFAS instrumentation do not exceed the limits assumed in
the accident analyses. The proposed changes will have no impact on
the protective boundaries, safety limits, or margin of safety.
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
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: July 22, 1993, as supplemented February
4, 1994
Description of amendment request: The proposed amendment would
modify the Facility Operating License (OL) and Technical Specifications
(TSs) to permit uprated power operation. The plant is currently
licensed for operation at 3323 megawatts thermal (MWt), although many
of the original analyses were performed at a design power level of 3467
MWt. The proposed changes would redefine rated thermal power to be 3467
MWt, which represents an approximately 4.3 percent increase over the
currently licensed power level. Implementation of the power uprate
would require minor modifications, such as, resetting of the low set
safety relief setpoints, as well as the recalibration of plant
instrumentation to reflect the uprated power. The proposed changes
follow the generic guidelines for boiling water reactor power uprate
described in General Electric Topical Report, NEDC-31897P-1, ``Generic
Guidelines for General Electric Boiling Water Reactor Power Uprate,''
June 1991.
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:
Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
OL2C(1), TS 1.34 - Increase in Rated Thermal Power to
3467 MWt.
The changes in the OL and TS were evaluated and it was
determined that the probability (frequency) of a DBA [design-basis
accident] or other licensing event occurring is not a significant
function of the power level because the design and regulatory
criteria originally established for plant equipment (ASME [American
Society of Mechanical Engineers] code, IEEE [Institute of Electrical
and Electronics Engineers] standards, NEMA [National Electrical
Manufacturers Association] standards, Regulatory Guide criteria,
etc.) are still imposed for the uprated power level. Scram setpoints
are established such that there will be no significant increase in
scram frequency due to power uprate.
The consequences of hypothetical accidents which would occur
from 102% of the uprated power, as opposed to that previously
evaluated from 102% of the original power, are in all
cases insignificant, since the accident evaluations from 102% of
uprated power do not result in exceeding the NRC-approved acceptance
limits. A spectrum of hypothetical accidents and transients has been
investigated for uprated conditions and the bounding events have
been shown to meet the same regulatory criteria to which they are
currently licensed. In the area of core design, for example, the
fuel operating limits such as Maximum Average Planar Linear Heat
Generation Rate (MAPLHGR) and Safety Limit Minimum Criteria Power
Ratio (SLMCPR) are still met at the uprated power level.
The analysis of all limiting events (Section 9) [of General
Electric Topical Report, NEDC-31994P, ``Power Uprate Licensing
Evaluation for Nine Mile Point Nuclear Power Station Unit 2,''
Revision 1, May 1993] and cycle specific reload analyses will show
plant transients meet the criteria accepted by the NRC as specified
in NEDO-24011, GESTAR II. Challenges to fuel or ECCS [emergency core
cooling system] performance have been evaluated (Section 9.2) [of
NEDC-31994P] and shown to still meet the criteria of 10CFR50.46
using the methodology defined by Appendix K (Regulatory Guide 1.70,
USAR [Updated Safety Analysis Report] Section 6.3). Challenges to
the containment have been evaluated for uprated power (Section 4.1)
[of NEDC-31994P] and still meet 10CFR[Part]50 Appendix A Criterion
38, Long Term Cooling and Criterion 50, Containment. Radiological
Release events have been evaluated [Sections 8.4 and 8.5] [of NEDC-
31994P] and shown to be a small fraction of the criteria of
10CFR[Part]100 (Regulatory Guide 1.70 USAR Chapter 15).
The results of these analyses as discussed above demonstrate
that operation of [at] the power uprate level does not significantly
increase the probability or consequences of any accident previously
evaluated.
OL2C(7) - Change in Allowable Feedwater Temperature
This change is made to maintain an equivalent 20 deg.F allowable
operating range of final feedwater temperature for uprated power
(405 to 425 deg.F) as compared to the presently licensed range (400
to 420 deg.F). No change is made to the current method and criteria
for operation of the feedwater heating systems. The limiting
transient (Feedwater Controller Failure - Maximum Demand) has been
evaluated at a feedwater temperature of 405 deg.F (Section 9.1.3)
[of NEDC-31994P] to demonstrate compliance with all current thermal
limits criteria. Previous feedwater nozzle evaluations have shown
that operation with the proposed feedwater temperature operating
range is acceptable. Therefore, there is no significant increase in
the probability or consequences of an accident previously evaluated.
TS Table 2.2.1-1 - Reactor Protection System Instrument
Setpoints
The increases in steam dome high pressure scram instrument
setpoints are made to ensure that there is no significant increase
in the frequency of scrams due to operation at the higher pressure.
The increase of the high pressure scram setpoint by the same amount
as the increase in the planned operating pressure maintains the same
level of trip avoidance for scram as originally provided. The high
pressure scram is used as a backup to other scram signals. It has
been shown that this role is still adequate for uprated operation
with the revised setpoints (e.g., vessel overpressure protection).
Since the backup protectionfunctions and the current margins to trip
avoidance are maintained with the revised setpoints, there is no
significant increase in the probability or consequences of an
accident previously evaluated.
TS Bases Table B2.1.2-2 - Add footnote to applicability
of table to uprated operation.
The parameters listed in TS Bases Table B2.1.2-1 come from the
original statistical analysis performed for BWR4/5 core designs
(including NMP2) [Nine Mile Point Nuclear Station, Unit 2]. As
discussed in Section 3.2 of LTR2 (Reference 11-1) [General Electric
Topical Report NEDC-31984P, ``Generic Evaluation of General Electric
Boiling Water Reactor Power Uprate,'' July 1991], the uprated
average bundle power is used to determine the applicability of the
generic Safety Limit Minimum Critical Power Ratio (SLMCPR) basis for
each plant. The average bundle power for NMP2 after power uprate is
4.538 MWt per bundle. This value is acceptable for application of
the generic SLMCR statistical analysis to uprated NMP2. The generic
analysis is documented through NEDC-24011-P-A (GESTAR II) and NEDC-
31152P (GE Fuel Bundle Design). Therefore, there is no significant
increase in the probability or consequences of an accident
previously evaluated.
TS 4.1.5.c and TS 4.1.5.d2 - Increases in SLCS [standby
liquid control system] Surveillance Test Pressure and SLCS pump
discharge relief valve setpoint.
The standby liquid control system test pressure increase ensures
continued ability of the system to pump the required amount of
sodium pentaborate at the higher operating pressure associated with
power uprate. This test pressure increase is consistent with the
ATWS [anticipated transient without scram] analysis provided in
Section 9.3 [of NEDC-31994P]. The higher pressure setpoint for the
SLCS relief valve does not exceed the design capability of the SLCS
components. Surveillance testing at the increased pressure will
maintain the system's design capability for operation at uprated
conditions. These changes therefore do not increase the probability
or consequences of a previously evaluated accident.
TS Table 3.3.1-1 - Note (i), footnote (**), Action 6
footnote (*), and Table 3.3.4.2-1 - footnote (**)
The setpoints for the bypass of T/G [turbine generator] trip
scram and RPT [recirculation pump trip] at 30% of rated power are
changed to 125.8 psig and 136.4 psig to be consistent with uprated
power.
These changes reflect the redefinition of rated conditions. They
are consistent with the approach discussed in Section F.4.2(c) of
LTR1 (Reference 11-2) [General Electric Topical Report NEDC-31897P-
1, ``Generic Guidelines for General Electric Boiling Water Reactor
Power Uprate,'' June 1991]. There is no significant impact on the
transient safety analyses which establish core thermal operating
limit since T/G trips at this partial power setpoint are not
limiting. Therefore, there is no significant increase in the
probability or consequences of an accident previously evaluated.
TS Table 3.3.2-2 Item 1.C.3 - Increase in main
steamline high flow isolation differential pressure setpoint and
allowable value.
The main steamline high flow trip TS changes reflect the
redefinition of rated steam flow during uprated power operation and
the application of GE [General Electric Company] setpoint
methodology. The current analytical basis of 140% of rated steam
flow is maintained for uprated operation to ensure that an adequate
trip avoidance margin is maintained (e.g., for disturbances caused
by full closure testing of MSIVs [main steam isolation valves] or
turbine inlet valves). The revised setpoints ensure that there is no
effect on the probability of inadvertent isolation; they have no
effect on the probability of occurrence of a main steamline break.
The same isolation initiation function for the main steam line break
accident is maintained (Section 5.1.2.5) [of NEDC-31994P]. Therefore
these setpoint changes do not significantly increase the consequence
of the main steamline break accident.
TS Table 3.3.2-2 Item 1d - Increase the main steamline
tunnel temperature setpoints.
These isolation setpoints are changed to reflect the slight
increase (about 1 deg.F) in the operating temperature expected for
uprated operation. Margins between trip setpoints and operating
temperature are maintained. The increases will avoid unnecessary
trips. The revised trip setpoints were derived using the GE setpoint
methodology (documented in NEDC-31336). The setpoints still perform
their isolation function equivalent to current operation.
Therefore no significant increase in the probability or
consequences of an accident previously evaluated results from these
changes.
TS Table 3.3.4.1-2 - Increases in the ATWS RPT reactor
vessel high pressure trip and allowable setpoint.
The ATWS RPT high pressure setpoints are increased to correspond
to the increase in the steam dome operating pressure due to power
uprate. This increase maintains the current margin between the
operating condition and the trip setpoint to avoid unnecessary
trips. The capability of the system to adequately perform its ATWS
function with the new setpoints is shown in Section 9.3 [of NEDC-
31994P]. Therefore the change does not cause a significant increase
in the probability or consequences of an accident previously
evaluated.
TS Figure 3.4.1.1-1 - The figure is revised to reflect
new definition of rated thermal power in terms of megawatts thermal.
This change is made to be consistent with the new definition of
rated thermal power. The current restrictions on operation within
the restricted power/flow zone are unchanged. The basis for this
change is described in Section 3.2 of LTR2 (Reference 11-1) [NEDC-
31984P]. There is no significant change in the previously evaluated
potential for initiation of core thermal hydraulic instability.
Therefore, this TS change ensures that power uprate operation will
not cause a significant increase in the probability of [or]
consequences of an accident previously evaluated.
TS 3.4.2 - Increase of spring setpoints for the two
lowest set SRVs [safety relief valves].
The two low set SRV setpoints are increased to accommodate the
change in operating pressure after power uprate. This increase in
the SRV setpoints ensures that approximately the same difference is
maintained between the RPV [reactor pressure vessel] pressure and
the lowest SRV setpoint such that there is no increase in the number
of unnecessary SRV actuations. The increase in the spring setpoints
by the same amount as the increase planned for normal operation also
maintains acceptable simmer margin for the SRVs. The SRVs are
capable of operating at uprated temperatures and pressures as
evaluated generically in Section 4.6 of LTR2 [NEDC-31984P]. As
described in Sections 3.2 and 9.3.1 [of NEDC-31994P] a higher RPV
peak pressure results due to uprate conditions but it is maintained
well within the ASME Code allowable peak pressure of 1375 psig.
Therefore, no significant increase in the probability or
consequences of an accident previously evaluated is caused by this
change.
TS 4.4.6.1.3-1 - Revision to the neutron fluence lead
factor.
The increase in the lead factor from 0.41 to 0.46 reflects
updated calculations for the higher power level and projected
fluence distributions. This calculation accounted for the locations
of the NMP2 specimen capsules (at three locations on the vessel wall
around the core beltline region), and the projected uprated
equilibrium cycle spatial power distributions. Since the revised
lead factor is consistent with the requirements for vessel
surveillance, the change causes no significant increase in the
probability or consequences or [of] an accident previously
evaluated.
TS 3.4.6.2 and TS 4.4.6.2 - Increase of reactor steam
dome operating pressure limit.
This change to the dome operating pressure limit is consistent
with and meets the current design criteria used for evaluation of
steady state operating conditions and for the most limiting
transient and accident events, i.e., vessel overpressure protection
and a loss-of-coolant accident (Sections 3.2 and 4.3) [of NEDC-
31994P]. Therefore, there is no significant increase in the
probability or consequences of an accident previously evaluated.
TS 4.7.4b - Increase in RCIC [reactor core isolation
cooling] Surveillance Test Pressure.
The increase in the RCIC surveillance test pressure requires
system testing at the higher operating pressure with power uprate.
The RCIC system has been evaluated and demonstrated to be capable of
injecting its design flow rate at the higher reactor pressure
associated with power uprate as discussed in Section 4.2 of LTR2
(Reference 1) [NEDC-31984P]. This evaluation applies to NMP2 as
described in Section 3.8 [of NEDC-31994P]. Therefore, this TS change
ensures that power uprate operation will not cause a significant
increase in the probability or consequences of an accident
previously evaluated.
TS Bases 3/4.2 (References), and TS 6.9.1.9.b(1)
(Administrative Control) - Revised the reference for the LOCA [loss-
of-coolant accident] analysis methodology to the SAFER/GESTR-LOCA
methodology report.
These changes are made to incorporate the power uprate LOCA
licensing basis. Reference 1 of TS Bases 3/4.2 (References) and the
report noted in TS 6.9.1.9.b(1) are changed to reflect the improved
SAFER/GESTR-LOCA methodology used for the NMP2 loss-of-coolant
accident analysis for power uprate. This methodology has been
previously approved by the NRC. These changes are made for
documentation consistency and there is no significant increase in
the probability or consequences of an accident previously evaluated.
TS Bases Table B3.2.1-1 - Significant input parameters
used in the LOCA analysis.
The changes in the plant parameters used in the uprated LOCA
analysis are provided from the power uprate analysis (Reference 4-
15, Section 4) [General Electric Report NEDC-31830P, ``NMP2 SAFER/
GESTR-LOCA Loss-of-Coolant Accident Analysis,'' Revision 1, November
1990]. The power and steam flow values are consistent with
Regulatory Guide 1.49. The new analysis parameters will also be
included in USAR Table 6.3-1 as uprate is implemented. Detailed
information about application of the SAFER/GESTR methodology is
provided in Reference 1 of TS Bases 3/4.2 (Reference 4-14, Section
4) [General Electric Report NEDE-23785-1-PA, ``The GESTR/LOCA and
SAFER Models for the Evaluation of the Loss of Coolant Accident,''
Revision 1, October 1984]. The LOCA analysis (Section 4.3) [of NEDC-
31994P] shows that all required criteria are met for operation with
the uprated parameters.
Footnote (*) is revised to provide the correct reference for the
LOCA analysis parameters for NMP2 power uprate. Since the LOCA
analysis for uprated operation meets all required criteria, these TS
changes do not cause an increase in the probability or consequences
of an accident previously evaluated.
TS Bases B3/4.5.1 and B3/4.5.2 - Increase in required
capability of the HPCS [high pressure core spray] pump and the
corresponding differential pressure.
The increase in the differential pressure for HPCS pump flow
accommodates the increase in SRV setpoint valves as discussed for
changes to TS 3.4.2 earlier. This change maintains the currently
designed functional capability of the HPCS system to provide coolant
inventory during isolation conditions after a loss of feedwater flow
transient (backup to RCIC) and during a main steam line break
(outside containment) accident.
The small change in the HPCS pump flow (517 versus 516) [gpm] is
made to be consistent with the value used in the NMP2 SAFER/GESTR
analysis for a loss of coolant accident. This small change corrects
the Technical Specification bases for this parameter.
Therefore, this TS change ensures that power uprate operation
will not cause a significant increase in the probability or
consequences of an accident previously evaluated.
TS Bases B3/4.6.1.2, B3/4.6.1.5, and B3/4.6.2 - Maximum
containment pressure for leakage testing.
The bases for the value currently in the TS for the maximum
containment pressure are reworded to clarify that the maximum
containment pressure after power uprate will be maintained below the
current value used for containment leak rate testing. Section 4.1
[of NEDC-31994P] documents the containment analysis for power uprate
and shows a peak DBA-LOCA calculated pressure of 36.8 psig (less
than the current testing requirement of 39.75 psig). There is no
impact on currently approved requirements and test procedures.
Therefore, there is no significant increase in the probability or
consequences of an accident previously evaluated.
Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The Operating License changes in power level and allowable
feedwater temperature, and the associated Technical Specification
changes (all listed in Table 11-1) [of NEDC-31994P] will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Equipment that could be impacted by power uprate has been
evaluated. No new operating mode, equipment lineup, accident
scenario, or equipment failure mode has been identified. The full
spectrum of accident considerations defined in Regulatory Guide 1.70
have been reviewed and no new or different kind of accident has been
identified. Power uprate uses already developed technology and
applies it within the capabilities of already existing plant
equipment in accordance with presently existing regulatory criteria
to include NRC approved codes, standards, and methods. GE has
designed BWRs of higher power levels than the uprated power of any
of the currently operating BWR fleet and no new power dependent
accidents have been identified.
The Technical Specifications changes required to implement power
uprate require minor changes to the configuration of the plant, and
all the Technical Specification changes have been evaluated and are
acceptable.
Will the change involve a significant reduction in a margin of
safety?
OL2C(1), TS 1.34 - Increase in Rated Thermal Power to
3467 MWt.
Power uprate will not involve a significant reduction in a
margin of safety, since the licensing evaluations were performed
either at plant conditions higher than the proposed uprate
conditions, or used approved methodologies which incorporate
appropriate allowances for uncertainties. As discussed throughout
this report (e.g., Section 11.1) [of NEDC-31994P] and in Section 5
of Reference 11-2 [NEDC-31897P-1], the safety margins prescribed by
the Code of Federal Regulations have been maintained by meeting the
appropriate regulatory criteria. Similarly, the margins provided by
the application of the American Society of Mechanical Engineers
(ASME) design acceptance criteria where applicable have been
maintained (e.g., see Section 3.2) [of NEDC-31994P]. Other margin-
assuring acceptance criteria have also been maintained.
All limiting accident and transient analyses have been
reperformed at uprated power operating conditions consistent with
the requested Technical Specification changes. The NRC-approved
SAFER/GESTR-LOCA methodology was used in the LOCA analysis.
Additionally, Reference 11-2 [NEDC-31897P-1] addresses the BWR
generic acceptability of analytical evaluations for the loss of
feedwater transient, stability, core spray distribution, safety
limit minimum critical power ratio, containment atmosphere
combustibility, materials and coolant chemistry, and anticipated
transients without scram (ATWS).
As discussed in Section 5.2.3 of Reference 11-2 [NEDC-31897P-1],
offsite doses for the DBA/LOCA will increase proportionally to
reactor power and can be compared on a consistent basis.
As evaluated in Sections 8.5 and 9.2 [of NEDC-31994P], the
results remain a small fraction of the acceptance criteria of
10CFR[Part]100.
The radiological doses resulting from the DBA/LOCA and MSLB
[main steam line break] accidents were initially analyzed at 3489
MWt (105% of 3323 MWt). For the power uprate program, the increase
in the analyzed power level is only 1.3% (from 3489 to 3536 MWt)
which provides the uncertainty factor (2%) required by Regulatory
Guide 1.49.
It is concluded that there is no significant decrease in a
margin of safety.
OL2C(7) - Change in Allowable Feedwater Temperature
The increase in the lower limit of the allowable operating range
of the final feedwater temperature for uprate power was evaluated by
reanalyzing the Feedwater Controller Failure - Maximum Demand
transient at a feedwater temperature of 405 deg.F (Section 9.1.3)
[of NEDC-31994P]. In addition, the reactor pressure vessel feedwater
nozzle has been evaluated for the 20 deg.F range of feedwater
temperature. The results of these evaluations demonstrate that
current fuel thermal limits criteria and ASME Code criteria are met.
Therefore, there is no significant decrease in a margin of safety.
TS Table 2.2.1-1 - Reactor Protection System Instrument
Setpoints
The increases in the steam dome high pressure scram instrument
setpoints for uprated power were evaluated by determining if the
high pressure scram, which is used as a backup to other scram
signals, provides adequate overpressure protection. The evaluation
demonstrates that the backup protection function, with the revised
setpoints, continues to provide adequate overpressure protection at
uprated power conditions by meeting the applicable ASME Code
criteria. It is concluded that there is no significant decrease in a
margin of safety.
TS Bases Table B2.1.2-2 - Add footnote to applicability
of table to uprated operation
The nominal values of parameters used in the statistical
analyses of the fuel cladding integrity safety limit were re-
evaluated at power uprate conditions. This evaluation demonstrates
that the average bundle power at uprated conditions for NMP2 is
acceptable for application of the generic safety limit minimum
critical power ratio statistical analysis. Therefore, it is
concluded that there is no significant decrease in a margin of
safety.
TS 4.1.5.c and TS 4.1.5.d2 - Increase in SLCS
Surveillance Test Pressure and SLCS pump discharge relief valve
setpoint.
The SLCS surveillance test pressure was increased to provide
periodic demonstration of the ability of the SLCS to provide the
required amount of sodium pentaborate solution at the higher
pressure associated with an ATWS event postulated to occur at power
uprate conditions. At this increased SLCS discharge pressure, the
system provides an adequate shutdown backup capability by having the
ability to bring the isolated reactor from full power to a cold,
Xenon-free shutdown condition, assuming that the withdrawn control
rods remain fixed in the uprated power pattern.
For power uprate, the capability of the SLCS to respond with
adequate margin to a postulated ATWS event was confirmed. The most
limiting ATWS events evaluated for peak vessel pressure and peak
suppression pool temperature were: (1) closure of all MSIVs and (2)
inadvertent opening of a relief valve. The peak pressure for the
MSIV closure event which included simulation of the higher relief
setpoints and two relief valves out of service, demonstrates that
the peak pressure, 1325 psig, remains below the ASME emergency
overpressure protection criteria of approximately 1500 psig, which
is applicable to an ATWS event. The reactor pressure is controlled
by the relief valves (after the initial peak) within the pressure
specified in this revised Technical Specification. SLCS injection
takes place during this period with the relief valves controlling
pressure.
The peak suppression pool temperature for the inadvertent
opening of a relief valve was demonstrated to remain below the ATWS
peak pool temperature criteria of 190 deg. for a Mark II containment
design, which is applicable to NMP2. Peak containment pressure was
well below the 45 psig containment design pressure. For this event,
SLCS injection will be at vessel pressures bounded by the revised
Technical Specification. The higher pressure setpoint of the SLCS
pump discharge relief valve provides adequate overpressure
protection of the SLCS pressure boundary by meeting the applicable
ASME Code criteria (equal to or less than the system piping design
pressure).
In summary, peak vessel pressure is below ASME code criteria,
and suppression pool temperature is below the ATWS peak pool
temperature criterion for Mark II containment design, peak
containment pressure is well below the containment design pressure,
the SLCS injection pressure during the bounding events is within the
new Technical Specification testing requirement, and the SLCS
pressure boundary is maintained in compliance with ASME Code
criteria. Therefore, it is concluded that there is no significant
decrease in a margin to safety.
TS Table 3.3.1-1 - Note (i), footnote (**), Action 6,
footnote (*), and Table 3.3.4.2-1 footnote (**)
The increase in the setpoints for the bypass of T/G trip scram
and RPT at 30% power are made to be consistent with uprated power.
These increased setpoints do not significantly reduce a margin of
safety since the T/G trips at this partial power setpoint continue
to be non-limiting events.
TS Table 3.3.2-2 Item 1.c.3 - Increase in main steam
line high flow differential pressure setpoint and allowable valve.
The increase in the main steam line high flow differential
pressure setpoint and allowable value reflect the redefinition of
rated conditions. The increased setpoint will maintain the same
inadvertent trip avoidance margin, thereby avoiding any increase in
the frequency of occurrence of isolation events. The closure of the
MSIV remains assured during the limiting event (the steam line break
accident). The break flow rate (controlled by the flow restrictor)
will be about 190% (Section 3.5) [of NEDC-31994P], so the setpoint
at less than 140% will sense the accident as effectively as for
current operation. It is concluded that this change does not result
in a significant decrease in a margin of safety.
TS Table 3.3.2-2 Item 1.d - Increase in main steam line
tunnel temperature setpoints.
These isolation setpoints are changed to reflect the slight
increase (about 1 deg.F) in the steam tunnel operating temperature
expected for uprated operation. The increase in these setpoints
ensures adequate inadvertent trip avoidance. The analytical upper
limits for these setpoints are not changed so that their safety
functions are not impacted by the Technical Specification changes.
For example, the instruments will act at the same setpoints assumed
in previous analysis for a main steamline break, ensuring that
offsite radiological doses remain a small fraction of 10CFR[Part]100
criteria and within GDC19 criteria for control room doses.
Therefore, there is no significant decrease in a margin of safety.
TS Table 3.3.4.1-2 - Increases in the ATWS RPT reactor
vessel high pressure trip and allowable setpoints
The purpose of the high pressure RPT is to reduce reactor power
level during a postulated pressurization transient with scram
assumed to fail (ATWS). The physical phenomenon involved is that the
void reactivity feedback from a pressurization transient adds
positive reactivity to the reactor system. However, the high
pressure RPT system trips both recirculation pumps to the low speed
condition, thereby increasing core void fraction and creating
negative reactivity to reduce the power transient. This enables the
safety/relief valves to maintain peak pressure within the ASME
overpressure emergency limit for the bounding ATWS case (Section
9.3.1) [of NEDC-31994P].
For power uprate, the capability of the SLCS to respond to a
postulated ATWS event with adequate margin was confirmed (Section
9.3.1) [of NEDC-31994P]. By reducing reactor power until the SLCS
can be injected to achieve full shutdown, the RPT also reduces
suppression pool temperature for isolation cases (also shown to be
acceptable for power uprate conditions in Section 9.3.1) [of NEDC-
31994P]. Therefore, it is concluded that there is no significant
decrease in a margin of safety.
TS Figure 3.4.1.1-1 - The figure is revised to reflect
the new definition of rated thermal power in terms of megawatts
thermal.
This change is made to be consistent with the new definition of
rated thermal power. As described in Section 3.2 of LTR2 (Reference
11-1) [NEDC-31984], the change to the power flow restricted zone is
made to maintain the same operating constraints and stability margin
that were established for the current power level. This change
avoids any increase in the possibility of occurrence or any increase
in the potential effects of power oscillations. Therefore, there is
no significant decrease in a margin of safety.
TS 3.4.2 - Increase of spring setpoints for the two
lowest set SRVs.
The two low set SRV setpoints are increased to accommodate the
change in operating pressure after power uprate. This change
maintains a simmer margin of greater than 120 psig. Power uprate
analysis shows that the revised SRVs still maintain the peak RPV
pressure within the ASME Code Upset limit of 1375 psig for the
limiting pressurization event (MSIV closure when credit is only
taken for the backup high neutron flux scram) and provide adequate
protection for postulated ATWS events. See Sections 3.2 and 9.3.1
[of NEDC-31994P] for further discussion. Therefore, it is concluded
that there is no significant decrease in a margin of safety.
TS 4.4.6.1.3-1 - Revision of the neutron fluence lead
factor.
The increase in the lead factor includes consideration of the
higher power level and projected spatial power distributions for an
uprated equilibrium cycle. The evaluation at uprated conditions
utilized the same calculational approach as performed for the
current neutron fluence lead factor, but using more precise input
parameters for uprated conditions. Therefore, it is concluded that
there is no significant decrease in a margin of safety.
TS 3.4.6.2 and TS 4.4.6.2 - Increase of reactor steam
dome operating pressure limit.
The change to the dome operating pressure limit is made to be
consistent with the new operating pressure for uprated thermal
power. This change is used as a direct initial condition analysis
input or sensitivity study parameter in the evaluation of steady
state operating conditions and for the most limiting transients and
accident events, i.e., vessel overpressure protection and LOCA. With
this revised limit, peak vessel pressure remains below ASME Code
criteria, and LOCA fuel performance satisfies the requirements of
10CFR50.46 and 10CFR[Part]50 Appendix K. Therefore, there is no
significant decrease in a margin of safety.
TS 4.7.4b - Increase in RCIC Surveillance Test
Pressure.
The RCIC surveillance test pressure was increased to provide
periodic demonstration of the ability of RCIC system to perform
consistent with the requirements of the analyses at the higher
operating pressure associated with power uprate conditions. An
evaluation of the RCIC system confirmed its ability to operate at
slightly higher turbine speed and provide its design flow rate at
power uprate conditions. RCIC system performance will be confirmed
during the initial power ascension to uprated conditions (and
periodically thereafter per the Technical Specification). Therefore,
it is concluded that there is no significant decrease in a margin of
safety.
TS Bases 3/4.2 (References) and TS 6.9.1.9.b (1)
(Administrative Control) - Revised the references for the LOCA
analysis methodology to the SAFER/GESTR-LOCA methodology report.
These changes are made to incorporate the power uprate LOCA
licensing basis. Since SAFER/GESTR-LOCA methodology has been
previously approved by the NRC and is acceptable for use for NMP2,
it is concluded this change does not significantly decrease a margin
of safety.
TS Bases Table B3.2.1-1 - Significant input parameters
used in the LOCA analyses.
The changes in the plant parameters reflect the power uprate
condition. These changes have been reflected as input parameters in
the LOCA analyses consistent with Regulatory Guide 1.49. Since the
LOCA analysis demonstrates that 10CFR50.46 and 10CFR[Part]50
Appendix K criteria are met for operation with the uprated
parameters, it is concluded that there is no significant decrease in
a margin of safety.
TS Bases B3/4.5.1 and B3/4.5.2 - Increase in required
capability of the HPCS pump and the corresponding differential
pressure.
The increase in differential pressure accommodates the increase
in SRV setpoint values previously discussed to TS 3.4.2. The
increased differential pressure ensures there is sufficient HPCS
flow, assuming the two lowest setpoint SRVs are out of service, and
the commencement of flow as reflected in the analysis of isolation
events. The increase in HPCS pump flow is reflected in the LOCA
analyses (516 to 517 gpm). This small change corrects the Technical
Specification bases for this parameter. The revised parameters for
the HPCS pump differential pressure and flow are reflected as inputs
to the LOCA analyses and analyses of isolation events. Since the
LOCA analysis meets 10CFR50.46 criteria and 10CFR[Part]50 Appendix K
criteria, and the isolation events meet all required criteria (e.g.
top of fuel remains covered for the loss of feedwater transient) it
is concluded that there is no significant decrease in a margin of
safety.
TS Bases B3/4.6.1.2, B3/4.6.1.5 and B3/4.6.2 - Maximum
containment pressure for leakage testing.
The bases for the value currently in the TS for the maximum
containment pressure are reworded to clarify that the maximum
containment pressure for power uprate has been calculated to remain
below the current value used for containment leak rate testing.
Therefore, it is concluded that there is no significant decrease 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 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Robert A. Capra
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 can
not 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 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. 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 further changes
to the seismic monitoring instrumentation system will be evaluated
for the effect of the 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 shut down 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 instumentation
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
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: December 13, 1993, as supplemented
February 2, 1994
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to allow for the use and
subsequent storage of fuel with an initial enrichment of 5.0 w/o
[weight percent] Uranium 235. The TS currently allow the use of fuel
with a maximum enrichment of 4.25 w/o Uranium 235. The proposed
amendment would also revise the restrictions on fuel storage in regions
1 and 2 of the spent fuel pool to ensure that the design basis for
preventing criticality is maintained in the event of absorber panel
shrinkage.
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 change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
There is no increase in the probability of an accident because
the physical characteristics of a fuel assembly are not changed when
fuel enrichment is increased. Fuel assembly movement will continue
to be controlled by approved fuel handling procedures.
There is no increase in the consequences of an accident because
fuel cycle designs will continue to be analyzed with NRC-approved
codes and methods to ensure the design bases for VCSNS [Virgil C.
Summer Nuclear Station] are satisfied. The double contingency
principle of ANSI/ANS 8.1-1983 can be applied to any postulated
accident in the spent fuel pool which could cause reactivity to
increase beyond the analyzed conditions. As shown in Attachment IV,
the level of boron in the VCSNS spent fuel pool is sufficient to
maintain Keff [effective neutron multiplication factor] less than or
equal to 0.95. There is no postulated accident which could cause
reactivity to increase beyond the analyzed conditions in the new
fuel rack.
The radiological consequence analyses [...] performed to support
the installation of replacement steam generators at VCSNS included
the development of source terms which bound fuel enrichments up to
5.0 w/o U235 [Uranium 235] and average discharge burnups up to
65,730 MWD/MTU [megawatt days per metric ton uranium], which bounds
the currently licensed burnup for fuel at VCSNS. These source terms
were used to calculate offsite doses for accidents that are
postulated to result in the release of fission products to the
environment, including the fuel handling accident. In all cases, the
dose results are within 10CFR100 limits.
2. The change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed Technical Specification changes do not involve any
physical changes to the plant or any changes to the method in which
the plant is operated. They do not affect the performance or
qualification of safety related equipment. Therefore the possibility
of a different type of accident or malfunction than previously
considered is not created.
3. The change does not involve a significant reduction in a
margin of safety.
Criticality analyses [...] have been performed for the spent
fuel pool to allow for storage of fuel assemblies with enrichments
up to 5.0 without U-235. The proposed Technical specification
changes include those necessary to maintain Keff less than or equal
to 0.95, including conservative allowances for uncertainties and
biases, when the pool is flooded with unborated water.
The new fuel racks have been previously analyzed [...] for
storage of fuel assemblies with enrichments up to 5.0 w/o U-235. For
the flooded condition Keff does not exceed 0.95 including
conservative allowances for uncertainties and biases. For the
normally dry condition Keff does not exceed 0.98 for the low density
optimum moderation condition. However, the proposed Technical
Specification changes require fuel assemblies with enrichment above
4.0 w/o U-235 to contain integral fuel burnable absorbers such that
the maximum reference fuel [infinite neutron multiplication factor]
is less than or equal to 1.460 in unborated water at 68 deg.F due to
restrictions on spent fuel storage.
Since the proposed changes ensure that the design basis for
preventing criticality in the fuel storage areas is preserved and
since fuel cycle designs will continue to be analyzed with NRC-
approved codes and methods to ensure the design bases for VCSNS are
satisfied, there is no 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: Fairfield County Library,
Garden and Washington Streets, Winnsboro, South Carolina 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: S. Singh Bajwa
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama
Date of amendments request: February 16, 1994
Description of amendments request: The proposed amendment will
change the Technical Specifications to modify the description of fuel
and control rod assemblies in TS 5.3.1, Fuel Assemblies. The change to
the fuel assembly description will permit the limited substitution of
zirconium alloy, zircaloy-4, ZIRLOTM, or stainless steel filler
rods for fuel rods in accordance with the NRC-approved applications of
fuel rod configurations that have been analyzed with NRC-approved
methods. This change will allow timely removal of fuel rods that are
found to be a probable source of future leakage. The change will make
provisions for the loading of lead test assemblies into the reactor
without requiring a specific TS change. This amendment also allows the
use of ZIRLOTM clad fuel as lead test assemblies. The specific
descriptions of the fuel and control rod assemblies contained in the TS
which are restrictive due to the unnecessary details are being deleted.
The change will also make line item improvements in the Technical
Specifications in accordance with Generic Letter 90-02, Supplement 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change to the Technical Specifications allowing
reconstitution will not involve a significant increase in the
probability or consequences of an accident previously evaluated
because it will not result in a change to any of the process
variables that might initiate an accident or affect the radiological
release for an accident. The operating limits will not be changed
and the analysis methods to demonstrate operation within the limits
will remain in accordance with NRC-approved methodology. Other than
the changes to the fuel assemblies, there are no physical changes to
the plant associated with this Technical Specification change. The
consequences of an accident previously evaluated will not be
increased because the safety analysis to be performed for each cycle
will continue to demonstrate compliance with all fuel safety design
bases. The ability to remove potentially leaking fuel rods should
result in a reduction in the radiological consequences of any
transients or accidents.
The probability or consequences of an accident previously
evaluated are not significantly increased with the use of
ZIRLOTM cladding. The VANTAGE 5 fuel assemblies containing
ZIRLOTM clad fuel rods meet the same fuel assembly and fuel rod
design bases as other VANTAGE 5 fuel assemblies. In addition, the 10
CFR 50.46 criteria will be applied to the ZIRLOTM clad fuel
rods. The use of these fuel assemblies will not result in a change
to the proposed Farley VANTAGE 5 reload design and safety analysis
limits. Since the original design criteria are being met, the
ZIRLOTM clad fuel rods will not be an initiator for any new
accident. The ZIRLOTM clad material is similar in chemical
composition and has similar physical and mechanical properties as
that of zircaloy-4. Thus, the cladding integrity is maintained and
the structural integrity of the fuel assembly is not affected. The
ZIRLOTM clad fuel rod improves corrosion performance and
dimensional stability. No concerns have been identified with respect
to the use of an assembly containing a combination of both zircaloy-
4 and selected ZIRLOTM clad fuel rods. Since the dose
predications in the Farley safety analyses are not sensitive to the
fuel rod cladding material used, the radiological consequences of
accidents previously evaluated in the Farley safety analysis remain
valid. Therefore the probability or consequences of an accident
previously evaluated are not significantly increased.
The proposed removal of detailed descriptions of fuel and
control rod assemblies will not involve a significant increase in
the probability or consequences of an accident previously evaluated
because it will not result in a change to any of the process
variables that might initiate an accident. The operating limits will
not be changed and the analysis methods to demonstrate operation
within the limits will remain in accordance with NRC-approved
methodology. The consequences of an accident previously evaluated
will not be increased because the safety analyses to be performed
for each cycle will continue to demonstrate compliance with all fuel
safety design bases.
2. This change to the Technical Specifications allowing
reconstitution will not create the possibility of a new or different
kind of accident from any accident previously evaluated because it
will only affect the assembly configuration and will be limited to
NRC-approved applications of fuel rod configurations. The other
aspects of plant design, operation, limitations and responses to
events will remain unchanged.
The possibility for a new or different kind of accident from any
accident previously evaluated is not created by the use of
ZIRLOTM cladding since the VANTAGE 5 fuel assemblies containing
ZIRLOTM clad fuel rods will satisfy the same design bases as
that used for other VANTAGE 5 fuel assemblies. All design and
performance criteria will continue to be met and no new single
failure mechanisms have been defined. In addition, the use of these
fuel assemblies does not involve any alterations to plant equipment
or procedures that would introduce any new or unique operational
modes or accident precursors. Therefore, the possibility for a new
or different kind of accident previously evaluated is not created.
The removal of detailed descriptions of fuel and control rod
assemblies will not create the possibility of a new or different
kind of accident from any accident previously evaluated because they
will be limited to NRC-approved applications of fuel rod
configurations. The other aspects of plant design, operation,
limitations and responses to events will remain unchanged.
3. The use of zirconium alloy, zircaloy-4, ZIRLOTM, or
stainless steel filler rods in fuel assemblies will not involve a
significant reduction in a margin of safety because analyses using
NRC-approved methods will be performed for each configuration to
demonstrate continued operation within the limits that assure
acceptable plant response to accidents and transients. These
analyses will be performed using NRC-approved methods that have been
approved for application to the fuel configuration.
The margin of safety is not significantly reduced by the use of
ZIRLOTM clad [sic] since the VANTAGE 5 fuel assemblies
containing ZIRLOTM clad fuel rods do not change the proposed
Farley VANTAGE 5 reload design and safety analysis limits. The use
of these fuel assemblies will take into consideration the normal
core operating conditions allowed for in the Technical
Specifications. For each cycle reload core, the fuel assemblies will
be evaluated using NRC Staff-approved reload design methods. This
will include consideration of the core physics analysis peaking
factors and core average linear heat rate effects. Therefore, the
margin of safety as defined in the bases to the Farley Technical
Specifications and VANTAGE 5 Licensing Amendment Request is not
significantly reduced.
The removal of detailed descriptions of fuel assemblies will not
involve a significant reduction in a margin of safety because
analyses using NRC-approved methods will be performed for each
configuration to demonstrate continued operation within the limits
that assure acceptable plant response to accidents and transients.
These analyses will be performed using NRC-approved methods that
have been approved for application to the fuel configuration.
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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: James H. Miller, III, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: S. Singh Bajwa
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: February 7, 1994 TS 93-19
Description of amendment request: The proposed change would revise
Technical Specification 5.3.1 to allow the substitution of filler rods
for fuel rods in fuel assemblies. This would permit the timely removal
of fuel rods that are found to be leaking or are determined to be the
probable source of future leaks.
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:
TVA has evaluated the proposed technical specification change
and has determined that it does not represent a significant hazards
consideration based on criteria established in 10 CFR 50.92(c).
Operation of Sequoyah Nuclear Plant (SQN) in accordance with the
proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The substitution of filler rods will be justified using NRC-
approved methodology. This methodology will demonstrate that the
existing design limits and safety analyses criteria are met.
Therefore, the proposed change does not increase the consequences of
an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change involves the substitution of filler rods for
fuel rods. This substitution requires the utilization of NRC-
approved methodology. This methodology will ensure that the specific
analyses will not cause any new or different kind of accident from
that previously analyzed.
3. Involve a significant reduction in a margin of safety.
The substitution of filler rods for fuel rods would result in
less active fuel in the core. Therefore, the amounts of radiological
effluents that may be released offsite would not increase. The NRC-
approved methodology by which any reanalyses would be performed
already accounts for the affects on grid strength or the mass,
stiffness, and fundamental frequency of the fuel assembly during
seismic and loss-of-cooling accident conditions. Thus, the margin of
safety is not reduced when substituting filler rods for fuel rods.
The NRC 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: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
NuclearPlant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: February 7, 1994 TS 93-11
Description of amendment request: The proposed change would revise
Surveillance Requirement (SR) 4.7.9.i, ``Snubber Service Life
Program,'' to replace the present wording that describes the service
life hydraulic snubber monitoring and evaluation program with that from
the Westinghouse Electric Corporation Standard Technical
Specifications, Revision 4a. This would eliminate the need to perform
an engineering evaluation for drag-force increases of 50 percent or
greater of the previously measured value and substitute a requirement
to establish a monitoring program. This program would require that a
maximum service life for the snubber components be determined and the
monitoring program be established to ensure that the maximum service
life is not exceeded based on test results and failure history. A
proposed change to SR 4.7.9.c would remove the wording that is
inconsistent with Generic Letter 90-09 by removing the term ``if
applicable'' for performance of an as-found functional test and the
requirement related to tests of hydraulic snubbers that have uncovered
fluid ports.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
TVA proposes to delete the current TS requirements to perform an
evaluation of snubber test data when there is a greater than 50
percent increase in drag force for mechanical snubbers. The 50
percent evaluation requirement is considered unnecessary where
snubbers have small drag forces during their previous test. During
subsequent testing, small increases in drag forces (when compared
with the rated load of the snubber) may exceed 50 percent of the
previous test value. The relative change in drag force is small when
compared with the overall rating of the snubber; however, under the
current TS, an engineering evaluation for impending failure will
still be required. Eliminating the current evaluation requirement
from SQN's TSs will reduce the burden associated with performing
unnecessary evaluations. The proposed change is consistent with the
standard TS (Revision 4a), and a Snubber Service Life Program
continues to exist at SQN. Therefore, there is no increase in the
probability or consequences of an accident previously evaluated.
In addition, the testing language associated with Visual
Inspection Performance and Evaluation has been deleted to provide
consistency with Generic Letter 90-09. It should be noted that SQN
tests its hydraulic snubbers in either direction as necessary. This
is more conservative than the present TS requirement; therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The two proposed changes involve deleting a requirement to
perform unnecessary analyses and a potentially nonconservative
testing requirement. These changes do not alter any plant operation,
maintenance requirements, or system design or function. Therefore, a
new or different kind of accident is not created by this proposed
change.
3. Involve a significant reduction in a margin of safety.
The proposed changes to the visual inspection and the service
life section will not modify the plant or revise its mode of
operation or the present safety analysis. The trending criteria to
be utilized provide adequate assurance that snubber impending
failure will be predicted in a timely manner. The deleted sections
will not change the requirement to test and trend data for snubbers
to predict failure; therefore, there is not a reduction in any
margin of safety.
The NRC 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: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: February 8, 1994 TS 93-14
Description of amendment request: The proposed change would revise
the setpoints in Technical Specification Table 3.3-4, ``Engineered
Safety Feature Actuation System Instrumentation,'' for the pressure
switches used to control switchover of the motor-driven Auxiliary
Feedwater pump suction from the normal condensate storage tank supply
to the essential raw cooling water supply. The setpoints would be
changed from the present trip setpoint of 2 psig and allowable value of
1 psig, to new values of 3.21 psig and 2.44 psig, respectively.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The auxiliary feedwater (AFW) system is designed to mitigate the
effects of the design basis accidents and anticipated operational
transients listed below:
A. Loss of normal feedwater
B. Loss of offsite power to station auxiliaries
C. Accidental depressurization in the main steam system
D. Rupture of a main steam line
E. Major rupture of a main feedwater pipe
F. Steam generator tube rupture
G. Small break loss of coolant accident
The AFW system only provides mitigation of the events listed
above and cannot initiate design-basis accident. Therefore, the
proposed change in the low-pressure setpoint of the motor-driven AFW
pump supply line will not result in an increase in the probability
of a previously analyzed accident. In addition, the proposed change
does not affect the overall water supply to the AFW system. Instead,
the proposed change results in a transfer from the condensate
storage tanks (CST) to the essential raw cooling water system at a
slightly higher CST water level, thus enhancing the continuous
supply of water. Therefore, this change will not result in an
increase in the consequences of a previously analyzed accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
As discussed previously, the AFW system provides only mitigation
functions. In addition, the proposed change does not affect the
overall function and operation of the AFW system or its associated
water supplies. Instead, this change will provide additional
assurance of the proper operation of the AFW system. Therefore, the
proposed revision of the low-pressure setpoint of the motor-driven
AFW pump supply line will not create the possibility of a new or
different kind of accident from any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The TS bases for the AFW system require that AFW be available to
ensure that the reactor coolant system (RCS) can be cooled down to
less than 350 degrees Fahrenheit from normal operating conditions in
the event of a total loss of offsite power. In addition, the TS
bases for the CST require that a minimum water volume be available
to maintain the RCS at hot standby condition for two hours with
steam discharge to the atmosphere concurrent with a total loss of
offsite power.
The proposed TS revision does not affect the overall operation
of either the AFW system or the CST. The proposed setpoint revision
does slightly reduce the usable volume of water in the CST. However,
sufficient margin remains to ensure compliance with the bases of the
SQN TSs.
Therefore, the proposed changes to the SQN TSs do not involve a
reduction in the margin of safety.
The NRC 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: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: February 8, 1994 TS 93-14
Description of amendment request: The proposed change would revise
the setpoints in Technical Specification Table 3.3-4, ``Engineered
Safety Feature Actuation System Instrumentation,'' for the pressure
switches used to control switchover of the motor-driven Auxiliary
Feedwater pump suction from the normal condensate storage tank supply
to the essential raw cooling water supply. The setpoints would be
changed from the present trip setpoint of 2 psig and allowable value of
1 psig, to new values of 3.21 psig and 2.44 psig, respectively.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The auxiliary feedwater (AFW) system is designed to mitigate the
effects of the design basis accidents and anticipated operational
transients listed below:
A. Loss of normal feedwater
B. Loss of offsite power to station auxiliaries
C. Accidental depressurization in the main steam system
D. Rupture of a main steam line
E. Major rupture of a main feedwater pipe
F. Steam generator tube rupture
G. Small break loss of coolant accident
The AFW system only provides mitigation of the events listed
above and cannot initiate design-basis accident. Therefore, the
proposed change in the low pressure setpoint of the motor-driven AFW
pump supply line will not result in an increase in the probability
of a previously analyzed accident. In addition, the proposed change
does not affect the overall water supply to the AFW system. Instead,
the proposed change results in a transfer from the condensate
storage tanks (CST) to the essential raw cooling water system at a
slightly higher CST water level, thus enhancing the continuous
supply of water. Therefore, this change will not result in an
increase in the consequences of a previously analyzed accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
As discussed previously, the AFW system provides only mitigation
functions. In addition, the proposed change does not affect the
overall function and operation of the AFW system or its associated
water supplies. Instead, this change will provide additional
assurance of the proper operation of the AFW system. Therefore, the
proposed revision of the low-pressure setpoint of the motor-driven
AFW pump supply line will not create the possibility of a new or
different kind of accident from any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The TS bases for the AFW system require that AFW be available to
ensure that the reactor coolant system (RCS) can be cooled down to
less than 350 degrees Fahrenheit from normal operating conditions in
the event of a total loss of offsite power. In addition, the TS
bases for the CST require that a minimum water volume be available
to maintain the RCS at hot standby condition for two hours with
steam discharge to the atmosphere concurrent with a total loss of
offsite power.
The proposed TS revision does not affect the overall operation
of either the AFW system or the CST. The proposed setpoint revision
does slightly reduce the usable volume of water in the CST. However,
sufficient margin remains to ensure compliance with the bases of the
SQN TSs.
Therefore, the proposed changes to the SQN TSs do not involve a
reduction in the margin of safety.
The NRC 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: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: February 9, 1994 TS 93-21
Description of amendment request: The proposed change would revise
Technical Specification Table 3.3-11, ``Fire Detection Instruments,''
by adding one detector to Fire Zones 184, 185, 186, and 187 for each
Unit. These fire zones are located in the 6.9 kv shutdown board room
corridors in the auxiliary building.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). The operation of Sequoyah Nuclear Plant (SQN) in
accordance with the proposed amendment will not:
1.Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to the fire detection instrumentation adds
two additional cross-zone detectors in each of the Units 1 and 2
6900-volt shutdown board room corridors on Elevation 734 of the
auxiliary building. The additional fire detection instrumentation
provides additional assurance that the fire
1detection instrumentation will operate as required in the event
of a fire. However, neither the fire detection instrumentation nor
the equipment associated with this instrumentation is considered to
be the source of an accident. In addition, this equipment is not
taken credit for in the safety analysis. Therefore, there is no
increase in the probability or consequences of a previously
evaluated accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The fire detection and/or suppression functions affected by this
change enhance fire mitigation functions only and do not result in a
change in plant functions. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any preciously analyzed.
3. Involve a significant reduction in a margin of safety.
The equipment functions affected by the proposed changes are not
assumed for any accident in the SQN safety analysis and are not an
input to the TS margin of safety. Therefore, the proposed change
will not result in a reduction in the margin of safety.
The NRC 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: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: January 31, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.1.1.2 to permit the reduction of
boron concentration of water within the reactor coolant system (RCS),
subject to certain restrictions, when the reactor is in Mode 5 and RCS
flow is less than 2800 gpm. The proposed amendment is related to
Amendment No. 176, which was issued by the NRC on December 8, 1992, and
incorporated a similar revision for Mode 6 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:The Nuclear Regulatory
Commission has provided standards in 10 CFR 50.92(c) for determining
whether a significant hazard exists due to a proposed amendment to an
Operating License for a facility. A proposed amendment involves no
significant hazards if operation of the facility in accordance with the
proposed changes would: (1) Not involve a significant increase in the
probability or consequences of an accident previously evaluated; (2)
Not create the possibility of a new or different kind of accident from
any accident previously evaluated; or (3) Not involve a significant
reduction in a margin of safety. Toledo Edison has reviewed the
proposed change and determined that a significant hazards consideration
does not exist because operation of the Davis-Besse Nuclear Power
Station, Unit Number 1, in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions or assumptions are significantly affected by the proposed
changes. The proposed change to Technical Specification (TS) 3/
4.1.1.2 would revise an exception to make it applicable in Mode 5 as
well as Mode 6. The revised exception would allow water of a lower
boron concentration than the Reactor Coolant System (RCS) to be
added to the RCS, with the flowrate of reactor coolant through the
RCS less than 2800 gpm, provided that the water to be added meets
the requirements of TS 3.1.1.1 (Mode 5) or TS 3.9.1 (Mode 6). TS
3.1.1.1 requires that in Mode 5, the boron concentration of the RCS
be maintained such that the Shutdown Margin shall be less than or
equal to one percent delta k/k. TS 3.9.1 requires that in Mode 6,
the boron concentration of all filled portions of the RCS and the
refueling canal shall be maintained uniform and sufficient to ensure
that the more restrictive of two reactivity conditions is met. If
the RCS meets these reactivity condition requirements, and water is
added to the RCS that also meets the reactivity condition
requirements of TS 3.1.1.1 or TS 3.9.1, then the RCS is assured to
remain in compliance with the reactivity condition requirements. The
possibility that the added water may be of lower boron concentration
than the RCS, therefore, does not significantly increase the
probability of an accident previously evaluated.
The proposed change to TS 3/4.9.8.1 makes TS 3/4.9.8.1 and TS 3/
4.9.8.2 consistent with the current TS 3/4.1.1.2, and is considered
to be administrative in nature.
The proposed changes to TS Bases 3/4.1.1.2 and TS Bases 3/4.9.8
are considered to be administrative in nature.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no accident conditions or
assumptions are affected by the proposed changes. As discussed in
item 1a. above, the proposed revision of the exception to TS 3/
4.1.1.2 will not cause a condition that would result in the RCS not
meeting the requirements of TS 3.1.1.1 or TS 3.9.1, as applicable.
The proposed changes do not alter the source term, containment
isolation, or allowable releases. The proposed changes, therefore,
will not increase the radiological consequences of a previously
evaluated accident. As also discussed in item 1a. above, the
proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases
3.4.9.8 are considered to be administrative in nature.
2a. Not create the possibility of a new kind of accident from
any accident previously evaluated because no new accident initiators
or assumption are introduced by the proposed changes. The proposed
changes do not alter any accident scenarios. As discussed in item
1a. above, the proposed revision of the exception to TS 3/4.1.1.2
will not cause a condition that would result in the RCS not meeting
the requirements of TS 3.1.1.1 or TS 3.9.1. The proposed changes to
TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases 3/4.9.8 are
considered to be administrative in nature. None of the proposed
changes creates the possibility of a new kind of accident.
2b. Not create the possibility of a different kind of accident
from any accident previously evaluated because no different accident
initiators or assumptions are introduced by the proposed changes.
The proposed changes do not alter any accident scenarios. As
discussed in item 1a. above, the proposed revision of the exception
to TS 3/4.1.1.2 will not cause a condition that would result in the
RCS not meeting the requirements of TS 3.1.1.1 or TS 3.9.1. The
proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases
3/4.9.8 are considered to be administrative in nature. None of the
proposed changes creates the possibility of a different kind of
accident from any accident previously evaluated.
3. Not involve a significant reduction in the margin of safety
because the proposed change to TS 3/4.1.1.2, as described above,
will not cause a condition that would result in the RCS not meeting
the requirements of TS 3.1.1.1 or TS 3.9.1. The margin of safety
will be maintained by adhering to the limits specified in these TSs.
The proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1 and TS
Bases 3/4.9.8 are considered to be administrative in nature.
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 Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Project Director: John N. Hannon
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: July 14, 1993
Description of amendment request: The proposed amendment would
modify Sections 3.6 and 4.6 of the Technical Specifications to add
Reactor Coolant System leakage detection requirements to address
Generic Letter 88-01, ``NRC Position on Intergranular Stress Corrosion
Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1.The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed amendment would add a more conservative
requirement into the plant Technical Specifications, in addition to
those that presently exist. Hence, approval of this change will have
no affect on any previously evaluated accident scenario.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No physical changes are being made to the plant and now
new operating techniques or procedures are being proposed. The
proposed amendment would add an additional Limiting Condition for
Operation and an increased Surveillance
Requirement to plant Technical Specifications. Hence, approval
of this change will not create the possibility of a new or different
kind of accident.
3. The proposed amendment will not involve a significant
reduction in a margin of safety. The proposed change adds more
restrictive requirements into the Technical Specifications. Hence,
approval of this change would not reduce the margin of
safety.GI21The Commission has also provided guidance concerning the
application of these standards by providing certain examples (March
6, 1986, 51FR7751). An example of an amendment that is considered
not likely to involve a significant hazards consideration is Example
(ii) which is an additional limitation, restriction or control not
presently included in the Technical Specifications. This proposed
amendment provides for an additional Limiting Condition of Operation
and an increased Surveillance Requirement in the plant Technical
Specifications. Therefore, based on the above, it is determined this
change does not constitute a significant hazards consideration as
defined in 10CFR 50.92(c).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301
Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,
One International Place, Boston, Massachusetts 02110-2624
NRC Project Director: Walter R. Butler
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia
Date of amendment request: December 27, 1993
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the Surry Power Station, Units
No. 1 and No. 2 (SPS-1&2). The proposed changes revise the review
responsibilities of the Station Nuclear Safety and Operating Committee
(SNSOC) and the Management Safety Review Committee (MSRC).
The SPS-1&2 TS address the organization and responsibilities of
both the onsite and offsite review groups: SNSOC and MSRC,
respectively. The responsibilities of the SNSOC include the review of
new procedures and changes to procedures that affect nuclear safety.
The MSRC review responsibilities include the review of safety
evaluations and SNSOC meeting minutes and reports. The extent of these
review activities would be revised by the proposed changes to ensure
the two review groups are focusing on nuclear safety issues and not
spending an unnecessary amount of time on activities of minimal safety
significance. Specifically, the proposed changes would revise the
review responsibilities of SNSOC regarding procedure changes. Rather
than reviewing all procedure changes, SNSOC would only review procedure
changes that require a safety evaluation. The proposed changes also
would revise the review responsibilities of the MSRC. Rather than
reviewing all of the safety evaluations and SNSOC meeting minutes and
reports as presently required by the TS, the MSRC would only review a
representative sample of these documents.
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:
[T]he elimination of the SNSOC review of procedure changes that
do not require a safety evaluation, revising the wording for
approval of procedure changes, and the modification of the MSRC's
duties regarding their review of safety evaluations and SNSOC
meeting minutes and reports will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. As administrative
changes, the proposed Technical Specifications changes have no
direct or indirect effect on accident precursors. No plant
modifications are being implemented and operation of the plant is
unchanged. SNSOC review of new procedures and procedure changes that
require a safety evaluation ensures that activities that could
affect nuclear safety are being properly reviewed. The MSRC's
overview of representative samples of safety evaluations and SNSOC
meeting minutes and reports based on performance ensures these
programs are being properly implemented and nuclear safety is not
being compromised; or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated since physical modifications
are not involved and systems and components will be operated as
before the change. The proposed changes are wholly administrative in
nature and have no impact on plant operations or accident
considerations. These changes modify the scope of SNSOC review of
procedure changes and MSRC's review functions concerning safety
evaluations and SNSOC meeting minutes and reports. Procedure changes
will continue to receive management review in accordance with
administrative procedures, however, only changes that require a
safety evaluation will require SNSOC approval. MSRC review of
representative samples of safety evaluations and SNSOC meeting
minutes and reports based on performance will continue to provide
adequate assurance that nuclear safety is being properly considered;
or
3. Involve a significant reduction in a margin of safety as
defined in the basis of any Technical Specification since the
responsibilities of the SNSOC and MSRC are not addressed by the
existing Technical Specification Bases, nor are review requirements
for procedures. The proposed changes are administrative in nature
and have no impact on, nor were they considered in, existing UFSAR
accident analyses. Safety significant procedure changes, i.e.,
changes that require a safety evaluation to be prepared, will
continue to be reviewed by SNSOC, as will new procedures. Procedure
changes still require cognizant management approval and preparation
of an activity screening to determine whether or not the change
impacts nuclear safety. This ensures activities important to nuclear
safety are being appropriately reviewed. The effectiveness of the
safety evaluation program, and the thoroughness of SNSOC meetings
and reports will be assured through the MSRC's plant overview
function which is based on observed performance.
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: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
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: December 6, 1993
Description of amendment request: The proposed amendment would
modify the testing requirements for the Main Steam Relief Valves
(MSRVs) in the Technical Specifications (TS). Specifically, the
proposed amendment would allow deferral of MSRV Position Indication
Channel Calibration, including the Channel Functional Test that is the
focus of the request, for 24 hours after the plant reaches conditions
that would allow the Channel Functional Test to be conducted under
operating conditions (above 10% rated reactor power). In addition, the
proposed change would extend the current deferral for two related TS,
one for MSRVs and the other for Automatic Depressurization System (ADS)
valves, from 12 hours to the proposed 24 hours after reaching
conditions that would allow the Channel Functional Test to be conducted
under operating conditions.
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 of the
change that would defer Safety/Relief Valve Position Indicators Channel
Calibration, specifically the Channel Functional Test, is presented
below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
No credit is taken for the MSRV position indication in the
initiation or mitigation of any analyzed accident. The inoperability of
valve position indication does not affect the manual or automatic
actuation of the MSRVs. The analysis for inadvertent opening of an MSRV
(FSAR Section 15.1.4) assumes that the alarm function of the MSRV
discharge line temperature sensors and Reactor Pressure Vessel (RPV)
level control systems provide the signals for manual and automatic
system actuation, respectively. Therefore, because previously analyzed
accidents are mitigated without the use of the MSRV position
indication, this change that would allow temporary plant operation in
modes 1 and 2 with inoperable MSRV position indication to allow testing
does not 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 mode of operation of any equipment results from the delay of
the Channel Functional surveillance. The valve position indication
provides information for operator response to previously evaluated
accidents, and does not provide any automatic system actuations that
could initiate an accident or abnormal operating occurrence sequence.
This 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?
The margin of safety involved in this amendment is the time it
takes to identify inadvertent MSRV operation to initiate either
automatic or manual plant response. Since no credit is taken for MSRV
position indication in the WNP-2 safety analyses for initiation of
automatic or operator manual response, the lack of MSRV position
indication for a 24 hour time period will not affect the analyzed time
to identify inadvertent MSRV operation. The proposed change does not,
therefore, affect the margin of safety.
The staff's evaluation of the licensee's analysis of the change
that would allow an increase in the current deferral of Safety/Relief
Valve and Automatic Depressurization System (ADS) valve testing of 12
hours to the proposed 24 hours is presented below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change represents an additional 12 hours in which the
operability of the MSRV position indication would not be verified, and
could be inoperable. No credit is taken for the MSRV position
indication in the initiation or mitigation of any analyzed accident.
The inoperability of valve position indication does not affect the
manual or automatic actuation of the MSRVs as discussed in the
following: The analysis for inadvertent opening of an MSRV (FSAR
Section 15.1.4) assumes the alarm function of the MSRV discharge line
temperature sensors and RPV level control systems for manual and
automatic system actuation, respectively. Thus, the inoperability of
the MRSV position indication for an additional 12 hours would not
affect the probability or consequences of an accident previously
evaluated.
The proposed change represents an additional 12 hours in which the
operability of the ADS valves, which are also the MSRVs, would not be
verified, and could be inoperable. The cause of any potential
inoperability would also be undetermined, and any unidentified failure
mode affects the uncertainties assumed in the probability of
inadvertent MSRV opening, which is an analyzed accident (FSAR Section
15.1.4). In the worst case, the probability of inadvertent MSRV opening
would be increased due to an unidentified problem with the MSRVs. The
additional 12 hours, considering plant operation for approximately 5800
hours of power operation each year (assuming 60 day annual refueling
outage and 80% capacity factor for the remaining time), represents only
an estimated 0.2% increase in time at power without verifying
operability of the ADS valves, which contributes a very small increase
in the probability of an inadvertent MSRV opening due to an
undiscovered fault condition. Performing the surveillance testing using
other possible methods, either removal of the MSRV position indication
from the valve and testing separately, or testing in modes 3, 4, or 5
such that adequate steam back pressure does not exist, introduce other
uncertainties and potential for valve damage that would, using
engineering judgement, create a greater increase in probability of an
inadvertent MSRV opening than the additional 12 hours would contribute.
The proposed change would not, therefore, significantly increase the
probability of an accident previously evaluated.
Regarding the potential effect on the consequences of an
inadvertent MSRV opening, the accident analysis in FSAR Section 15.1.4
analyzes consequences assuming an MSRV sticks open. Thus, increasing
the time that a potential malfunction of a valve may go undetected does
not affect the consequences of an MSRV that is already considered open
as assumed in the accident analysis. Consequently, the increase in time
allowed for conducting the surveillance testing, and therefore the time
that the operability of the ADS-related MSRV is not verified does not
affect the 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?
No new or modified mode of operation of any structure, system, or
component results from delaying verification of the status of
operability of the MSRVs for an additional 12 hours. The proposed
change does 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?
The margin of safety involved in this proposed amendment is the
time, established by the current TS, that operation in modes 1 and 2 is
allowed with the operability of the MSRVs and associated ADS valves
undetermined. The current TS allow 12 hours to conduct testing to
verify operability of these valves. The proposed TS would extend that
time to 24 hours. The additional 12 hours, considering plant operation
for approximately 5800 hours of power operation each year (assuming 60-
day annual refueling outage and 80% capacity factor for the remaining
time), represents only an estimated 0.2% increase in time at power
without verifying operability of the ADS valves, which the staff
considers, by engineering judgement, as a very small reduction in the
margin to safety affected by the proposed change. This change would,
therefore, not involve a significant reduction in the margin of safety
provided by the current TS.
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 Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: January 26, 1994
Description of amendment request: The proposed amendments would
change Technical Specification Section 15.3.0, ``General
Considerations.'' This section specifies the actions which should be
taken for conditions not directly addressed in the action statements of
the Technical Specifications. The changes would provide more
operational flexibility. Changes to the applicable bases and editorial
changes are also proposed.
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 this facility under the proposed Technical
Specifications change will not create a significant increase in the
probability or consequences of an accident previously evaluated.
This proposed change will incorporate requirements contained in
NUREG 1431, Revision 0, ``Westinghouse Owner's Group Improved
Technical Specifications,'' into the Point Beach Technical
Specifications Section 15.3.0, ``General Considerations.'' The
proposed revisions will not remove any existing requirements.
Several new requirements will be added. However, the proposed
revisions will allow a longer period of time for the affected
unit(s) to be placed in hot shutdown should one of the applicable
Limiting Conditions for Operation not be met. This longer time limit
is identical to the time limit specified in NUREG 1431, Revision 0,
and permits the shutdown to proceed in a controlled and orderly
manner that is well within the capabilities of the unit(s), assuming
that only the minimum required equipment is operable. This reduces
thermal stresses on components of the Reactor Coolant System and the
potential for a plant transient that could challenge plant safety
systems. The amount of time to reach cold shutdown would decrease
from 48 hours to 37 hours. The slightly longer time to reach hot
shutdown is more than offset by the shorter time to cold shutdown,
thereby reducing the total amount of time a unit may be operated in
a condition in which a system or component required to mitigate the
consequences of an accident is unavailable, or that is otherwise
prohibited by the specifications.
Should a shutdown of both units be required, 15.3.0.A and
15.3.0.B allow an orderly and sequential shutdown of each unit to
take place. This ensures that the plant operators are not overloaded
during the shutdown process and allows the units shutdowns to
proceed in a controlled and orderly manner. The revised
specifications will clarify the requirements, enhancing the
effectiveness of the Point Beach Technical Specifications. There is
no physical change to the facility, its systems, or its operation.
Thus, an increased probability or consequences of an accident
previously evaluated cannot occur.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This proposed change will incorporate requirements contained in
NUREG 1431, Revision 0, ``Westinghouse Owner's Group Improved
Standard Technical Specifications,'' into the Point Beach Technical
Specifications Section 15.3.0, ``General Considerations.'' The
proposed revisions will not remove any existing requirements.
Several new requirements will be added.
However, the proposed revisions will allow a longer period of
time for the affected unit(s) to be placed in hot shutdown should
one of the applicable Limiting Conditions for Operation not be met.
This longer time limit is identical to the time limit specified in
NUREG 1431, Revision 0, and permits the shutdown to proceed in a
controlled and orderly manner that is well within the capabilities
of the unit(s), assuming that only the minimum required equipment is
operable. This reduces thermal stresses on components of the Reactor
Coolant System and the potential for a plant transient that could
challenge plant safety systems. The amount of time to reach cold
shutdown would decrease from 48 hours to 37 hours. The slightly
longer time to reach hot shutdown is more than offset by the shorter
time to cold shutdown, thereby reducing the total amount of time a
unit may be operated in a condition in which a system or component
required to mitigate the consequences of an accident is unavailable,
or that is otherwise prohibited by the specifications.
Should a shutdown of both units be required, 15.3.0.A and
15.3.0.B allow an orderly and sequential shutdown of each unit to
take place. This ensures that the plant operators are not overloaded
during the shutdown process and allows the units shutdown to proceed
in a controlled and orderly manner. The revised specifications will
clarify the existing specifications and will add some additional
requirements, enhancing the effectiveness of the Point Beach
Technical Specifications. There is no physical change to the
facility, its systems, or its operation. Thus, a new or different
kind of accident cannot occur.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety. This proposed change will incorporate requirements
contained in NUREG 1431, Revision 0, ``Westinghouse Owner's Group
Improved Standard Technical Specifications,'' into the Point Beach
Technical Specifications Section 15.3.0, ``General Considerations.''
The proposed revisions will not remove any existing requirements.
Several new requirements will be added.
However, the proposed revisions will allow a longer period of
time for the affected unit(s) to be placed in hot shutdown should
one of the applicable Limiting Conditions for Operation not be met.
This longer time limit is identical to the time limit specified in
NUREG 1431, Revision 0, and permits the shutdown to proceed in a
controlled and orderly manner that is well within the capabilities
of the unit(s), assuming that only the minimum required equipment is
operable. This reduces thermal stresses on components of the Reactor
Coolant System and the potential for a plant transient that could
challenge plant safety systems. The amount of time to reach cold
shutdown would decrease from 48 hours to 37 hours. The slightly
longer time to reach hot shutdown is more than offset by the shorter
time to cold shutdown, thereby reducing the total amount of time a
unit may be operated in a condition in which a system or component
required to mitigate the consequences of an accident is unavailable,
or that is otherwise prohibited by the specifications.
Should a shutdown of both units be required, 15.3.0.A and
15.3.0.B allow an orderly and sequential shutdown of each unit to
take place. This ensures that the plant operators are not overloaded
during the shutdown process and allows the units shutdowns to
proceed in a controlled and orderly manner. The revised
specifications will clarify the existing specifications and will add
some additional requirements, enhancing the effectiveness of the
Point Beach Technical Specifications. There is no physical change to
the facility, its systems, or its operation. Thus, a significant
reduction in a margin of safety cannot occur.
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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
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.
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 amendments request: November 4, 1993
Description of amendments request: The proposed amendments would
allow the removal of an orifice plate in the containment vent/purge
line to allow greater flow through the line. The restoration of full
flow capability will result in less time required to vent the
containment. A reanalysis of the maximum hypothetical accident, as
currently described in the Updated Final Safety Analysis Report, was
performed to support the requested amendments. The results of the
reanalysis indicate that the consequences of the accident previously
analyzed would be increased. Although the consequences result in an
increase in the fission product release, the total doses are well
within the limits of 10 CFR Part 100, ``Factors to be considered when
evaluating sites.''Date of Publication of Individual Notice in Federal
Register: February 25, 1994 (59 FR 9254)
Expiration date of individual notice: March 28, 1994
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra
Duke Power Company, Docket No. 50-414, Catawba Nuclear Station,
Unit No. 2, York County, South Carolina
Date of amendment request: January 10, 1994
Description of amendment request: The proposed amendment would
change the method of measuring the reactor coolant system flow rate
(Technical Specification 2.0 and 3/4.2) during the 18-month
surveillance for Catawba, Unit 2. Date of publication of individual
notice in Federal Register: March 1, 1994 (59 FR 9785)
Expiration date of individual notice: March 31, 1994
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of application for amendment: February 7, 1994
Brief description of amendment request: The proposed amendment
would allow an increase in reactor coolant temperature in order to
support operation at the rated thermal power of 3565 megawatts thermal
(MWt). The proposed amendment would change reactor protection system
setpoints by increasing the nominal reactor coolant average temperature
from 581.2 deg.F to 586.5 deg.F, changing the axial flux difference
penalties and setpoint uncertainty allowances. The proposed amendment
also increases the maximum indicated reactor coolant system average
temperature from 585.0 deg.F to 590.5 deg.F.
Date of individual notice in Federal Register: February 15, 1994
(59 FR 7269)
Expiration date of individual notice: March 17, 1994
Local Public Document Room location: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
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. 50-318, Calvert
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland
Date of application for amendment: November 1, 1993, as
supplemented on February 1, 1994
Brief description of amendment: The amendment revises the heatup
and cooldown curves which will allow operation beyond the current 12
effective full-power years (EFPY) to approximately 13.8 EFPY. The
increase in this EFPY will allow Unit 2 to operate until its next
refueling outage (RFO-10) in accordance with the requirements of 10 CFR
Part 50, Appendix G.
Date of issuance: March 1, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 162
Facility Operating License No. DPR-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 24, 1993 (58
FR 62150)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 location: Calvert County Library, Prince
Frederick, Maryland 20678.
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 application for amendment: August 27, 1993, as supplemented
November 10, 1993 and February 1, 1994
Brief description of amendment: The amendment revises the Technical
Specifications and allows elimination of the existing reactor coolant's
resistance temperature detector (RTD) bypass manifold system and the
substitution of a new design with RTDs mounted in thermowells that
extend directly into the flow stream of the reactor coolant system.
Date of issuance: February 18, 1994
Effective date: February 18, 1994
Amendment No. 43
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 15, 1993 (58
FR 48379)The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 18, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
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 application for amendment: July 16, 1993, as supplemented
February 17, 1994
Brief description of amendment: The amendment revises the Technical
Specifications (TS) 3/4.2.1, Axial Flux Difference, 3/4.2.2, Heat Flux
Hot Channel Factor (FQ), deletes surveillance requirement 4.2.2.2
and 4.2.2.3, and changes 6.9.1.6, Core Operating Limit Report, and
associated Bases related to the transition from nuclear fuel supplied
by Westinghouse to nuclear fuel supplied by Siemens Power Corporation
(SPC) beginning with Cycle No. 6.
Date of issuance: March 1, 1994
Effective date: March 1, 1994
Amendment No. 44
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41500)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 location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
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 application for amendment: May 15, 1993, as supplemented
February 17 and February 25, 1994
Brief description of amendment: The amendment modifies the SHNPP
Operating License to provide for a one-time exemption from compliance
with License Condition 2.C.(8) which requires periodic emergency diesel
generator engine teardowns for component inspections.
Date of issuance: March 3, 1994
Effective date: March 3, 1994
Amendment No. 45
Facility Operating License No. NPF-63. Amendment revises the
Operating License Condition 2.C.(8) as defined in Attachment 1 to
Operating License NPF-63.
Date of initial notice in Federal Register: June 9, 1993 (58 FR
32378) The February 17, 1994, letter provided supplemental information
and the February 25, 1994, letter modified the May 15, 1993, letter and
requested a one-time exemption. Neither supplemental letter changed the
initial proposed determination of no significant hazards consideration
as published in the Federal Register.The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
March 3, 1994.No significant hazards consideration comments received:
No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
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 application for amendment: July 16, 1993, as supplemented
February 17, 1994
Brief description of amendment: The amendment revises the SHNPP
Technical Specification to incorporate changes to reactor core safety
limits, reactor trip system instrumentation setpoints, power
distribution limits, and shutdown boron concentration control in
support of the transition from nuclear fuel supplied by Westinghouse
Electric Corporation to nuclear fuel supplied by Siemens Power
Corporation and a reactor core safety average temperature reduction
effort.
Date of issuance: March 3, 1994
Effective date: March 3, 1994
Amendment No. 46
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR 50966) The February 17, 1994, letter provided clarifying information
that did not change the initial determination of no significant hazards
consideration as published in the FEDERAL REGISTER.The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated March 3, 1994. No significant hazards consideration comments
received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: August 13, 1993, as
supplemented by letters dated September 15, 1993, September 16, 1993,
December 17, 1993, January 19, 1994, February 11, 1994, and February
24, 1994.
Brief description of amendments: These amendments revise Technical
Specification 3/4.4.5, ``Steam Generators,'' to allow sleeving of
defective steam generator tubes as an alternative to tube plugging. Two
different methods of sleeving are approved for Byron and Braidwood
stations: Westinghouse laser-welded tube sleeving and Babcock and
Wilcox (B&W) kinetic-welded tube sleeving.
Date of issuance: March 4, 1994
Effective date: March 4, 1994
Amendment Nos.: 47, 47, 59, and 59
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: October 27, 1993 (58 FR
57846) The additional information contained in the supplemental letters
dated September 15, 1993, September 16, 1993, December 17, 1993,
January 19, 1994, February 11, 1994, and February 24, 1994, served to
clarify the amendments, were within the scope of the initial notice,
and did not affect the Commission's proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated March 4, 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. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: September 2, 1993, as
supplemented by letters dated January 7, 1994, and February 10, 1994.
Brief description of amendments: The amendments revise the Byron
and Braidwood Technical Specifications (TS) to allow replacement of the
125 volt DC Gould batteries with new 125 volt DC AT&T batteries and
rephrase the specification for their design duty cycle. In addition,
the amendments revise the crosstie loading limitations and crosstie
breaker limitations. The associated TS Bases are also revised to
include the purpose for the crosstie limitations and a discussion of
the design duty cycle requirements.
Date of issuance: March 4, 1994
Effective date: March 4, 1994
Amendment Nos.: 59, 59, 47, and 47
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
4936) The February 10, 1994, supplemental submittal did not change the
original no significant hazards consideration determination. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated March 4, 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. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: December 15, 1989, as
supplemented September 16, 1992.
Brief description of amendments: The amendments revise Technical
Specifications (TSs) 3.0.4, 4.0.3, and 4.0.4 to incorporate guidance
provided by the NRC in Generic Letter 87-09, ``Sections 3.0 and 4.0 of
the Standard Technical Specifications (STS) on the Applicability of
Limiting Conditions for Operation and Surveillance Requirements.''
Date of issuance: February 24, 1994
Effective date: February 24, 1994
Amendment Nos.: 94 and 78
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 12, 1992 (57
FR 53785)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 24, 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: January 28, 1994
Brief description of amendments: The amendments minimize
unnecessary testing for certain instruments in the Reactor Protection
System and the End-of Cycle Recirculation Pump Trip system for LaSalle
County Station, Units 1 and 2 technical specifications.
Date of issuance: February 25, 1994
Effective date: February 25, 1994
Amendment Nos.: 95 and 79
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.Public comments requested as to
proposed no significant hazards consideration: No (59 FR 6062 dated
February 9, 1994). That notice provided an opportunity to submit
comments on the Commission's proposed no significant hazards
consideration determination. No comments have been received. The notice
also provided for an opportunity to request a hearing by March 11,
1994, but indicated that if the Commission makes a final no significant
hazards consideration determination any such hearing would take place
after issuance of the amendment. The Commission's related evaluation of
the amendment, finding of exigent circumstances, and final
determination of no significant hazards consideration is contained in a
Safety Evaluation dated February 25, 1994.
Attorney to licensee: Michael I. Miller, Esquire; Sidley & Austin,
One First National Plaza, Chicago, Illinois 60690.
Local Public Document Room location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: November 4, 1993
Brief description of amendments: The amendment revises the
Technical Specifications surveillance frequency and acceptance criteria
requirements for the steam generator safety valves.
Date of issuance: March 2, 1994
Effective date: March 2, 1994
Amendment Nos.: 154 and 142
Facility Operating License Nos. DPR-39 and DPR-48. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64605)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: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: May 24, 1993
Brief description of amendment: The amendment revises Technical
Specification 4.6.2.1 to allow a one-time relief from the requirement
to perform accelerated Type A integrated leak rate tests (ILRT) after
two consecutive tests fail to meet the acceptance criteria.
Concurrently, the Commission granted a one-time exemption from the
requirement in 10 CFR Part 50, Appendix J, III.A.6.(b) to perform the
Type A containment ILRTs on an accelerated frequency following failure
of two previous Type A tests.
Date of issuance: February 22, 1994
Effective date: February 22, 1994, with full implementation within
45 days.
Amendment No.: 97
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43925)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 22, 1994 No significant hazards
consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: October 19, 1993
Brief description of amendments: The amendments revise both the
surveillance requirements of the Unit 1 Technical Specification (TS)
3.9.D and Unit 2 TS 3/4.8.2, ``Electrical Power Monitoring for Reactor
Protection System,'' to add time delays to the reactor protection
system electrical power monitoring trips.
Date of issuance: February 24, 1994
Effective date: To be implemented within 60 days from the date of
issuance.
Amendment Nos.: 191 and 130
Facility Operating License Nos. DPR-57 and NPF-5. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67846)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 24, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: November 9, 1993
Brief description of amendments: The amendments revise the
Surveillance Requirements of Hatch Unit 1 Technical Specification (TS)
Section 4.9, ``Auxiliary Electrical Systems,'' and Hatch Unit 2 TS
Section 4.8, ``AC Sources - Operating,'' to change the requirements for
diesel generator testing under hot initial conditions.
Date of issuance: February 24, 1994
Effective date: To be implemented within 60 days from the date of
issuance.
Amendment Nos.: 192 and 131
Facility Operating License Nos. DPR-57 and NPF-5. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67846)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 24, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Energy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 8, 1993, as supplemented by
letter dated February 3, 1994.
Brief description of amendment: The amendment revises the River
Bend Station, Unit 1 technical specifications to permit extending
certain surveillance intervals so that testing can be performed during
the refueling outage scheduled to start April 16, 1994, rather than
requiring an earlier shutdown solely to perform the testing.
Date of issuance: February 18, 1994
Effective date: February 18, 1994
Amendment No.: Amendment No. 72
Facility Operating License No. NPF-47: The amendment revised the
technical specifications.
Date of initial notice in Federal Register: January 18, 1994 (59 FR
2630) The February 3, 1994, letter provided clarifying information and
did not change the initial no significant hazards consideration
determination.The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 18, 1994. No
significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: February 1, 1990, as supplemented by
letters dated November 27, 1990, June 5, 1991, November 3, 1992,
November 11, 1992, August 16, 1993, October 22, 1993, November 5, 1993
(two letters), and November 29, 1993.
Brief description of amendments: The amendments consist of changes
to the technical specifications (TS) to change the allowed outage times
(AOTs) and/or surveillance test intervals (STIs) as a result of the
South Texas probabilistic safety assessment (PSA) considered in
conjunction with other information. Ten of the TS changes were based on
changes to the core damage frequency as calculated using the PSA. Five
additional TS changes to the AOTs and STIs were not specifically
modeled in the PSA, but had little or no impact on risk.
Date of issuance: February 17, 1994
Effective date: February 17, 1994, to be implemented within 45 days
of issuance.
Amendment Nos.: Unit 1 - Amendment No. 59; Unit 2 - Amendment No.
47
Facility Operating License Nos. NPF-76 and NPF-80: Amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: March 21, 1990 (55 FR
10535) The supplemental letters provided additional clarifying
information, were within the scope of the original application and did
not change the original proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 17, 1994. No
significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: December 14, 1993
Brief description of amendment: The amendment revises Technical
Specification (TS) 3/4.8.2, ``DC Sources,'' to delete two notes which
indicated that two 125-volt full capacity battery chargers were
required when the Uninterruptible Power Supply was powered by its
backup DC power supply. These notes applied to the Divisions I and II
DC sources during operating and shutdown conditions. The amendment also
revises TS 3/4.8.2 to increase the minimum allowable electrolyte
temperature for the 125-volt batteries from 60 deg.F to 65 deg.F. In
addition, the amendment makes administrative changes to TS 4/3.8.4,
``Electrical Equipment Protective Devices,'' and the TS Bases.
Date of issuance: March 2, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 55
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2868) 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: 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: May 14, 1993
Brief description of amendment: The amendment modifies the
Technical Specifications relating to the spent fuel pool (SFP) by
removal of the cell blockers in Region C, thus increasing by 234 fuel
assemblies the storage capacity of the SFP. To accommodate the
reactivity requirements, the required burnup of fuel in Region C has
been increased and neutron absorbing (poison) rodlets (pins) are
required to be introduced in fuel assemblies not meeting the maximum
burnup requirements for fuel assemblies without rodlets.
Date of issuance: March 1, 1994
Effective date: As of the date of issuance to be implemented
within30 days.
Amendment No.: 172
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 10, 1993 (58 FR
42581) The submittals of November 30, 1993, December 1, 1993, and
January 27, 1994, provided clarifying 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 March 1, 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.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-313,
DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: December 22, 1992 (Reference
LAR 92-09)
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 Section 2.2, ``Reactor Trip System
InstrumentationSetpoints,'' would be revised to change the
Overtemperature Delta-R reactor trip setpoint as a result of a non-
conservativsm in the Westinghouse methodology used to calculate the
f(delta I) penalty function.
Date of issuance: February 24, 1994
Effective date: During 6th refueling outage for Units 1 and 2,
March andSeptember 1994, respectively
Amendment Nos.: 88 and 87
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 3, 1993 (58 FR
7003)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 1974.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: January 10, 1994, as
supplemented February 3, 1994 (Reference LAR 94-01)
Brief description of amendments: The proposed amendments revise the
combined Technical Specifications (TS) for the Diablo Canyon Power
Plant Unit Nos. 1 and 2 to change TS 3/4.3.2, ``Engineered Safety
Features Actuation System Instrumentation,'' and TS 3/4.6.2.3,
``Containment Cooling System.'' Specifically, TS 3/4.3.2, Table 3.3-3,
``Engineered Safety Features Actuation System Instrumentation,'' and
Table 4.3-2, ``Engineered Safety Features Actuation System
Instrumentation Surveillance Requirements,'' are revised to include
Mode 4 applicability requirements for the high-high containment
pressure signal. In addition, TS 3/4.6.2.3 is revised to clarify
acceptable containment fan cooling unit (CFCU) configurations that
satisfy the safety analysis requirements and to clarify the minimum
required component cooling water flow supplied to the CFCU cooling
coils.
Date of issuance: March 2, 1994
Effective date: March 2, 1993
Amendment Nos.: 89 and 88
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 28, 1994 (59 FR
4121)The February 3, 1994 Federal Register submittal provided
clarifying information and did not affect the no significant hazards
consideration determination. 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: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
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: May 21, 1993
Brief description of amendment: The amendment revises the TS
Surveillance Requirements and associated Bases for the emergency diesel
generator (EDG) fuel oil transfer system and the EDG air starting
compressors to clarify that testing of these systems/components can be
conducted either concurrently or independently of the monthly EDG
tests. The changes would also clarify the Bases for EDG fuel quality
testing and make an editorial change.
Date of issuance: February 23, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 205
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36443)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 23, 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.
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: December 22, 1993
Brief description of amendment: The amendment adds Limiting
Conditions for Operation and Surveillance Requirements to Tables
3.12.1, ``Water Spray/Sprinkler Protected Areas,'' and 4.12.1, ``Water
Spray/Sprinkler System Tests,'' and clarifies the associated Bases to
reflect the installation of a new full-area fire suppression system in
the east and west cable tunnels. This new full-area fire suppression
system was installed because the previous sprinkler system did not
provide coverage to some cable trays and the sprinkler head orientation
did not provide full coverage of the cable trays where it was
installed. The amendment also corrects other portions of Tables 3.12.1
and 4.12.1 to ensure consistency with changes made to reflect the east
and west cable tunnel modification.
Date of issuance: February 28, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 206
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 1994 (59 FR
2634) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 28, 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.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: August 30, 1993
Brief description of amendment: The amendment made the following
changes:1. Revised Technical Specification (TS) TS Table 3.8.4.1-1 to
delete Breaker No. 52-263022 which was disconnected and converted to
spare status by a plant design change, and 2. Revised TS 3.11.2.7 to
change radioactivity rate units and associated action statement
reference from HOT STANDBY to read HOT SHUTDOWN, and changed the
reference name of the Offgas Radioactivity Monitor.
Date of issuance: February 28, 1994
Effective date: As of the date of issuance and shall be implemented
within 60 days of the date of issuance.
Amendment No.: 66
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR 50974)The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 28, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket No. 50-311, Salem
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
Date of application for amendment: January 25, 1993
Brief description of amendment: The amendment revised the pressure-
temperature limit curves for heatup, cooldown, hydrostatic tests and
criticality from 10 effective full power years to 15 effective full
power years.
Date of issuance: February 22, 1994
Effective date: February 22, 1994
Amendment No.: 129
Facility Operating License No. DPR-75: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2871)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 22, 1994.No significant hazards
consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Sacramento Municipal Utility District, Docket No. 50-312, Rancho
Seco Nuclear Generating Station, Sacramento County, California
Date of application for amendment: January 19, 1993, and
supplemented May 14, and December 22, 1993. The supplemental
information submitted May 14, and December 22, 1993, did not affect the
proposed no significant hazards consideration determination.
Brief description of amendment: This amendment modifies the nuclear
organization of the Sacramento Municipal Utility District (SMUD) that
will oversee the operation of Rancho Seco at least through the
Custodial SAFSTOR phase of the decommissioning of Rancho Seco.
Date of issuance: February 23, 1994
Effective date: February 23, 1994
3Amendment No.: 121Facility Operating License (Possession Only) No.
DPR-54: This amendment modifies the nuclear organization of SMUD that
will oversee the operation at Rancho Seco at least through the
Custodial SAFSTOR phase of decommissioning.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34092). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 23, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: Central Library, Government
Documents 828 I Street, Sacramento, California 95814.
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: September 19, 1990 as
supplemented on February 26, 1993
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.6.1.3 to specify actions in the event a
containment air lock interlock mechanism becomes inoperable and to
clarify the limitations on the use of an inoperable air lock. The
associated bases were also revised per telecon of February 2, 1994.
Date of issuance: February 23, 1994
Effective date: February 23, 1994
Amendment No.: 56
Facility Operating License No. NPF-58. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 15, 1991 (56 FR
22479) The application for amendment was renoticed on April 14, 1993
(58 FR 19473). The telecon of February 2, 1994, just changed the Bases
section of the TS and did not affect the initial proposed finding of no
significant hazards consideration determination.The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated February 23, 1994.No significant hazards consideration comments
received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: July 16, 1993, as supplemented
November 15, 1993
Brief description of amendments: These amendments implement the
revised 10 CFR Part 20, Standards for Protection Against Radiation
Date of issuance: February 17, 1994
Effective date: February 17, 1994
Amendment Nos.: 178 and 159
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43937)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 17, 1994 No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: March 18, 1993 and December 9,
1993
Brief description of amendments: The amendments revise the NA-1&2
TS which will allow plant personnel to effect repairs to the Rod
Control System in an orderly manner while continuing to ensure that the
control and shutdown banks are capable of performing their designed
safety function.
Date of issuance: March 1, 1994
Effective date: March 1, 1994
Amendment Nos.: 79 and 160
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: April 14, 1993 (58 FR
19492)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 1, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Virginia Electric and Power Company, 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. The December 10, 1993 letter provided clarifying
information within the scope of the original amendment application and
did not change the staff's no significant hazards consideration
determination.
Brief description of amendments: These amendments update the
augmented inspection program for sensitized stainless steel to the
newer Code requirements.
Date of issuance: February 18, 1994
Effective date: February 18, 1994
Amendment Nos.: 187 and 187
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46241) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 18, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
CreekGenerating Station, Coffey County, Kansas
Date of amendment request: February 7, 1994
Brief description of amendment: The changes allow an increase in
reactor coolant temperature in order to support operation at the rated
thermal power of 3565 megawatts thermal (MWt). The amendment changes
reactor protection system overtemperature and overpower delta-
temperature setpoints by increasing the nominal reactor coolant
temperature from 581.2 deg.F to 586.5 deg.F, changing the axial flux
difference penalties, and changing the setpoint uncertainty allowances.
The amendment also increases the maximum indicated reactor coolant
system average temperature of Technical Specification 3/4.2.5, DNB
Parameters, from 585. deg.F to 590.5 deg.F.
Date of issuance:March 3, 1994
Effective date: March 3, 1994, to be implemented within 30 days of
issuance.
Amendment No.: Amendment No. 72
Facility Operating License No. NPF-42. Amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: Yes (59 FR 7269 dated February 15,
1994). That notice provided an opportunity to submit comments on the
Commission's proposed no significant hazards consideration
determination. No comments have been received. The notice also provided
for an opportunity to request a hearing by March 17, 1994, but
indicated that if the Commission makes a final no significant hazards
consideration determination any such hearing would take place after
issuance of the amendment.The Commission's related evaluation of the
amendment, finding of exigent circumstances, and final determination of
no significant hazards consideration is contained in a Safety
Evaluation dated March 3, 1994.Local Public Document Room Locations:
Emporia State University, William Allen White Library, 1200 Commercial
Street, Emporia, Kansas 66801 and Washburn University School of Law
Library, Topeka, Kansas 66621
Dated at Rockville, Maryland, this 7th day March 1994.
For the Nuclear Regulatory Commission.
Steven A. Varga,
Director of Reactor Projects - I/II
[Doc. 94-5971 Filed 03-15-94; 8:45 am]
BILLING CODE 7590-01-F
