[Federal Register Volume 60, Number 70 (Wednesday, April 12, 1995)]
[Notices]
[Pages 18621-18640]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-8845]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 17, 1995, through March 31, 1995. The
last biweekly notice was published on March 29, 1995 (60 FR 16181).
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 [[Page 18622]] no significant hazards
consideration. Under the Commission's regulations in 10 CFR 50.92, this
means that operation of the facility in accordance with the proposed
amendment would not (1) involve a significant increase in the
probability or consequences of an accident previously evaluated; or (2)
create the possibility of a new or different kind of accident from any
accident previously evaluated; or (3) involve a significant reduction
in a margin of safety. The basis for this proposed determination for
each amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By May 12, 1995, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for 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, 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 [[Page 18623]] 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, and at the local public document room for
the particular facility involved.
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 amendment request: March 23, 1994, as supplemented on July
26, 1994, February 15, 1995, and February 28, 1995.
Description of amendment request: In the submittals of March 23 and
July 26, 1994, the licensee requested revisions to the plants'
technical specifications (TSs) to permit the use of a slightly positive
reactor core moderator temperature coefficient (MTC). The February 15,
1995, submittal requested approval to expand the operating limits
report (OLR) to include a cycle specific MTC value and requested
approval to maintain the MTC value within the limits specified in the
OLR. The maximum upper MTC limit would be specified in the TSs. The
February 28, 1995, submittal provided a revised Significant Hazards
Consideration. This supplements the information that was published in
the Federal Register on August 31, 1994 (59 FR 45037).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
An analysis program was pursued by Commonwealth Edison to
justify a positive MTC, reduced reactor coolant system thermal
design flow, and increased steam generator tube plugging levels.
This analysis identified a need for corresponding increases in the
boron concentration levels in the refueling water storage tank
(RWST) and safety injection accumulators to assure subcriticality
requirements are met following a postulated loss-of-coolant accident
(LOCA). The increases in boron concentration are based on the
maximum upper limit of the MTC. The corresponding Technical
Specification changes required as a result of this analysis program
were previously approved by the NRC, including the increases in
boron concentration limits, with the exception of the positive MTC
change. The safety analyses necessary to support this program are
documented in WCAP-13964. The results were reviewed by Commonwealth
Edison and found to be acceptable. All Departure from Nucleate
Boiling Ratio (DNBR) design limits were determined such that there
was a 95 percent probability at a 95 percent confidence level that
DNB would not occur on the most limiting fuel rod for any Condition
I or Condition II event. The present Technical Specification limit
for Nuclear Enthalpy Rise Hot Channel Factor, ... , of less than
1.65 ensures that the DNB design basis stated above would be met,
thus fuel integrity will not be challenged.
The accidents which are sensitive to MTC were analyzed as part
of the overall program and the results were found to be acceptable.
The safety functions of the evaluated systems and components remain
unchanged. The analysis performed using the increased MTC value does
not affect the integrity of the safety related systems and
components such that their function to control radiological
consequences is affected and all fission barriers will remain
intact. The effects on offsite doses have been considered. The
incorporation of a positive MTC, in conjunction with the previously
approved reduction in reactor coolant system thermal design flow
rate and increase in steam generator tube plugging levels, will
result in a small increase in offsite doses; however, the total
doses remain a small fraction of the 10 CFR 100 limits. As such, the
accident analysis acceptance criteria continue to be satisfied.
On a cycle-by-cycle basis, a deterministic evaluation of the
impact on ATWS risk will be performed. An Unfavorable Exposure Time
(UET) will be calculated, where UET is defined as the amount of time
during the operating cycle for which the reactivity feedback is not
sufficient to prevent Reactor Coolant System (RCS) pressure from
exceeding 3200 psig for a given plant configuration. The UET
methodology is consistent with the Westinghouse Owner's Group
methodology presented in WCAP 11992, ``ATWS Rule Administration
Process'' and WCAP 11993, ``Assessment of Compliance with ATWS Rule
Basis for Westinghouse PWRs''. Corrective actions will be taken, as
necessary, to assure a UET of less than 5 percent of cycle length.
The relocation of the cycle-specific core operating limits for
the MTC from the Technical Specifications has no influence or impact
on the probability or consequences of any accident previously
evaluated. Byron and Braidwood Stations will continue to operate
within the cycle-specific MTC limits contained in the OLR. The
proposed amendment will require exactly the same action to be taken
when the OLR limits are exceeded as are required by the current
Technical Specification. Any change to the MTC values in the OLR
will be performed based on NRC-approved methodology as delineated in
Section 6.9.1.9 of the Technical Specifications. Each accident
analysis addressed in the Updated Final Safety Analysis Report
(UFSAR) will be examined with respect to changes in cycle dependent
parameters, which are obtained from application of NRC-approved
reload design methodologies, to ensure that the transient evaluation
of new reloads are bounded by previously accepted analysis. This
examination, which will be performed under the requirements of 10
CFR 50.59, ensures that future reloads will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Therefore, implementation of a positive MTC will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different type of accident from any accident previously
evaluated.
The methodology and manner of plant operation as a result of the
proposed changes is unaffected. Implementation of a positive MTC
does not impact the safe operation of the reactor provided that the
Limiting Conditions for Operation (LCOs) and the associated action
requirements are satisfied. The assumptions do not create any new
failure modes that could adversely impact safety related equipment.
The reload safety limits and LCOs in the plant Technical
Specifications will be evaluated and satisfied for each future
reload core design via the 10 CFR 50.59 process. All DNBR limits
have been satisfied. Currently installed equipment will not be
operated in a manner different than previously designed. No new
credible limiting single failure has been created. No new or
different accidents or failure modes have been identified for any
systems or components important to safety.
The relocation of the cycle specific MTC values to the OLR will
not create the possibility of a new or different type of accident.
No safety related equipment or safety function will be altered as a
result of this proposed change. The cycle specific values are
calculated using NRC-approved methods and submitted to the NRC to
allow the Staff to continue to trend these limits. The Technical
Specifications will continue to require operation within the
analyzed core operating limits and appropriate actions will be
taken, when, or if, the limits are exceeded. [[Page 18624]]
Therefore, there is not a potential for creating the possibility
of a new or different type of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The performance and integrity of the evaluated safety related
systems and components are not affected by the proposed change to
the MTC. The radiological consequences of all previously analyzed
accidents remain within acceptable limits. The proposed change to
the MTC will have no effect on the availability, operability, or
performance of the evaluated safety related systems or components.
The incorporation of a positive MTC, in conjunction with the
previously approved reduction in reactor coolant system thermal
design flow rate and increase in steam generator tube plugging
levels, will result in a small increase in offsite doses; however,
the total doses remain a small fraction of the 10CFR100 limits. The
methodology, discussed in Attachment E, describes the determination
and use of the UET values in the calculation of the Primary Pressure
Relief node for the ATWS event tree to determine an overall ATWS
risk value. The methodology will be used by ComEd to ensure that a
core designed with a positive MTC will not result in an unacceptable
risk to core damage frequency due to an ATWS event. The margin of
safety associated with the licensing basis safety analysis is not
significantly reduced by the proposed changes. All acceptance
criteria for the specific UFSAR Chapter 15 safety analyses (non-LOCA
and LOCA) have been satisfactorily evaluated and verified using NRC
approved methodologies.
The margin of safety is not affected by the relocation of the
cycle specific MTC limits from the Technical Specifications. The
proposed amendment continues to require operation within the core
limits as determined by the NRC-approved reload design and safety
analysis methodologies. Appropriate actions will be taken, when, or
if, limits are exceeded.
The development of the MTC limits for future reloads will
continue to conform to those methods described in the NRC-approved
documentation. In addition, each future reload will involve a 10 CFR
50.59 safety review to assure that operation of the unit within the
cycle specific limits will not involve a reduction in the margin of
safety as defined in the basis for any Technical Specification.
Therefore, there is no significant reduction in the margin of
safety as defined in the bases of any Technical Specification.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
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.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: October 20, 1992
Description of amendment request: The proposed amendment would
comply with the requirements of Amendment 135 to the Palisades
Operating License, dated February 11, 1991, which included a change to
Technical Specification 5.3.1a, Primary Coolant System. The safety
evaluation for Amendment 135 included a requirement that changes to
Section 4.2 of the Palisades Final Safety Analysis Report (FSAR) be
made through a formal amendment process. The proposed FSAR change is a
result of the steam generator replacement project and includes the
following: (1) deletion of a design load since this was not treated as
a necessary design condition in the new steam generators; (2) a change
in the feedwater temperature from 70 deg.F to 40 deg.F, since this
assumption was changed in the analysis for the replacement steam
generators; and (3) editorial changes.
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 following summary supports the finding that the proposed
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The probability of an accident previously evaluated in the FSAR
will not be increased by deleting the design load change of 15% per
minute or decreasing the minimum feedwater temperature from 70 deg.F
to 40 deg.F. There is no design requirement that the plant be
capable of 15% per minute load changes. No accident has as an
initial condition a 15% per minute load change taking place, and
since this FSAR change is the result of the replacement steam
generators design, no accident probabilities are increased. The
40 deg.F feedwater temperature affects the steam generators, but
nothing else is affected in the primary coolant system (PCS). The
replacement steam generators have been shown by the design analysis
report to be able to withstand the same number of cycles of the
addition of 40 deg.F water as the old steam generators could with
70 deg.F water.
The consequences of an accident previously evaluated in the FSAR
are not increased by either of these two changes. Deleting the
design load rate of 15% per minute deals with normal plant operation
and would not affect the course of a Chapter 14 event since none of
the Chapter 14 events involve power level changes with respect to
the steam generators. Also, reducing the maximum design load change
rate is a conservative change.
Lowering the feedwater temperature could increase the
consequences of the main steam line break (MSLB) accident by
increasing the likelihood of a return to power event caused by
increased core cooling; however, the current FSAR analysis in
Section 14.14 used 32 deg.F as the auxiliary feedwater temperature
and thus bounds [the] 40 deg.F [temperature].
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The possibility of a new or different type of accident is not
created by these FSAR changes. By deleting the 15% per minute load
change rate from the FSAR, the operation of the plant is unaffected
because the 5% per minute limit on load rate change is more
limiting. There is no license requirement to be able to change power
at 15% per minute except as described in the proposed FSAR deletion.
Furthermore, FSAR Section 4.3.7.2 states that the pressurizer
heaters cannot be uncovered by the outward surge of water following
load increases; a 10% step increase and 15% ramp increase. FSAR
Section 1.2.4.9.a states that the nuclear steam supply system (NSSS)
is capable of a ramp change from 15% to 100% power at 5% per minute,
and at a greater rate over smaller load changes up to a step change
of 10%.
Another consideration is that the analysis for the original
steam generators was not as detailed or exact as the analysis for
the replacement steam generators. The thermal analysis section of
the original steam generator design analysis report states for the
three power change cases, 5% per minute, 15% per minute and a 10%
step change, that ''... the transient thermal effects of the power
changes are small and [negligible]. The situations of significance
are due to cycling between steady state conditions at different
power levels.'' Thus, the rate of change was not a consideration in
the original design analysis. The replacement steam generator
analysis calculated the transient temperature changes with respect
to time, so the rate of change was considered. Therefore, the
replacement steam generator analysis is more accurate, but does not
consider a 15% per minute rate change. The original steam generators
were not designed for 15% per minute power changes but could
withstand power increases from 50% to 100% [a total of] 15,000 times
without considering the rate of power change.
Reducing the analyzed feedwater temperature from 70 deg.F to
40 deg.F does not change the possibility of whether another type of
accident or malfunction can occur since the steam generator is
analyzed for this.
3. Involve a significant reduction in a margin of safety.
The margin of safety as defined by plant licensing basis is not
reduced due to the replacement steam generators not being analyzed
for a 15% per minute power ramp [[Page 18625]] because the 15% per
minute ramp rate was not a licensing basis of the plant design. The
original plant Safety Evaluation Report does not mention the design
power ramp rates. The basis for Technical Specification 3.1.2 states
that all components are designed to withstand the effects of cyclic
loads due to primary coolant system temperature and pressure changes
induced by load changes, trips, and start-ups and shutdowns. FSAR
Section 4.2.2 is referenced. The change of eliminating the analyzed
ability to make 15% per minute power changes does not reduce the
margin of safety because:
a. the plant is not operated in a manner wherein 15% per minute
power increases are made. Rapid power decreases during emergency
conditions are not covered by this analysis since they are not
controlled to 15% per minute but should be considered analyzed by
the 500 trips or 10% step change analysis and,
b. the original steam generator did not use the ramp rate in the
analysis and,
c. a 15% per minute power change from 50% to 100% power is a
fairly benign change for the steam generator with respect to
pressure and temperature changes as compared to heatups and
cooldowns because the total changes are small.
The only requirement from the NRC with respect to the number and
type of loads is contained in Section II of the NRC Standard Review
Plan (SRP) 3.9.1 which states ''...The section of the applicant's
SAR which pertains to transients will be acceptable if the transient
conditions selected for equipment fatigue evaluation are based upon
a conservative estimate of the magnitude and frequency of the
temperature and pressure conditions resulting from those
transients.'' ''... Transients and resulting loads and load
combinations with appropriate specified design and service limits
must provide a complete basis for design of the reactor coolant
pressure boundary for all conditions and events expected over the
service lifetime of the plant.''
In the intervening years between design of the original steam
generators and the replacement steam generators, Combustion
Engineering (ABB-CE) decided that a 15% per minute power ramp rate
was beyond what was necessary and expected to occur. This position
was acceptable to the NRC since ABB-CE letter CPC-90-170, dated
October 24, 1990, states that the replacement steam generators are
identical in design to the Palo Verde (Arizona Public Service) steam
generators. (The ABB-CE letter was concerned with the stress
analysis for steam line breaks, therefore, the reference to being
identical was with respect to that stress analysis.)
The change in feedwater temperature from 70 deg.F to 40 deg.F
maintains the margin of safety because the replacement steam
generators have been shown by the design analysis report to be able
to withstand the same number of cycles of the addition of 40 deg.F
water as the old steam generators could 70 deg.F water.
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: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: Cynthia A. Carpenter, Acting
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: September 13, 1993
Description of amendment request: The proposed amendment would
relocate audit frequencies of Section 6.5.2.8 of the Technical
Specifications to the Quality Assurance Program in Section 17.2 of the
Updated Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change to relocate the audit program frequency
requirements to the Quality Assurance Program does not:
(1) involve a significant increase in the probability or
consequences of an accident previously evaluated,
This change is administrative in nature and does not impact the
operation of the plant or the plant's response to an accident.
Because it will allow more flexibility in assigning resources to
assess weak or declining performance areas, the plant safety
performance will be improved.
(2) create the possibility of a new or different kind of
accident from any accident previously evaluated,
This change is administrative in nature and does not affect the
operation or design of the plant; therefore, there is no change in
the possibility of a new or different kind of accident from any
accident previously evaluated.
(3) involve a significant reduction in a margin of safety.
This change is administrative in nature and does not affect the
operation of the plant; therefore, there is no change in the margin
of safety. Relocating the audit program frequency requirements to
the Quality Assurance program will allow a more dynamic and
responsive audit program. Audits will be able to be scheduled more
effectively based on performance and the status of related
activities. This should result in a more effective audit program
that will contribute to an improvement in safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226
NRC Project Director: Cynthia A. Carpenter, Acting
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368,
ArkansasNuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County,
Arkansas
Date of amendment request: August 30, 1994, with supplement dated
January 19, 1995.
Description of amendment request: The proposed amendment changes
requirements related to the site parimeter security system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, excerpts of this analysis are presented below:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated
The accident mitigation features of the plant are not affected
by the proposed change. This change provides an equivalent level of
protection as required by 10CFR73.55(c)(4), does not significantly
decrease the effectiveness of the security program, and is adequate
for preventing an unacceptable risk to public health and safety.
Ample protection against a design basis security threat continues to
be provided. Therefore, the probability or consequences of an
accident previously evaluated is not significantly increased.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from Any Previously Evaluated
This change clarifies the existing configuration of the
protected area barrier at the ANO intake structure. New systems,
modes of equipment operation, failure modes, or other plant
perturbations are not introduced by this change. Therefore, the
possibility of a new or different kind of accident from amy
previously evaluated is not created.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety
This change clarifies the existing configuration of the
protected area barrier at the ANO intake structure. The proposed
change does not alter a safety limit, a limiting condition of
operation, or a surveillance requirement on equipment to operate the
plant. Adequate physical protection of the plant is maintained.
Therefore, the margin of safety is not significantly
reduced. [[Page 18626]]
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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center
(DAEC), Linn County, Iowa
Date of amendment request: March 1, 1995
Description of amendment request: The proposed License Amendment
would revise Technical Specification (TS) Sections 4.5 and 4.8 of the
DAEC TS to reflect the changes to pump and valve testing criteria. The
proposed amendment changes the testing frequency for certain pumps and
valves in the Low Pressure Coolant Injection subsystem; Core Spray
subsystems; and the Residual Heat Removal Service Water, High Pressure
Coolant Injection, Emergency Service Water, and River Water Supply
systems. The frequency would change from testing every three months to
that specified by DAEC ASME Section XI Inservice Testing (IST) program.
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 affected pumps and valves in Sections 4.5 and 4.8 will
continue to be tested in accordance with ASME Section XI OM-6 and
OM-10. The affected pumps and valves will continue to function as
before and this change will not result in a decrease in their
availability to mitigate the consequences of certain accidents and
transients. The proposed amendment will not affect the consequences
of these accidents and transients. Therefore, the
proposed amendment does not involve a change in the probability
or consequences of an accident previously evaluated.
(2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated. The safety functions of the affected pumps and
valves will remain unchanged. This amendment will result in no
physical changes to the affected pumps, valves or systems.
Consequently, the proposed license amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) The proposed amendment will not reduce the margin of safety.
The actual operation of the affected pumps and valves will remain
unchanged. Testing in accordance with ASME Section XI OM-6 and OM-10
will continue to provide assurance that degradation in tested
components will be detected and addressed.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis
& Bockius, 1800 M Street, N.W., Washington, DC 20036-5869NRC Acting
Project Director: Gail H. Marcus
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: January 24, 1995, as supplemented March
22, and March 29, 1995.
Description of amendment request: The amendment request would
revise the Technical Specification Section 3.2.3.1.a and Table 2.2-1 to
decrease the acceptance criterion for measured reactor coolant system
(RCS) flow rate from 387,480 gallons per minute (gpm) to 371,920 gpm.
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 (SHC), which is presented below:
...The proposed changes do not involve an SHC because the
changes would not:
1. Involve a Significant Increase in the Probability or
Consequence of an Accident Previously Evaluated.
An evaluation of the 4% decrease in the RCS total flow rate
limit has shown that the change does not significantly impact the
design basis analyses. Therefore, the change will not increase the
consequences of an accident previously evaluated.
There are no actual plant changes that will result from this
technical specification change. Instead, the technical specification
requirement for minimum total RCS flow rate is being changed to
provide operational benefit without compromising safety. Since there
are no plant changes, there is no effect on the probability of
occurrence of previously evaluated accidents.
The change will have a negligible impact on the small break loss
of coolant accident (LOCA) and large break LOCA analyses. The PCT
[peak cladding temperature] acceptance criteria will continue to be
met with the assumption of a 4% reduction in RCS flow rate.
For the steam generator tube rupture event, both the FSAR [Final
Safety Analysis Report] offsite dose analysis and the margin of
steam generator (SG) overfill were evaluated. It was determined that
the 4% reduction in RCS flow rate will not adversely affect the
offsite doses or the margin to SG overfill and, therefore, the FSAR
conclusions remain unchanged.
In the evaluation of non-LOCA transients, the DNB [departure
from nucleate boiling] is the most affected parameter due to a
change in flow rate. It was concluded that the 4% reduction in RCS
flow was acceptable and there was margin to the DNB limit.
It is concluded that there is sufficient margin to the system
pressure, PCT and DNB limits to offset the effect of the 4% flow
rate decrease and the calculated radiological releases associated
with the analysis are not affected. Therefore, there is no effect on
the consequences of previously evaluated accidents.
2. Create the Possibility of a New or Different Kind of Accident
from any Previously Evaluated.
The low loop flow trip setpoint specified in Technical
Specification Table 2.2-1 is set as a fraction of total flow. The
flow fraction is not being changed and no hardware changes are
required due to the reduction in minimum flow. Also, the reduction
in minimum flow will not change the operation of any plant equipment
and it does not modify plant operation.
Therefore, the reduction in minimum flow does not introduce any
new failure modes or malfunctions and it does not create the
potential for a new unanalyzed accident.
3. Involve a Significant Reduction in the Margin of Safety.
The proposed 4% decrease in the technical specification limit
for total RCS flow rate will not adversely affect the results of the
FSAR accident analysis, and it is concluded that this change is
safe. The change does not adversely affect any equipment credited in
the safety analysis, and it does not affect the probability of
occurrence of any plant accident. Also, the change has a negligible
impact on the PCT, and it does not increase the offsite doses or
decrease the DNB below its acceptance limit.
Therefore, the change does not have any significant impact on
the protective boundaries, and there is no reduction in the margin
of safety as specified in the technical specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574
[[Page 18627]] New London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: March 1, 1995
Description of amendment request: The proposed amendment to the
technical specifications (TS) would make administrative changes to TS
2.5, 2.8, 2.11, 3.2, and 3.10 and, in accordance with Generic Letter
(GL) 93-07, ``Modification of the Technical Specification
Administrative Control Requirements for Emergency and Security Plans,''
to TS 5.5 and 5.8.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed revisions to Technical Specifications (TS) 5.5 and
5.8 are administrative in nature and follow the guidance of Generic
Letter (GL) 93-07. The review and audit functions of the site
security and emergency plans and procedures will be retained in a
manner that fully satisfies regulatory requirements. Therefore, the
proposed revisions do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed revision to TS 2.5 will still require backup water
for the emergency feedwater storage tank to be available. However,
several other available sources of water are preferred over river
water, such as, the water plant demineralized water system and the
outside condensate storage tank. Therefore, the proposed revision
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed deletion of TS 2.8(8) pertaining to fuel handling
cranes, deletion of TS 2.11 pertaining to overhead cranes in the
Containment and Auxiliary Buildings, and deletion of statements in
the bases of TS 2.8 pertaining to crane interlocks does not involve
a significant increase in the probability or consequences of an
accident previously evaluated. Specifications 2.8(8), 2.11 and the
deleted statements in the bases of Specification 2.8 need not be
retained in the TS based upon Criteria 1 through 4 of the ``Final
Policy Statement on Technical Specifications Improvements for
Nuclear Power Reactors,'' dated July 22, 1993 (58 FR 39132).
Controls and limitations for the operation and testing of these
cranes and interlocks will be incorporated into the Updated Safety
Analysis Report (USAR). The requirements of TS 2.8(8) and
restrictions of TS 2.11 are currently contained in Station
procedures to ensure that the handling of fuel assemblies, control
element assemblies (CEAs) and heavy loads is accomplished safely and
effectively. These revisions make the FCS Technical Specifications
more similar to Standard Technical Specifications (STS), which do
not contain requirements or restrictions concerning the operation of
fuel handling cranes or overhead cranes.
The revision proposed for TS 3.2, Table 3-5, Item 1 will make
its surveillance frequency identical to the frequency specified in
STS 3.1.5.7. The proposed frequency will require testing CEA drop
times prior to reactor criticality after each removal of the reactor
vessel closure head, which is the most appropriate time to perform
the surveillance. The proposed frequency will ensure that the CEAs
drop into the core within the time specified in the safety analysis
and, therefore, does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed deletion of TS 3.2, Table 3-5, Item 5, which
currently requires testing refueling system interlocks prior to the
refueling outage does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Table 3-5, Item 5, does not need to be retained in the TS based upon
Criteria 1 through 4 of the ``Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors,'' dated July
22, 1993. Controls and limitations for testing the refueling system
interlocks will be incorporated into the USAR. The requirements for
testing refueling system interlocks are already contained in Station
procedures. This revision makes the FCS Technical Specifications
more similar to STS, which do not contain requirements or
restrictions pertaining to testing refueling system interlocks.
The proposed revision to TS 3.2, Table 3-5, Item 10, ensures
consistent use of terminology among the frequencies specified in
Table 3-5. The proposed revision clarifies the wording and
introduces additional operational flexibility such that the
surveillance could be performed before 720 hours of system
operation, if warranted by plant conditions or beneficial to plant
operation. Therefore, the proposed revision does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The remaining TS revisions are administrative in nature in that
they correct references, titles, misspelling(s), and page numbers,
or revise wording to be consistent with defined intervals within the
TS. Therefore, they do not increase the probability or consequences
of an accident previously evaluated. None of the proposed TS
revisions will impact the function or method of operation of plant
systems, structures, or components.
(2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed revisions to TS 5.5 and 5.8 which delete the review
and/or audit of the emergency, site security and safeguards
contingency plans and implementing procedures from the TS are
administrative in nature and in accordance with the guidance of GL
93-07. The proposed revisions will not affect the operation of any
system, structure, or component and therefore do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed revision to TS 2.5 will still require a backup
supply of water for the emergency feedwater storage tank to be
available. However, several other available sources of water are
preferred over river water, such as, the water plant demineralized
water system and the outside condensate storage tank. Therefore, the
proposed revision does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed deletion of TS 2.8(8) pertaining to fuel handling
cranes, deletion of TS 2.11 pertaining to overhead cranes in the
Containment and Auxiliary Buildings and deletion of statements in
the bases of TS 2.8 pertaining to crane interlocks does not create
the possibility of a new or different kind of accident from any
accident previously evaluated. Specifications 2.8(8), 2.11 and the
deleted statements in the bases of Specification 2.8 need not be
retained in the TS based upon Criteria 1 through 4 of the ``Final
Policy Statement on Technical Specifications Improvements for
Nuclear Power Reactors,'' dated July 22, 1993.
The requirements of TS 2.8(8) and restrictions of TS 2.11 are
currently contained in Station procedures to ensure that the
handling of fuel assemblies, CEAs and heavy loads is accomplished
safely and effectively. These revisions make the FCS Technical
Specifications more similar to STS, which do not contain
requirements or restrictions concerning the operation of fuel
handling cranes or overhead cranes.
The proposed revision to TS 3.2, Table 3-5, Item 1, is an
administrative revision to the frequency of CEA drop time testing.
The proposed frequency is the most appropriate time to perform the
surveillance to ensure that the CEAs drop into the core within the
time specified in safety analysis and is identical to the frequency
specified in STS 3.1.5.7. Therefore, the proposed revision does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
The proposed deletion of TS 3.2, Table 3-5, Item 5, which
currently requires testing the refueling system interlocks prior to
the refueling outage, does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Table 3-5, Item 5, does not need to be retained in the TS based upon
Criteria 1 through 4 of the ``Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors,'' dated July
22, 1993. The requirements for testing refueling
[[Page 18628]] system interlocks are currently contained in Station
procedures. This revision makes the FCS Technical Specifications
more similar to STS, which do not contain requirements or
restrictions pertaining to testing refueling system interlocks.
The proposed revision to TS 3.2, Table 3-5, Item 10, ensures
consistent use of terminology among the frequencies specified in
Table 3-5. The proposed revision clarifies the wording and
introduces additional operational flexibility such that the
surveillance could be performed before 720 hours of system
operation, if warranted by plant conditions or beneficial to plant
operation. Therefore, the proposed revision does not create the
possibility of a new or different kind of accident from any
previously evaluated.
The remaining TS revisions are administrative in nature in that
they correct references, titles, misspelling(s), and page numbers,
or revise wording to be consistent with defined intervals within the
TS. Therefore, they do not create the possibility of a new or
different kind of accident.
(3) The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed revisions to TS 5.5 and 5.8 concerning the review
and/or audit of the emergency, site security and safeguards
contingency plans and implementing procedures do not involve a
significant reduction in a margin of safety. The audit and review
processes are administrative functions which will be retained
outside the TS in a manner that fully satisfies regulatory
requirements.
Removing the requirement of TS 2.5 that Missouri River water
from the fire water system shall be available to provide a backup
water supply to the emergency feedwater storage tank improves
operational flexibility without reducing any safety margins. Better
sources of backup water are available to replenish the emergency
feedwater storage tank. Although deleted from TS 2.5, the fire water
system is still required to be available to meet the requirements of
paragraph 3.F of the FCS Operating License. Therefore, the proposed
revision does not involve a significant reduction in a margin of
safety.
The proposed deletion of TS 2.8(8) pertaining to fuel handling
cranes, deletion of TS 2.11 pertaining to overhead cranes in the
Containment and Auxiliary Buildings and deletion of statements in
the bases of TS 2.8 pertaining to crane interlocks does not involve
a significant reduction in a margin of safety. Specifications
2.8(8), 2.11 and the deleted statements in the bases of
Specification 2.8 do not need to be retained in the TS based upon
Criteria 1 through 4 of the ``Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors,'' dated July
22, 1993.
The requirements of Specification 2.8(8) and restrictions of
Specification 2.11 are currently contained in Station procedures to
ensure that the handling of fuel assemblies, CEAs and heavy loads is
accomplished safely and effectively. These revisions make the FCS
Technical Specifications more similar to STS, which do not contain
requirements or restrictions concerning the operation of fuel
handling cranes or overhead cranes.
The proposed revision to TS 3.2, Table 3-5, Item 1, is an
administrative revision to the frequency of CEA drop time testing.
The proposed frequency is the most appropriate time to perform the
surveillance to ensure that the CEAs drop into the core within the
time specified in the safety analysis and is identical to the
frequency specified in STS 3.1.5.7. Therefore, the proposed revision
does not involve a significant reduction in a margin of safety.
The proposed deletion of TS 3.2, Table 3-5, Item 5, which
currently requires testing the refueling system interlocks prior to
the refueling outage does not involve a significant reduction in a
margin of safety. Table 3-5, Item 5, does not need to be retained in
the TS based upon Criteria 1 through 4 of the ``Final Policy
Statement on Technical Specifications Improvements for Nuclear Power
Reactors,'' dated July 22, 1993. The requirements for testing
refueling system interlocks are currently contained in Station
procedures. This revision makes the FCS Technical Specifications
more similar to STS, which do not contain requirements or
restrictions pertaining to testing refueling system interlocks.
The proposed revision to TS 3.2, Table 3-5, Item 10, ensures
consistent use of terminology among the frequencies specified in
Table 3-5. The proposed revision clarifies the wording and
introduces additional operational flexibility such that the
surveillance could be performed before 720 hours of system operation
if warranted by plant conditions or beneficial to plant operation.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
The remaining TS revisions are administrative in nature in that
they correct references, titles, misspelling(s), and page numbers,
or revise wording to be consistent with defined intervals within the
TS. Therefore, 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875
Connecticut Avenue, NW., Washington, DC 20009-5728
NRC Project Director: William H. Bateman
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: March 6, 1995
Description of amendments request: The proposed amendment would
relocate the seismic and meteorological monitoring instrumentation from
the Technical Specifications to the Final Safety Analysis Report in
accordance with the ``Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors,'' dated July
22, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change relocates information from the TS to the
FSAR and has no impact on physical plant operation or configuration.
The continued capability of the seismic and meteorological
instrumentation to perform its intended function will be ensured
through controlled change processes governed by 10 CFR 50.59.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The sole function of the seismic and meteorological
monitoring instrumentation is to record data. The proposed change
will not involve any design change or modification to the plant. The
proposed change will not alter the operation of the plant or the
manner in which it is operated. Any subsequent change to the Seismic
and Meteorological Monitoring Instrumentation requirements will
undergo a review in accordance with the criteria of 10 CFR 50.59 to
endure that the change does not involve an unreviewed safety
question.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed change will relocate Seismic and Meteorological
Monitoring Instrumentation requirements from the TS to licensee
controlled documents subject to the criteria of 10 CFR 50.59. The
proposed change will have no adverse impact on any protective
boundary or safety limit.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial
[[Page 18629]] Library, 212 W. Burdeshaw Street, Post Office Box 1369,
Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: William H. Bateman
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: November 15, 1994; superseded March 7,
1995 (TS 94-12).
Description of amendment request: The proposed change would remove
the frequency for each of the audits specified in the administrative
controls section of the technical specifications (TS), except those
related to the fire protection system. The requirements to perform the
audits would be retained, but the frequency for their performance would
be controlled by a requirement to be added to the Nuclear Quality
Assurance Plan. This would require that the audits listed in the TS
(except those related to the fire protection system) be performed on a
biennial frequency. In addition, the proposed change would remove the
requirement to perform site Radiological Emergency Plan, Physical
Security Plan, and the Safeguard Contingency Plan reviews and audits
from the TS, since these requirements presently exist in their
respective Plans.
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 standards used to arrive at a determination that a Technical
Specification change request involves no significant hazards
consideration are included in the Commission's regulations, 10 CFR
50.92, which states that no significant hazards considerations are
involved if the operation of the facility in accordance with the
proposed amendment would not: (1) involve a significant increase in
the probability or consequences of an accident previously evaluated;
or (2) create the possibility of a new or different kind of accident
from any accident previously evaluated; or (3) involve a significant
reduction in a margin of safety. Each standard is addressed as
follows:
1. Operation of the facility in accordance with the proposed
technical specifications would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The likelihood that an accident will occur is neither increased
or decreased by the Technical Specification change which only
affects review and audit frequencies. This Technical Specification
change will not impact the function or method of operation of plant
equipment. Thus, there is not a significant increase in the
probability of a previously analyzed accident due to this change. No
systems, equipment, or components are affected by the proposed
changes. Thus, the consequences of a malfunction of equipment
important to safety previously evaluated in the FSAR are not
increased by this change.
The proposed change only affects review and audit frequencies.
As such, the proposed change has no impact on accident initiators or
plant equipment, and thus, does not affect the probabilities or
consequences of an accident.
Therefore, we conclude that this change does not significantly
increase the probabilities or consequences of an accident.
2. Operation of the facility in accordance with the proposed
technical specifications would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve changes to the physical
plant or operations. Since program audits do not contribute to
accident initiation, a change related to audit functions cannot
produce a new accident scenario or produce a new type of equipment
malfunction. Also, this change does not alter any existing accident
scenarios. The proposed change does not affect equipment or its
operation, and, thus, does not create the possibility of a new or
different kind of accident. Therefore, the proposed change does not
create the possibility of a new or different kind of accident.
3. Operation of the facility in accordance with the proposed
technical specifications would not involve a significant reduction
in a margin of safety.
The proposed change concerning conduct of reviews and audits
does not directly affect plant equipment or operation. Safety limits
and limiting safety system settings are not affected by this
proposed change.
Therefore, use of the proposed Technical Specification would not
involve any reduction in the margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: March 2, 1995
Description of amendment request: The proposed changes would revise
Technical Specification 4.6.1.2.a to reference the testing requirements
of 10 CFR Part 50, Appendix J, and to state that the Nuclear Regulatory
Commission-approved exemptions to the applicable regulatory
requirements are permitted.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A discussion of these standards as they relate to this ...
amendment request follows.
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed change ... revises the North Anna Units 1 and 2
Technical Specification Surveillance Requirement 4.6.1.2.a to
reference the testing frequency requirements of 10 CFR 50 Appendix J
and to state that NRC approved exemptions to the applicable
regulatory requirements are permitted. The current Technical
Specification requires Type A tests be conducted in accordance with
Appendix J to 10 CFR 50. The proposed administrative change simply
includes the statement ``as modified by NRC-approved exemptions.''
No new requirements are added, nor are any existing requirements
deleted. Any specific changes to the requirements of Appendix J will
require a submittal from Virginia Electric and Power Company under
10 CFR 50.12 and subsequent review and approval by the NRC prior to
implementation. The proposed change is stated generically to avoid
the need for further Technical Specification changes if different
exemptions are approved in the future.
The proposed change, in itself, does not affect reactor
operations or accident analyses and has no radiological
consequences. The change provides clarification so that future
Technical Specifications changes will not be necessary to correspond
to applicable NRC-approved exemptions from the requirements of
Appendix J. This exemption request is consistent with the intent of
the regulation.
Therefore, this proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed Technical Specification amendment for Units 1 and 2
provides clarification to a specification that paraphrases a
codified requirement.
Since the ... proposed Technical Specifications change would not
change the [[Page 18630]] design, configuration, or method of
operation of the plant, the changes would not create the possibility
of a new or different kind of accident from any previously
evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed North Anna Units 1 and 2 Technical Specifications
change is administrative and clarifies the relationship between the
requirements of Technical Specification Surveillance Requirement
4.6.1.2.a, Appendix J, and any approved exemptions to Appendix J. It
does not, in itself, change a Safety Limit or a Limiting Condition
for Operation. The NRC will directly approve any proposed change or
exemption to Appendix J prior to implementation.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: David B. Matthews
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 amendment request: November 10, 1994
Description of amendment request: The proposed amendment request
will clarify the surveillance requirements for the reactor protection
and the engineered safeguards system instrumentation and actuation
logic.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of Surry Power Station in accordance with the proposed
Technical Specifications change will not:
1. Involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated.
The proposed change to clarify the surveillance requirements for
the Reactor Protection and Engineered Safeguards Systems
instrumentation and actuation logic has no impact on the probability
of an accident occurrence. The instrumentation and actuation logic
will continue to be operated in the same manner. The actual test
frequency is not changing. Rather, surveillance requirements are
being clarified to represent the actual testing and the licensing
and design bases. Testing of these instruments and actuation logic
are presently design limited and would otherwise require using
temporary modifications to complete the testing. Since the testing
is not changing, the clarification of the actual testing does not
contribute to the probability of any previously analyzed accident.
The Reactor Protection and Engineered Safeguards Systems
instrumentation and actuation logic will be operated in the same
manner and the system operability requirements are not being
altered. Therefore, the consequences of any design basis accident
are not being increased by the proposed change to clarify the
surveillance test requirements for the Reactor Protection and
Engineered Safeguards System instrumentation and actuation logic.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
There are no plant modifications or changes in methods of plant
operation introduced by this change in the clarification of the
testing for the Reactor Protection and Engineered Safeguards Systems
instrumentation and actuation logic. The plant is not being operated
or tested in a different manner due to the proposed change.
Therefore, no new accidents or accident precursors are generated by
the proposed change to clarify the surveillance test requirements.
Clarifying the surveillance test requirements to represent the
original licensing design basis and test conditions does not create
the possibility of a new or different accident than previously
analyzed.
3.Involve a significant reduction in a margin of safety.
Clarification of the testing for the Reactor Protection and
Engineered Safeguards Systems instrumentation and actuation logic
surveillance requirements does not affect the margin of safety in
that the operability requirements for these safety systems remain
unchanged. The existing testing is performed in accordance with
plant design and licensing basis and provides adequate indication of
the operability of the affected instrumentation or actuation logic.
The Reactor Protection and Engineered Safeguards Systems
instrumentation and actuation logic are fully tested on a refueling
cycle basis which includes complete operation of each relay and end
device. Therefore, the margin of safety is not altered by the
proposed clarification of the testing for the Reactor Protection and
Engineered Safeguards Systems instrumentation and actuation logic.
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: David B. Matthews
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 amendment request: November 22, 1994
Description of amendment request: The proposed amendment request
would delete unnecessary descriptive phrases regarding the number of
cells in the station and emergency diesel generator batteries.
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 deletion of the descriptive references regarding the number
of cells in the station and emergency diesel generator batteries is
an administrative change and therefore does not:
1. Involve an increase in the probability of occurrence or
consequences of an accident previously evaluated.
The proposed change to delete the descriptive references
associated with the station and emergency diesel generator batteries
(60 cell or 56 cell, respectively) has no impact on the probability
of an accident occurrence. The change is administrative in nature
and therefore does not affect the operation of the units. The
batteries will continue to be operated in the same manner as before
the change with operability based on design voltage and capacity
requirements necessary to ensure safety functions can be performed.
Prescribed surveillance testing will continue to ensure the
operability of individual battery cells. Consequently, the proposed
change does not contribute to the probability of occurrence or
consequences of any design basis accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
This is an administrative change to delete the descriptive
references associated with the station and emergency diesel
generator batteries. There are no plant modifications being
implemented by the proposed change and plant operations are not
being changed. Provided the required design voltage and capacity are
maintained, the batteries remain fully operable and capable of
performing their intended safety functions. Individual battery cell
surveillance requirements remain unchanged. Therefore, no new
accidents or accident precursors are created by the proposed change.
3. Involve a reduction in a margin of safety as defined in the
Technical Specifications. [[Page 18631]]
The proposed administrative change to delete the descriptive
references associated with the station and emergency diesel
generator batteries (60 cell or 56 cell, respectively) is
administrative in nature. Provided the required design voltage and
capacity are maintained, the batteries remain fully operable and
capable of performing their intended safety functions as assumed in
the safety analyses. Individual battery cell surveillance
requirements remain unchanged. Therefore, the analyzed margin of
safety is not reduced by the proposed change.
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: David B. Matthews
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 amendment request: January 24, 1995
Description of amendment request: The proposed amendment request
would increase the current Technical Specification pressurizer safety
valve lift setpoint acceptance criterion from plus or minus 1% as-found
and plus or minus 1% as-left to plus or minus 3% as-found and plus or
minus 1% as-left.
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 Technical Specifications change does not involve a
significant hazards consideration because operation of Surry Units 1
and 2 in accordance with this change would not:
a. involve a significant increase in the probability or
consequences of an accident previously evaluated. Affected safety-
related parameters were analyzed for a change to Surry Units 1 and 2
Technical Specification 3.1.A.3.b. It was determined that the
primary and secondary side overpressure safety limits would not be
exceeded in the most limiting overpressure transient (Loss of Load,
Locker Rotor, and Rod Withdrawal events) with the pressurizer safety
valve lift setpoint acceptance criterion increased to [plus or
minus] 3%. The DNBR [departure from nucleate boiling ratio] results
of transients impacted by the setpoint acceptance criterion increase
are not affected by the proposed change. The increased setpoint
acceptance criterion will not result in an inadvertent opening of
the pressurizer safety valves. Since the proposed change involves no
alterations to the physical plant, the probability of occurrence of
an accident or malfunction of equipment important to safety
previously evaluated is not increased.
b. create the possibility of a new or different kind of accident
from any accident previously identified. The proposed change to
Surry Units 1 and 2 Technical Specification 3.1.A.3.b does not
involve any alterations to the physical plant which would introduce
any new or unique operational modes or accident precursors. Only the
allowable tolerance about the existing setpoint will be changed.
c. involve a significant reduction in a margin of safety. It was
determined that the most limiting overpressure transients do not
result in maximum pressures in excess of the primary and secondary
side overpressure limits. The DNBR results of affected transients
are not made more limiting by the proposed setpoint tolerance
increase. Therefore, the margin of safety is unchanged by the
proposed increase in the safety valve setpoint acceptance criterion.
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: David B. Matthews
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: March 21, 1995
Description of amendment request: The amendment would revise
Surveillance Requirement 4.6.2.1.d for the containment spray system to
change the surveillance interval for the performance of the air or
smoke test through the containment spray header from once per 5 years
to once per 10 years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed reduced testing frequency of the Containment Spray
System nozzles does not change the way the system is operated or the
Containment Spray System's operability requirements. The proposed
change to the surveillance frequency of safety equipment has no
impact on the probability of an accident occurrence nor can it
create a new or different type of accident. NUREG-1366 concluded
that the corrosion of stainless steel piping is negligible during
the extended surveillance interval. Since the Containment Spray
System is maintained dry there is no additional mechanism that could
cause blockage of the spray nozzles. Thus, the nozzles in the
Containment Spray System will remain operable during the ten year
surveillance interval to mitigate the consequence of an accident
previously evaluated. No clogging or blockage of the nozzles in the
Containment Spray System has been discovered during the performance
of the five year surveillance tests. Therefore, the testing of the
Containment Spray System[']s nozzles at the proposed reduced
frequency will not increase the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed reduced frequency testing of the Containment Spray
System nozzles does not change the way the Containment Spray System
is operated. The reduced frequency of testing of the spray nozzles
does not change plant operation or system readiness. The reduced
frequency testing of the Containment Spray System nozzles does not
generate any new accident precursors. Therefore, the possibility of
a new or different kind of accident from any accident previously
evaluated is not created by the proposed changes in surveillance
frequency of the Containment Spray System nozzles.
3. The proposed change does not involve a significant reduction
in the margin of safety.
Reduced testing of the Containment Spray System nozzles does not
change the way the system is operated or the Containment Spray
System's operability requirements. NUREG-1366 concluded that the
corrosion of stainless steel piping is negligible during the
extended surveillance interval. Since the Containment Spray System
is maintained dry there is no additional mechanism that could cause
blockage of the Containment Spray System nozzles. Thus, the proposed
reduced testing frequency is adequate to ensure spray nozzle
operability. The surveillance requirements do not affect the margin
of safety in the operability requirements of the Containment Spray
System remains unaltered. The existing safety analysis remains
bounding. Therefore no margins of safety are adversely affected by
this proposed change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
[[Page 18632]] satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
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
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: March 24, 1995
Description of amendment request: The proposed amendment would add
a new action statement to Technical Specification 3.5.1 which would
provide a 72-hour allowed outage time (AOT) for one accumulator to be
inoperable because its boron concentration did not meet the 2300-2500
parts per million (ppm) band. The amendment would also change the
current allowed outage time for other reasons of inoperability from 1
hour to 24 hours.
Changes to the surveillance requirements are also proposed to
incorporate the guidance of Generic Letter 93-05, ``Line-Item Technical
Specifications Improvements to Reduce Surveillance Requirements for
Testing During Operation.'' These proposed changes would base the
operability of the accumulator on the contained water volume and cover
pressure and would not require verification of the boron concentration
after an accumulator volume increase, provided the source of the makeup
water is the refueling water storage tank.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed change does not involve a significant Increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The overall protection system performance will remain within the
bounds of the accident analysis documented in Chapter 15 of the
Updated Safety Analysis Report [USAR], WCAP-1096-P, and WCAP-11883
since no hardware changes are proposed.
The safety injection accumulators are credited in Section 15.6.5
of the Updated Safety Analysis Report for large and small break LOCA
[loss-of-coolant accident]. There will be no effect on these
analyses, or any other accident analysis, since the analysis
assumptions are unaffected and remain the same as discussed in
Section 15.6.5. Design basis accidents are not assumed to occur
during allowed outage times covered by the Technical Specifications.
As such, the ECCS [emergency core cooling system] Evaluation Model
equipment availability assumptions made in Section 15.6.5 remain
valid.
The safety injection accumulators will continue to function in a
manner consistent with the above analysis assumptions and the plant
design basis. As such, there will be no degradation in the
performance of nor an increase in the number of challenges to
equipment assumed to function during an accident situation.
The proposed technical specifications changes do not involve any
hardware changes nor do they affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, ESF [engineered safety features] actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs. Therefore, these changes will not increase the probability
of an accident or malfunction.
The corresponding increase in CDF [core damage frequency] due to
the proposed change to increase the AOT of the accumulators from one
hour to 24 hours is insignificant. Pursuant to the guidance in
Section 3.5 of NSAC-125, the proposed increase in AOT does not
``degrade below the design basis the performance of a safety system
assumed to function in the accident analysis,'' nor does it
``increase challenges to safety systems assumed to function in the
accident analysis such that safety system performance is degraded
below the design basis without compensating effects.'' Therefore, it
is concluded that these changes do not increase the probability of
occurrence of a malfunction of equipment important to safety.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This change is administrative in nature and does not involve any
change to the installed plant systems or the overall operating
philosophy of WCGS [Wolf Creek Generating Station].
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these proposed changes. There will be no adverse effect or
challenges imposed on any safety-related system as a result of these
changes. Therefore, the possibility of a new or different type of
accident is not created.
There are no changes which would cause the malfunction of
safety-related equipment, assumed to be operable in the accident
analyses, as a result of the proposed technical specification
changes. No new mode failure has been created and no new equipment
performance burdens are imposed. Therefore, the possibility of a new
or different malfunction of safety-related equipment is not created.
(3) The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change does not involve an significant reduction in
a margin of safety. There will be no change to the Departure from
Nucleate Boiling Ratio (DNBR) Correlation Limit, the design DNBR
limits, or the safety analysis DNBR limits discussed in Bases
Section 2.1.1.
As discussed previously, the performance of the accumulators
will remain within the assumptions used in the large and small break
LOCA analyses, as presented in USAR Section 15.6.5. Also, there will
be no effect on the manner in which safety limits or limiting safety
system settings are determined nor will there be any effect on those
plant systems necessary to assure the accomplishment of protection
functions.
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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
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. [[Page 18633]]
Baltimore Gas and Electric Company, Docket No. 50-318, Calvert
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland
Date of amendment request: February 24, 1995
Brief description of amendments: The proposed amendment would
revise the Calvert Cliffs, Unit No. 2, Technical Specifications (TSs).
Specifically, TS 4.G.1.2 would reference 10 CFR Part 50, Appendix J,
directly, and any approved exemptions to the Type A testing frequency
requirements, rather than paraphrase the regulation. The proposed
wording is consistent with that used in NUREG-1432, ``Standard
Technical Specifications - Combustion Engineering Plants,'' dated
September 1992.Date of publication of individual notice in Federal
Register: March 8, 1995 (60 FR 12789)
Expiration date of individual notice: April 7, 1995
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: February 23, 1995, as supplemented March
21, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specifications 3.8.2.1 and 3.8.3.1 to allow
installation of a modification to replace the battery, main and tie
breakers in response to an Electrical Distribution Systems Functional
Inspection, conducted by the NRC in July 1991. The existing breaker
arrangement could result in a trip of both the battery and main
breakers if a fault occurs on one of the 125 VDC panelboards. The
licensee committed to have these breakers replaced in 1995 with a
better coordinated design to eliminate the concern.Date of publication
of individual notice in Federal Register: March 8, 1995 (60 FR 12791)
Expiration date of individual notice: April 7, 1995
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County,
Texas
Date of amendment request: March 1, 1995
Description of amendment request: The proposed amendment would
modify the steam generator tube plugging criteria in Technical
Specification 3/4.4.5, Steam Generators, and the allowable leakage for
Unit 1 in Technical Specification 3/4.4.6.2, Operational Leakage, and
the associated Bases.Date of individual notice in the Federal Register:
March 13, 1995 (60 FR 13478)
Expiration date of individual notice: April 12, 1995
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County,
Texas
Date of amendment request: March 1, 1995
Description of amendment request: The proposed amendment would
change Technical Specification 3/4.4.5, Steam Generators, and the
associated Bases to allow the use of an alternate plugging criteria
(known in the industry as F*) on steam generator tubes that are
defective or degraded within certain areas within the tubesheet. Date
of individual notice in the Federal Register: March 13, 1995 (60 FR
13481)
Expiration date of individual notice: April 12, 1995
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 amendment request: March 9, 1994
Description of amendment request: The proposed amendment would
revise the Nine Mile Point Nuclear Station, Unit 2, Technical
Specifications (TSs). Specifically, TS 4.6.1.2.a would be modified to
allow the second Primary Containment Integrated Leakage Rate Test (Type
A) to be performed at the fifth refueling outage (RF-05) or 72 months
after the first Type A test instead of the fourth refueling outage (RF-
04) as currently scheduled.
Date of publication of individual notice in Federal Register: March
23, 1995 (60 FR 15310)
Expiration date of individual notice: April 24, 1995
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Rochester, New York
Date of application for amendment: March 13, 1995
Brief description of amendment: The proposed amendment would revise
Ginna Station Technical Specification (TS) 4.4.2.4.a to replace
specific leakage testing frequencies for containment isolation valves.
This TS change will support a proposed Exemption to Title 10 of the
Code of Federal Regulations (10 CFR) Part 50, Appendix J, Section
III.D.3, requested under separate cover to exempt Type C testing of
certain valves during a 1995 refueling outage.
Date of publication of individual notice in Federal Register: March
22, 1995 (60 FR 15167)
Expiration date of individual notice: April 21, 1995
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
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 [[Page 18634]] under the special circumstances
provision in 10 CFR 51.12(b) and has made a determination based on that
assessment, it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station,Plymouth County, MassachusettsDate of application for
amendment: November 22, 1994
Brief description of amendment: The amendment revises the allowable
leak rate for the main steam isolation valves from the current 11.5
standard cubic feet per hour (scfh) for each valve, to a maximum
combined main steam line leak rate of 46 scfh.
Date of issuance: March 22, 1995
Effective date: March 22, 1995
Amendment No.: 160
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 1995 (60 FR
3671) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 22, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station,Plymouth County, Massachusetts
Date of application for amendment: September 6, 1994, as
supplemented February 15, 1995.
Brief description of amendment: This amendment revises Technical
Specifications (TSs) 3.7.B.1.a, 3.7.B.1.c, 3.7.B.1.e, 3.7.B.2.a, and
3.7.B.2.c and adds Sections 3.7.B.1.f and 3.7.B.2.e. The additional
section requires both trains of standby gas treatment and control room
high efficiency air filtration system to be operable for the initiation
of fuel movement. In the event either train becomes inoperable, the
other train must be demonstrated to be operable within 2 hours and fuel
handling operations may continue for 7 days with one train inoperable.
Additionally, this change allows one train to be defined as operable
without its associated emergency power supply, provided one source of
normal power (startup transformer or unit auxiliary power) is
available.
Date of issuance: March 22, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 161
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53837) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 22, 1995. No significant
hazards consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station,Plymouth County, Massachusetts
Date of application for amendment: September 6, 1994
Brief description of amendment: This amendment would reduce the
Reactor Pressure Setpoint at which the shutdown cooling system
automatically isolates. This setpoint also isolates the low pressure
coolant injection valves when the shutdown cooling system is in
operation.
Date of issuance: March 27, 1995
Effective date: To be implemented within 30 days following restart
from refueling outage 10
Amendment No.: 162
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53837) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 27, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
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 application for amendments: October 28, 1994, as
supplemented February 16, 1995.
Brief description of amendments: The proposed change will revise TS
requirements to increase the surveillance test intervals and the
allowable out of service times or instruments of the reactor protection
system, isolation actuation system, emergency core cooling system
actuation system, control rod withdrawal block system, control room
emergency ventilation system, anticipated transient without scram,
recirculation pump trip (RPT), end-of-cycle RPT, and the reactor core
isolation cooling actuation system.
Date of issuance: March 30, 1995Effective date: March 30, 1995
Amendment Nos.: 175 and 206
Facility Operating License Nos. DPR-71 and DPR-62. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 7, 1994 (59 FR
63114) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 30, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North CarolinaDate of application for amendment: October 24, 1994,
as supplemented December 6, 1994.
Brief description of amendment: The amendment allows the relocation
of TS 3/4.3.4, Turbine Overspeed Protection and associated Bases to be
consistent with the new Standard Technical Specifications for
Westinghouse plants.
Date of issuance: March 22, 1995
Effective date: March 22, 1995
Amendment No. 55
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: November 23, 1994 (59
FR 60379) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 22, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina
27605. [[Page 18635]]
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: December 22, 1992
Brief description of amendments: These amendments add new
requirements to the Technical Specifications (TS) to ensure that an
Essential Service Water system (SX) pump and crossover path are
available from a shutdown unit to serve as backup to an operating unit.
In addition, a new TS is added to require the unit crosstie to be open,
or capable of being opened, from the Main Control Room, whenever
either, or both units are in an operating mode (MODE 1, 2, 3, or 4).
Date of issuance: March 20, 1995
Effective date: March 20, 1995
Amendment Nos.: 71, 71, 62, and 62
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 3, 1993 (58 FR
6994) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 20, 1995. 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.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
PointNuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: September 19, 1994
Brief description of amendment: The amendment would revise
Technical Specification Section 4.4.A.3, Frequency of Containment
Integrated Leakage Rate Test, to reference 10 CFR Part 50, Appendix J,
as modified by approved exemptions, directly.
Date of issuance: March 17, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment No.: 181
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8744) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 17, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
PennsylvaniaDate of application for amendments: April 23, 1990, as
supplemented January 21, 1992 and March 17, 1995.
Brief description of amendments: These amendments revise the
Appendix A Technical Specifications (TSs) for Unit 1 and Unit 2 by (a)
deleting TS Table 3.6-1, ``Containment Penetrations,'' (b) rewording TS
Definition 1.8, ``Containment Integrity,'' and TSs 3.6.1.1, 3.6.1.2,
3.6.3.1, and 3.9.4 relating to containment integrity, containment
leakage, containment isolation valves, and containment building
penetrations respectively to account for the deletion of TS Table 3.6-
1, and (c) correcting terminology by replacing the word ``door'' with
``hatch'' in TS 3.9.4.a.
The Unit 1 amendment also modifies TS Table 3.3-5, ``Engineered
Safety Features Response Times,'' by changing the feedwater isolation
response time to reflect total isolation times for the main feedwater
regulating valve and bypass feedwater regulating valve. Minor editorial
changes were also incorporated in TS Table 3.3-5.
Date of issuance: March 28, 1995
Effective date: March 28, 1995
Amendment Nos.: 185 and 66
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 27, 1990 (55 FR
26283), as supplemented April 1, 1992 (57 FR 11107) The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated March 28, 1995. No significant hazards consideration
comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: June 22, 1994
Brief description of amendment: The amendment changes Technical
Specification (TS) Sections 1.6, 3.2.A, 3.9.f.5 and 4.2.A which specify
the Shutdown Margin (SDM) requirements that ensure the reactor can be
made subcritical and can be maintained sufficiently subcritical to
preclude inadvertent criticality in any core condition. The amendment
also includes a definition of Shutdown Margin, TS Section 1.45.
Administrative changes to TS Sections 1.7 and 3.2.b.2(b) are also
included to simplify definitions and eliminate unnecessary notes and
references.
Date of Issuance: March 21, 1995Effective date: As of the date of
issuance to be implemented within60 days
Amendment No.: 178
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37072) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated March 21, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: February 14, 1995
Brief description of amendment: The amendment revises Technical
Specification 3.8.2, ``AC Sources-Shutdown;'' 3.8.5, ``DC Sources-
Shutdown;'' and 3.8.8, ``Inverters-Shutdown.'' The changes revise the
operability requirements for the Division 3 diesel generator and the
Division 3 and 4 batteries, battery chargers and inverters to apply
only when the high pressure core spray system is required to be
operable.
Date of issuance: March 21, 1995
Effective date: March 21, 1995
Amendment No.: 99
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 17, 1995 (60
FR 9412) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 21, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois
61727. [[Page 18636]]
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: June 30, 1994, as supplemented
March 7, 1995
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.2.7.1 to add 8 check valves to Table 3.2.7.1.
These valves were installed to add additional protection of the low
pressure Core Spray system from the high pressure Reactor Coolant
system. Including the valves in the TSs will assure that the proper
surveillance testing is done to maintain a high reliability for the
valves to protect the Core Spray system.
Date of issuance: March 20, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 154
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39593) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 20, 1995.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: April 22, 1994
Brief description of amendment: The amendment deletes the
operability and surveillance requirements of the condenser air ejector
radiation monitor from the Millstone Unit 2 Technical Specification
Tables 3.3-12 and 4.3-12.
Date of issuance: March 27, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment No.: 186
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27058) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 27, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: March 31, 1994 and August 5,
1994
Brief description of amendment: This amendment revises: Technical
Specification (TS) 3.8.1.1.b.2 which maintains diesel operability for a
48-hour period when the fuel storage system of one or more diesel
generators contains less than a 7-day supply of fuel: TS 4.8.1.1.2.h.8
by deletion and replacement with surveillance requirement 4.8.1.1.2.k.1
which permits the 24-hour diesel generator endurance run to be
performed in any operational condition; establish surveillance
requirement 4.8.1.1.2.k.2 which allows the hot restart test to be
conducted not only after surveillance requirement 4.8.1.1.2.k.1, but
also after the diesel generator has operated between 4300 kw and 4400
kw for one hour or after any time the diesel generator operating
temperature has stabilized; revise TS 3.8.1.1 to eliminate the
requirements to start the Emergency Diesel Generator (EDG) with an
inoperable offsite circuit(s) of AC electrical power; add a provision
that eliminates required testing of remaining EDGs when one EDG is
inoperable due to an inoperable support system or an independently
testable component with no potential for common mode failure for the
remaining EDGs. In addition, if testing of the EDGs is required, the
surveillance will be performed within 16 hours instead of 24 hours as
currently specified; delete the requirement to perform a Loss of
Offsite Power (LOOP) test (Surveillance Requirement 4.8.1.1.2.h.b)
following the 24-hour EDG endurance run test in its place, a hot
restart test (no LOOP load sequencing) will be established.
Date of issuance: March 30, 1995
Effective date: March 30,1995
Amendment No.: 72
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29630) and October 12, 1994 (59 FR 51625) The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
March 30, 1995. No significant hazards consideration comments received:
No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
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 application for amendment: October 29, 1993, as
supplemented on March 11, 1994, May 18, 1994, September 20, 1994, and
October 20, 1994.
Brief description of amendment: The amendment changes Operating
License NPF-12 to delete License Conditions 2.C.13, 2.C.14, and 2.C.32.
Date of issuance: March 29, 1995
Effective date: March 29, 1995
Amendment No.: 123
Facility Operating License No. NPF-12. Amendment revises the
operating license.
Date of initial notice in Federal Register: February 16, 1994 (59
FR 7698) and April 28, 1994 (59 FR 22012), as corrected June 30, 1994
(59 FR 33795). The May 18, 1994, September 20, 1994, and October 20,
1994, submittals provided supplemental and 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 29, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: September 30, 1993, as
supplemented by letters dated November 16, 1993, January 18, 1995, and
February 2, 1995.
Brief description of amendments: These amendments revised the
technical specifications to (1) divide item 7 of Tables 3.3-3, 3.3-4,
3.3-5, and 4.3-2 into item 7a that addresses the existing loss-of-
voltage (LOV) function and item 7b that separately addresses the
degraded grid voltage (DGV) function; (2) add footnote (d) to Table
3.3-3 to indicate that the DGV actuation relay logic is applicable in
Modes 1, 2, 3, and 4 when the diesel generator circuit breaker is open;
(3) replace the reference to Figure 3.3-1 in item 7a of Tables 3.3-4
and 3.3-5 with definite voltage and time values; (4) add note 9 to
Table 3.3-5 to explain the response [[Page 18637]] time for an LOV
signal; and (5) delete Figure 3.3-1, ``Degraded Bus Voltage Trip
Setting,'' and the reference to this figure from Table 3.3-4.
Date of issuance: March 17, 1995
Effective date: Unit 2, as of the date of completion of the
currrent refueling outage and must be fully implemented before the
plant returns to power; Unit 3, as of the date of the completion of its
next refueling outage and must be fully implemented before the plant
returns to power.
Amendment Nos.: Unit 2 - Amendment No. 118; Unit 3 - Amendment No.
107
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59755). The additional information contained in the November 16,
1993, January 18, 1995 and February 2, 1995, letters was clarifying in
nature, within the scope of the initial notice and did not affect the
NRC staff's proposed no significant hazards consideration
determination.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 17, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
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: January 9, 1995
Brief description of amendments: The amendments change the
Technical Specifications to implement recommended changes from Generic
Letter (GL) 93-05, ``Line Item Technical Specification Improvements to
Reduce Surveillance Requirements for Testing During Power Operation,''
dated September 27, 1993. Specifically, the amendments implement TS
changes corresponding to the following GL 93-05 line-item improvement
issues and numbers: Control Rod Movement Test for Pressurized Water
Reactors (4.2.1); Radiation Monitors (5.14); Surveillance of Boron
Concentration in the Accumulator/Safety Injection/Core Flood Tank
(7.1); Containment Spray System (8.1); Hydrogen Recombiner (8.5); and
Special Test Exemptions (12).
Date of issuance: March 20, 1995
Effective date: March 20, 1995
Amendment Nos.: 113 and 104
Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8756) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 20, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
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 application for amendment: December 6, 1994
Brief description of amendment: This amendment deletes Technical
Specification (TS) Surveillance Requirement (SR) 4.1.3.2.2 for the
Axial Power Shaping Rods and relaxes surveillance intervals for TS 3/
4.1.3.1, ``Group Height - Safety and Regulating Rod Groups;'' TS 3/
4.4.6.2, ``Operational Leakage;'' TS 3/4.5.2, ``ECCS Subsystems - Tavg
equal to or greater than 280 deg.F;'' TS 3/4.6.2.1, ``Containment Spray
System;'' and TS 3/4.10.4, ``Special Test Exceptions Shutdown Margin.''
Date of issuance: March 21, 1995Effective date: March 21, 1995 and
implemented not later than 90 days after issuance
Amendment No.: 196
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8757) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 21, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
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 application for amendment: December 6, 1994
Brief description of amendment: This amendment revises Technical
Specification (TS) 4.0.5, ``Applicability'' and its associated Bases to
eliminate the need for NRC approval of relief requests prior to
implementation and relaxes surveillance test intervals for TS 3/
4.1.2.3, ``Reactivity Control Systems - Makeup Pump - Shutdown; TS 3/
4.1.2.4, ``Reactivity Control Systems - Makeup Pumps - Operating; TS 3/
4.1.2.6, Reactivity Control Systems - Boric Acid Pump - Shutdown; and
TS 3/4.1.2.7, ``Reactivity Control System - Boric Acid Pumps -
Operating'' from monthly to quarterly. Date of issuance: March 22, 1995
Effective date: March 22, 1995, and to be implemented within 90
days
Amendment No.: 197
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8758) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 22, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: December 9, 1994, as
supplemented on December 22, 1994.
Brief description of amendment: The amendment revises the Technical
Specification (TS) Surveillance Requirement 4.8.1.1.2f.7. The change
removes the requirement to perform the hot restart test within 5
minutes of completing the 24-hour endurance test and places that
requirement in a separate TS.
Date of issuance: March 20, 1995
Effective date: March 20, 1995, to be implemented within 30 days
Amendment No.: 95
Facility Operating License No. NPF-30. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6315) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 20, 1995. No significant hazards
consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251. [[Page 18638]]
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: August 4, 1994, as supplemented
on March 14, 1995 and March 28, 1995.
Brief description of amendment: The amendment replaces Technical
Specification (TS) 3/4.6.2.2, Spray Additive System, with a new TS 3/
4.6.2.2 entitled Recirculation Fluid pH control (RFPC) System. The
associated TS Surveillance Requirements and the Bases will also be
revised. In addition, the Bases section for the Refueling Water Storage
Tank (RWST) System will be revised.
Date of issuance: March 30, 1995
Effective date: March 30, 1995, to be implemented within 30 days
Amendment No.: 96
Facility Operating License No. NPF-30. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49440) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 30, 1995. The March 14,
1995, and March 28, 1995, letters provided supplemental information
that did not change the initial proposed no significant hazards
consideration determination. No significant hazards consideration
comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: September 8, 1994
Brief description of amendment: The amendment revises the Technical
Specification (TS) Bases Section 3/4.9 and changes Final Safety
Analysis Report (FSAR) Sections 9.1.3 ``Fuel Pool Cooling and
Cleanup,'' 9.1.4 ``Fuel Handling System'' and 15.4.6 ``Chemical and
Volume Control System Malfunction That Results in a Decrease in the
Boron Concentration in the Reactor Coolant. The changes established
procedural controls to address an unreviewed safety question.
Date of issuance: March 31, 1995
Effective date: March 31, 1995, to be implemented within 30 days
Amendment No.: 97
Facility Operating License No. NPF-30. Amendment revises the
Technical Specification Bases and FSAR.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11151) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 31, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: December 8, 1994, as
supplemented by letter dated February 16, 1995.
Brief description of amendment: The proposed amendment would change
Standby Gas Treatment Power Supply Requirements during refueling
operations.
Date of issuance: March 23, 1995
Effective date: As of the date of issuance, to be implemented
within 30 days
Amendment No.: 143
Facility Operating License No. DPR-28. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8759) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 23, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: October 31, 1994
Brief description of amendment: The amendment relocated
requirements regarding safety/relief valve position indication
instrumentation from the Technical Specifications to other licensee-
controlled documents.
Date of issuance: March 27, 1995
Effective date: March 27, 1995, to be implemented prior to restart
from the spring 1995 refueling outage
Amendment No.: 135
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65831) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 27, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
NuclearPower Plant, Kewaunee County, Wisconsin
Date of application for amendment: December 2, 1994
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant (KNPP) Technical Specification (TS) 3.2 by deleting
the requirements for the charging pumps, high concentration boric acid
in the boric acid storage tanks (BASTs), the boric acid transfer pumps,
and boric acid heat tracing. Changes to TS 3.3 and Table TS 3.5.3 add
requirements associated with the emergency core cooling system (ECCS)
accumulators, remove the requirements associated with the boric acid
storage tanks and increase the minimum required boron concentration in
the refueling water storage tank (RWST). Additionally, the surveillance
requirements involving the BASTs, associated valves and heat tracing
located in Table TS 4.1-1, Table TS 4.1-2 and Section 4.5 have been
deleted.
Date of issuance: March 28, 1995
Effective date: March 28, 1995, to be implemented within 20 days
Amendment No.: 116
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
508). The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 28, 1995. No significant hazards
consideration comments received: None.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Notice Of Issuance Of Amendments To Facility Operating LicensesAnd
Final Determination Of No Significant Hazards ConsiderationAnd
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
[[Page 18639]] and regulations. The Commission has made appropriate
findings as required by the Act and the Commission's rules and
regulations in 10 CFR Chapter I, which are set forth in the license
amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By May 12, 1995, the licensee
may file a request for a hearing with respect to issuance of the
amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for 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 [[Page 18640]] participate fully in the
conduct of the hearing, including the opportunity to present evidence
and cross-examine witnesses. Since the Commission has made a final
determination that the amendment involves no significant hazards
consideration, if a hearing is requested, it will not stay the
effectiveness of the amendment. Any hearing held would take place while
the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Northeast Nuclear Energy Company, Docket No. 50-245,
MillstoneNuclear Power Station, Unit 1, New London County,
Connecticut
Date of application for amendment: March 17, 1995
Brief description of amendment: The amendment revises Technical
Specification (TS) Surveillance Requirement 4.7.D.1.c.1 by replacing
the once per quarter stroke test for containment isolation valves
(CIVs) with the requirement that the CIVs be tested in accordance with
the inservice testing program. In addition, there are some editorial
changes, minor renumbering of subsections, to reflect the TS revisions.
Date of issuance: March 21, 1995
Effective date: As of the date of issuance to be implemented
immediately
Amendment No.: 81
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No. The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated March 21, 1995.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Dated at Rockville, Maryland, this 5th day of April, 1995.
For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV, Office of
Nuclear Reactor Regulation
[Doc. 95-8845 Filed 4-11-95; 8:45 am]
BILLING CODE 7590-01-F
