[Federal Register Volume 61, Number 21 (Wednesday, January 31, 1996)]
[Notices]
[Pages 3497-3508]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-20131]
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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 January 5, 1996, through January 19, 1996.
The last biweekly notice was published on January 22, 1996.
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 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 March 1, 1996, 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.
[[Page 3498]]
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 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.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: January 12, 1996
Description of amendment request: Compliance with 10 CFR Part 50,
Appendix J, provides assurance that the primary containment, including
those systems and components that penetrate the primary containment, do
not exceed the allowable leakage rate values specified in the Technical
Specifications and Bases. The allowable leakage rate is determined so
that the leakage assumed in the safety analyses is not exceeded.
On February 4, 1992, the NRC published a notice in the Federal
Register (57 FR 4166) discussing a planned initiative to begin
eliminating requirements marginal to safety that impose a significant
regulatory burden. Appendix J to 10 CFR Part 50, ``Primary Containment
Leakage Testing for Water-Cooled Power Reactors,'' was considered for
this initiative and the staff undertook a study of possible changes to
this regulation. The study examined the previous performance history of
domestic containments and examined the effect on risk of a revision to
the requirements of Appendix J. The results of this study are reported
in NUREG-1493, ``Performance-Based Leak-Test Program.''
Based on the results of this study, the staff developed a
performance based approach to containment leakage rate testing. On
September 12, 1995, the NRC approved issuance of this revision to 10
CFR Part 50, Appendix J, which was subsequently published in the
Federal Register on September 26, 1995, and became effective on October
26, 1995. The revision added Option B ``Performance-Based
Requirements'' to Appendix J to allow licensees to voluntarily replace
the prescriptive testing requirements of Appendix J with testing
requirements based on both overall and individual component leakage
rate performance.
Regulatory Guide 1.163, ``Performance-Based Containment Leak Test
Program,'' was developed as a method acceptable to the staff for
implementing Option B. Accordingly, the licensee has submitted, in its
application dated January 12, 1996, proposed changes to the TS to
implement 10 CFR Part 50, Appendix J, Option B, by referring to
Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test
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:
1The proposed change will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Containment leak rate testing is not an initiator of any
accident; the proposed change does not affect reactor operations or
accident analysis, and has no significant radiological consequences.
Therefore, this proposed change will not involve an increase in the
probability or consequences of any previously-evaluated accident.
2. The proposed change will not create the possibility of any
new accident not previously evaluated.
The proposed change does not affect normal plant operations or
configuration, nor does it affect leak rate test methods. The test
history at Catawba (no ILRT [integrated leak rate test] failures)
provides continued assurance of the leak tightness of the
containment structure.
3. There is no significant reduction in a margin of safety.
[[Page 3499]]
The proposed changes are based on NRC-accepted provisions, and
maintain necessary levels of reliability of containment integrity.
The performanced-based approach to leakage rate testing recognizes
that historically good results of containment testing provide
appropriate assurance of future containment integrity; this supports
the conclusion that the impact on the health and safety of the
public as a result of extended test intervals is negligible.
Based on the above, no significant hazards consideration is
created 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 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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: December 27, 1995
Description of amendment request: The proposed amendments would
modify Tables 3.3-11 and 4.3-7 of Beaver Valley Power Station Unit Nos.
1 and 2 (BVPS-1 and BVPS-2) Technical Specification (TS) 3.3.3.8 such
that only one valve position indication system for the power operated
relief valves and safety valves is required to be operable. The
licensee stated that the proposed amendments would then be consistent
with the NRC's Improved Standard Technical Specifications, NUREG-1431,
Revision 1, and with the guidance of Regulatory Guide 1.97, NUREG-0578,
and NUREG-0737.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves instrumentation which is redundant
in monitoring the position of valves and, as such, does not
influence the potential for an initiating event involving the power
operated relief valves (PORVs) or the safety valves (SVs).
Implementation of these changes will reduce the potential for
challenges to the plant due to a potential shutdown which should not
be necessary due to the restrictive nature of having unnecessary
redundant position indication in the technical specification. By
deleting the Unit No. 1 technical specification operability
requirements for the PORV acoustic detectors, and by deleting, on
both units, the technical specification operability requirements for
the SV temperature detector position indicators, the potential for
unnecessary shutdowns is reduced. When inoperable, the PORV acoustic
detectors and the SV temperature detectors presently invoke an
unnecessary action statement as another fully qualified safety-
related position indication system exists to provide indication. The
proposed change modifies Specification 3.3.3.8 actions and
surveillance requirements, but does not affect the BASES.
The remaining instrumentation on these tables [3.3-11 and 4.3-7]
will be unaffected. The remaining position indication systems for
the PORVs and SVs are fully qualified and satisfy regulatory
criteria for post accident monitoring of valve position. These
changes do not affect the ability to satisfy analysis assumptions
regarding operation of the PORVs and SVs. They do not affect the
ability to continue to meet the guidance of Regulatory Guide 1.97,
the post Three Mile Island criteria contained in NUREG 0578 and
NUREG 0737, and reflect the guidance provided in NUREG 1431,
``Improved Standard Technical Specifications'' (ISTS). Therefore, we
have concluded that these changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated in the Updated Final Safety Analysis Report
(UFSAR).
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change will reduce the potential to challenge
safety systems due to eliminating the potential for unnecessary
plant shutdowns. The proposed changes are limited to PORV and SV
position indication and do not involve any physical changes to the
PORVs or SVs or their setpoints. These changes do not delete any
design basis accident functions previously provided by the PORVs or
SVs nor has the probability of inadvertent opening been increased.
Accordingly, no new single failure has been identified as a result
of these changes. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated in the UFSAR.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes have been incorporated to eliminate a
degree of equipment redundancy and is consistent with the Improved
Standard Technical Specifications (ISTS). The Unit No. 1
specification presently requires operability of both redundant PORV
position indication systems and the primary and backup SV position
indication systems. The Unit No. 2 specification also requires
operability of the primary and backup SV position indication
systems. These changes will potentially eliminate some challenges
and potential unnecessary shutdowns by eliminating equipment
determined to be no longer necessary. Only one safety-related
position indication system is necessary to satisfy regulatory
criteria; therefore, operation of the plant in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: December 22, 1995
Description of amendment request: The proposed amendment would
revise the Duane Arnold Energy Center (DAEC) Technical Specifications
(TS) Sections 3.7.A and 4.7.A, ``Primary Containment,'' by deleting
information also contained in 10 CFR Part 50, Appendix J, Option A and
incorporating references to the Primary Containment Leakage Rate
Testing Program. These changes will allow the use of the performance
based option of containment leak testing. The request also adds
Operability and Surveillance Requirements (SRs) for the drywell air
lock. Minor administrative changes are also made. These changes are
consistent with comparable specifications in the Improved Standard
Technical Specifications (ITS), NUREG-1433. In addition to the
licensee's proposed revision to the DAEC TS, the staff will be
executing administrative changes and corrections to the TS Bases, as
submitted in letters(2) dated February 13, 1995. Sections that will be
changed or corrected are Section 1.2, Bases; Section 2.2, Bases Reactor
Coolant System Integrity; Section 3.2, Bases; Section 3.7.H/4.7.H,
Bases Containment Atmosphere Dilution; and Section 3.7.I/4.7.I, Bases
Oxygen Concentration.
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
[[Page 3500]]
consideration, which is presented below:
. The proposed revision does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Information contained in 10 CFR 50, Appendix J was deleted and
references to the Primary Containment Leakage Rate Testing Program
were added. These are administrative changes to allow the use of
performance-based containment leakage testing methods. The
containment testing program will conform with the requirements of
Option B of 10 CFR Part 50, Appendix J and approved exemptions. The
performance of the leakage tests themselves is not an input or
consideration in any accident previously evaluated, thus the
proposed change will not increase the probability of any such
accident occurring. The same operability requirements remain for the
primary containment, therefore the consequences of an accident are
not significantly increased.
Drywell air lock operability and surveillance requirements were
added. Actions for one air lock door inoperable have been added
consistent with the ITS. In addition, notes have been added to allow
entry and exit to perform repairs of the air lock components and to
explain that the previous overall leak test is not invalidated by an
inoperable door. This change represents an additional restriction on
plant operation, since the previous condition of one air lock door
inoperable did not require any actions to be taken. A requirement to
verify proper operation of interlock mechanism was also added. This
will ensure that one door is always closed which maintains primary
containment integrity.
The addition of these new drywell air lock requirements provides
more stringent provisions than previously existed in the [current
Technical Specifications]. The more stringent requirements will not
result in operation that will increase the probability of initiating
an analyzed event. If anything, the new requirements may decrease
the probability or consequences of an analyzed event by
incorporating the more restrictive changes discussed above. These
changes will not alter assumptions relative to mitigation of an
accident or transient event. The more restrictive requirements will
not alter the operation of process variables, structures, systems,
or components as described in the safety analyses.
The TS revision includes the relocation of certain requirements
from the current technical specification (CTS) to licensee
controlled documents. CTS 4.7.A.1.e contains a requirement to
replace the T-ring inflatable seals for the 18 inch purge valves
every four years. This provision is not in the ITS as it is a
maintenance issue and not a surveillance for operability. CTS
4.7.A.1.e also contains a requirement to verify (during Type C
testing) that the mechanical modification which limits the maximum
opening angle for the 18 inch purge valves is intact. The ITS only
requires this surveillance if the mechanical modification is not
permanent. At DAEC, the 18 inch purge valves are permanently blocked
to restrict opening to 30 deg.. These CTS provisions will be
relocated to plant procedures. Any changes to these relocated
requirements will require an evaluation in accordance with 10 CFR
50.59. CTS 4.7.A.1.a and 4.7.A.1.d contain some procedural details
that are not contained in Appendix J. These details will also be
relocated to plant procedures, consistent with the ITS. Since any
changes to these licensee controlled documents will be evaluated in
accordance with 10 CFR 50.59, no significant increase in the
probability or consequences of an accident previously evaluated will
be allowed.
The proposed revision does not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident, nor does it affect
any assumptions or conditions in the accident analysis. The proposed
revision does not degrade any existing plant programs, nor modify
any functions of safety related systems or accident mitigation
functions previously credited at the DAEC. The proposed changes do
not impact initiators of analyzed events. They also do not impact
the assumed mitigation of accidents or transient events. These TS
changes will not alter assumptions made in the safety analysis and
licensing basis.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed revision does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Deleting information from the TS which is contained in 10 CFR
50, Appendix J and adding references to the Primary Containment
Leakage Rate Testing Program are purely administrative changes to
allow the use of performance-based containment leakage testing
methods. The containment testing program will conform with the
requirements of Option B of 10 CFR Part 50, Appendix J and approved
exemptions. The use of Option B will maintain the containment safety
functions as a barrier to the release of radioactivity to the
environment.
The proposed revision does not make any physical or operational
changes to existing plant systems or components, nor does it alter
any plant parameters, revise any safety limit setpoint, or provide
any new release pathways. The proposed revision does not change any
transient responses assumed in the Design Bases of the plant.
The proposed changes which relocate requirements to licensee
controlled documents will not alter the plant configuration (no new
or different type of equipment will be installed) or change the
methods governing normal plant operation. These changes will not
alter assumptions made in the safety analysis or licensing basis.
The proposed changes which add more restrictive requirements to
the CTS will not alter the plant configuration (no new or different
type of equipment will be installed) or change the methods governing
normal plant operation. These changes do impose different
requirements. However, they are consistent with assumptions made in
the safety analyses.
Therefore, the revision does not create the possibility of a new
or different kind of accident previously evaluated.
3. The proposed revision will not significantly reduce any margin
of safety.
Deleting information from the TS which is contained in 10 CFR
50, Appendix J and adding references to the Primary Containment
Leakage Rate Testing Program do not involve a significant reduction
in the margin of safety. These changes are administrative in nature
and either eliminate a redundant requirement or clarify the
applicability and acceptability of an alternative, NRC approved,
leak rate testing provision within the TS. The containment testing
program will conform to the requirements of Option B of 10 CFR Part
50, Appendix J and approved exemptions. The use of Option B will
maintain the containment safety functions as a barrier to the
release of radioactivity to the environment.
The proposed revision does not require any modifications to
existing plant systems or equipment, safety limit settings, or
parameters utilized in the licensing bases for the safety analysis.
The proposed revision does not change any safety analysis or any
accident mitigation action for which DAEC has previously taken
credit. The proposed changes do not involve any technical changes;
they have no impact on any safety analysis assumptions. The addition
of new requirements either increases or does not affect the margin
of safety.
The proposed changes that relocate requirements from the CTS to
licensee controlled documents will not reduce a margin of safety
since they have no impact on any safety analysis assumptions. In
addition, the requirements to be relocated from the CTS to the
licensee controlled document are unchanged. Since any future changes
to this licensee controlled document will be evaluated in accordance
with the requirements of 10 CFR 50.59, no significant reduction in a
margin of safety will be allowed.
The proposed changes are consistent with NUREG-1433, which was
approved by the NRC Staff. The changes are also consistent with NRC
guidance provided for the implementation of Option B. The change
controls for proposed relocated details and requirements are
acceptable. Therefore, revising the TS to reflect the NRC accepted
level of detail and requirements ensures that there is no reduction
in a margin of safety.
Therefore, the proposed revision will not significantly reduce
any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
[[Page 3501]]
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869
NRC Project Director: Gail H. Marcus
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: December 14, 1995
Description of amendment request: The proposed amendment would
modify Technical Specifications 3.3.1.1, ``Reactor Protection System
(RPS) Instrumentation,'' and 3.3.6.1, ``Primary Containment and Drywell
Isolation Instrumentation,'' to eliminate periodic response time
testing of selected analog trip modules (ATMs). This request is
supported by analyses prepared by the Boiling Water Reactor Owners'
Group topical report NEDO-32291, ``System Analyses for Elimination of
Selected Response Time Testing Requirements,'' which demonstrate that
other periodic tests required by technical specifications, such as
channel calibrations, channel functional tests and logic system
functional tests, are adequate to ensure ATM response times remain
within acceptable limits.
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 purpose of the proposed Technical Specification (TS)
change is to eliminate response time testing requirements for
selected analog trip modules (ATMs) in the Reactor Protection System
(RPS) and the main steam isolation valve (MSIV) isolation actuation
instrumentation. The Boiling Water Reactor Owners' Group (BWROG) has
completed an evaluation which demonstrates that response time
testing is redundant to the other TS-required testing. These other
tests, in conjunction with actions taken in response to NRC Bulletin
90-01, ``Loss of Fill-Oil in Transmitters Manufactured by
Rosemount,'' and Supplement 1, are sufficient to identify failure
modes or degradations in instrument response time and ensure
operation of the associated systems within acceptable limits. There
are no known failure modes that can be detected by response time
testing that cannot also be detected by other TS-required testing.
This evaluation was documented in NEDO-32291, ``System Analyses for
Elimination of Selected Response Time Testing Requirements,''
January 1994. Illinois Power (IP) has confirmed the applicability of
this evaluation to Clinton Power Station (CPS). In addition, IP has
completed the actions identified in the NRC staff's safety
evaluation of NEDO-32291.
Because of the continued application of other existing TS-
required tests such as channel calibrations, channel checks, channel
functional tests, and logic system functional tests, the response
time of these systems will be maintained within the acceptance
limits assumed in plant safety analyses and required for successful
mitigation of an initiating event. The proposed changes do not
affect the capability of the associated systems to perform their
intended function within their required response time, nor do the
proposed changes themselves affect the operation of any equipment.
As a result, IP has concluded that the proposed changes do not
involve a significant increase in the probability or the
consequences of an accident previously evaluated.
(2) The proposed changes only apply to the testing requirements
for ATMs in the systems identified above and do not result in any
physical change to these or other components or their operation. As
a result, no new failure modes are introduced. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
(3) The current TS-required response times are based on the
maximum values assumed in the plant safety analyses. These analyses
conservatively establish the margin of safety. As described above,
the proposed changes do not affect the capability of the associated
systems to perform their intended function within the allowed
response time used as the basis for the plant safety analyses. The
potential failure modes for the components within the scope of this
request were evaluated for impact on instrument response time. This
evaluation confirmed that the remaining TS-required testing is
sufficient to identify failure modes or degradations in instrument
response times and to ensure that operation of the instrumentation
within the scope of this request is within acceptable limits. As a
result, it has been concluded that plant and system response to an
initiating event will remain in compliance with the assumptions of
the safety analysis.
Further, although not explicitly evaluated, the proposed changes
will provide an improvement to plant safety and operation by
reducing the time safety systems are unavailable, reducing the
potential for safety system actuations, reducing plant shutdown
risk, limiting radiation exposure to plant personnel, and
eliminating the diversion of key personnel resources to conduct
unnecessary testing. Therefore, IP has concluded that this request
will result in an overall increase in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606
NRC Project Director: Gail H. Marcus
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: December 11, 1995
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) Section 4.7, Surveillance
Requirements for Primary Containment Automatic Isolation Valves.
Specifically, the proposed amendment would delete TS Surveillance
Requirement 4.7.D.4, which requires replacement of the seat seals for
the drywell and suppression chamber purge and vent valves every 5
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:
The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
An evaluation of the operational performance of the 18-inch
purge and vent valves has concluded that deletion of the Monticello
Technical Specification surveillance requirement 4.7.D.4 will have
no adverse impact on the seat leakage performance of these primary
containment isolation valves, no adverse impact on the testing
performed in accordance with 10 CFR 50, Appendix J, and thus no
adverse impact on the containment isolation function of these
primary containment isolation valves. The material of which the T-
shaped elastomer seat is comprised of has been found to withstand
normal and accident thermal exposures for the design life of the
plant based on a thermal aging analysis. Radiation effects will not
have an adverse impact on the elastomer seat material. Therefore,
this amendment will not cause a significant increase in the
probability or consequences of an accident previously evaluated for
the Monticello plant.
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The proposed change to the Technical Specifications for the
Primary Containment Purge and Vent valves does not alter the
function of these components or their interrelationships with other
systems. Therefore, this amendment will not create the possibility
of a new or different kind of accident from any accident previously
analyzed.
The proposed amendment will not involve a significant reduction
in the margin of safety.
The operating experience of these valves has demonstrated that
the testing performed
[[Page 3502]]
in accordance with 10 CFR 50, Appendix J, provides a high level of
confidence in the ability of these valves to perform their safety
function with respect to valve leak tightness. The proposed
amendment will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: December 27, 1995
Description of amendment requests: The amendments would revise the
combined Technical Specifications (TS) 3/4.6.1.1, ``Containment
Integrity;'' 3/4.6.1.2, ``Containment Leakage;'' 3/4.6.1.3,
``Containment Air Locks;'' 3/4.6.1.6, ``Containment Structural
Integrity;'' 3/4.6.3, ``Containment Isolation Valves;'' and their
associated Bases; and would add TS 6.8.4.j, ``Containment Leakage Rate
Testing Program,'' to implement the performance-based leakage rate
testing program, as permitted by 10 CFR Part 50, Appendix J. These
changes will support the implementation of the performance-based
testing of Option B to Appendix J for Types A, B, and C containment
leakage rate testing and the appropriate rescheduling of testing. The
amendment changes the TS to implement 10 CFR Part 50, Appendix J,
Option B, by referring to Regulatory Guide 1.163, ``Performance-Based
Containment Leakage Test 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 proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to Technical Specification (TS) 3/4.6.1.1,
3/4.6.1.2, 3/4.6.1.3, 3/4.6.1.6, 3/4.6.3, and the addition of 6.8.4
j., to implement the performance-based Containment Leakage Rate
Testing Program have no effect on plant operation. The proposed
changes only provide mechanisms within the TS for implementing a
performance-based methodology for determining the frequency of leak
rate testing that has been approved by the Commission. The test type
and test method used for testing would not be changed. The test
acceptance criteria would not be changed, and containment leakage
will continue to be maintained within the required limits.
Directly referencing the Containment Leakage Rate Testing
Program for containment ILRT [integrated leak rate testing] and LLRT
[local leak rate test] requirements does not involve any
modification to plant equipment or affect the operation or design
basis of the containment. Leakage rate testing is not a precursor to
or an initiating event for any accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes only allow for the implementation of 10 CFR
50, Appendix J, Option B, testing frequencies and do not involve any
modifications to any plant equipment or affect the operation or
design basis of the containment. The proposed changes do not affect
the response of the containment during a design basis accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect or change a Safety Limit or
affect plant operations. The changes only implement the allowed 10
CFR 50, Appendix J, Option B testing frequencies that have been
determined by the Commission not to involve a safety concern. The
testing method, acceptance criteria, and basis for testing are not
changed and still provide assurance that the containment will
provide its intended function.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman
Saxton Nuclear Experimental Corporation (SNEC), Docket No. 50-146,
Saxton Nuclear Experimental Facility (SNEF), Bedford County,
Pennsylvania
Date of amendment request: November 21, 1995.
Description of amendment request: The proposed amendment would
change the license and technical specifications to add GPU Nuclear
Corporation (GPUN) as a licensee for the SNEF along with SNEC and would
transfer from SNEC to GPUN all management-related responsibilities for
the SNEF. Responsibility for safely maintaining the containment vessel
and performing characterization activities would change from SNEC to
GPUN. Technical specification organizational positions would be changed
from SNEC titles to GPUN titles. GPUN would take responsibility from
SNEC for administration of all SNEF functions, for radiation safety
activities, and for providing on-site management and continuing
oversight of production activities. The appointment of members to the
Saxton Radiation Safety Committee and the reporting of the Committee
would change from the SNEC President to the GPUN Vice President of the
Nuclear Services Division. The GPUN President would have the authority
to request audits and would receive audit reports instead of the SNEC
President. Procedure control methodology and the administrative
procedure for procedures would be changed from SNEC procedures to GPUN
procedures. The responsibility for records retention and reporting
would change from SNEC to GPUN. The organization chart for the facility
would be changed to reflect the addition of GPUN as a licensee.
Basis for proposed no significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve a significant hazards
considerations because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Because the proposed changes are administrative in nature they
would have no effect on the likelihood or impact on the potential
accidents of fire, flood or radiological hazard.
[[Page 3503]]
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Because the proposed changes are administrative in nature they
would not create the possibility of a new or different kind of
accident from any accident previously analyzed.
3. Involve a significant reduction in a margin of safety.
Because the proposed changes are administrative in nature they
would not involve any reduction in a margin of safety.
The NRC staff has reviewed the analysis of the licensee 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: Saxton Community Library, 911
Church Street, Saxton, Pennsylvania 16678 Attorney for the Licensee:
Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge,
2300 N Street, NW., Washington, DC 20037
NRC Project Director: Seymour H. Weiss
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: January 4, 1996 (TS 95-22)
Description of amendment request: The proposed change would extend
the functional testing interval for the following isolation radiation
monitor instruments from monthly to quarterly: (1) Engineered Safety
Feature Actuation System Instrumentation Surveillance Requirements
Table 4.3-2, Item 3.c.3, Containment Purge Air Exhaust Monitor
Radioactivity-High; (2) Radiation Monitoring Instrumentation
Surveillance Requirements Table 4.3-3, Item 1.a, Fuel Storage Pool Area
Radiation Monitor; (3) Table 4.3-3, Item 2.a, Containment Purge Air
Exhaust; (4) Table 4.3-3, Item 2.b.i, Containment Gaseous Activity RCS
Leakage Detection; (5) Table 4.3-3, Item 2.b.ii, Containment
Particulate Activity RCS Leakage Detection; and (6) Table 4.3-3, Item
2.c, Control Room Isolation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Review of the past history for the affected and similar
radiation monitors revealed that extending the functional testing
interval for these monitors will not adversely affect system
operability and will effectively increase system availability. These
radiation monitors are not accident initiating equipment, thus
increasing the surveillance interval on these monitors will not
affect the probability of any accident previously evaluated. Based
on the above statements, it is concluded that the probability or
consequences of an accident previously evaluated is not increased.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
No new type of accident or malfunction will be created since the
radiation monitors are not accident initiating equipment. The
proposed change merely increases the functional testing interval for
the affected radiation monitors, and does not change the method and
manner of plant operation. The safety design bases in the Updated
Final Safety Analysis Report have not been altered.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The proposed changes do not change the plant configuration in a
way that introduces a new potential hazard to the plant and do not
involve a significant reduction in the margin of safety. The
proposed changes do not affect applicable safety analysis acceptance
criteria and will not affect system operating conditions.
Additionally, plant operating experience with similar monitors has
shown that there has not been additional failures due to the
quarterly testing frequency. Thus, it is concluded that the margin
of safety is not reduced.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: November 22, 1995
Description of amendment request: The proposed amendment replaces
the requirements associated with the boron dilution mitigation system
(BDMS) in the Wolf Creek Generating Station Technical Specifications
with alarms, indicators, procedures, and controls to assure proper
resolution of potential inadvertent boron dilution events.
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 does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The only event potentially impacted by the proposed change is
the inadvertent boron dilution event. The discussion of the
probability and consequences of an inadvertent boron dilution event
at WCGS is provided in USAR [Updated Safety Analysis Report] Section
15.4.6. Primarily, the proposed changes revise the method of
detecting and mitigating the event. The only aspect of the changes
that impact[s] the potential causes of an inadvertent boron dilution
event is the increased requirement to isolate potential dilution
sources in Modes 3, 4 and 5. As a result, the overall probability of
the event is slightly decreased.
The alternate methods to detect and mitigate this event achieve
the same basic goal as the current BDMS; to prevent a return to
critical during an inadvertent dilution event. The proposed changes
to the BDMS will result in an improved system that will provide an
improved response to the inadvertent boron dilution event, and that
will prevent a return to critical. Thus, it can be concluded that
the proposed change will not significantly increase the consequences
of a postulated inadvertent boron dilution event.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The revisions to plant procedural requirements to either operate
a reactor coolant pump or to isolate/control potential dilution
sources does [sic] not create the potential for a new or different
kind of accident because these new requirements are configurations
which have always been allowed. Similarly, the new normal position
for the letdown divert valve does not create a new or different
accident because the new normal position has always been an allowed
position. The other procedural changes only increase the plant
operators' awareness of potential boron dilution problems or provide
the steps needed to respond to available indications and alarms to
mitigate the potential event. As a result, these procedural changes
do not create the possibility of a new or different kind of
accident.
The proposed changes also include addition of new redundant VCT
high level alarms and a new alarm indicating that the
[[Page 3504]]
letdown divert valve is not in the ``VCT'' position. Because the alarms
are passive, they do not create the possibility of a new or
different kind of accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The design criterion and margin of safety for the current BDMS
is that the dilution event is terminated prior to the loss of all
shutdown margin. The same criterion will be met following the
implementation of the proposed changes. Therefore, there is no
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room 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: December 20, 1995
Description of amendment request: This amendment request proposes
to revise Technical Specification 3/4.6.1.1, ``Containment Integrity,''
and 3/4.6.1.3, ``Containment Air Locks,'' and to add Technical
Specification 6.8.4i, ``Containment Leakage Rate Testing Program,'' to
implement the new performance-based leakage rate testing program as
permitted by 10 CFR 50, Appendix J. Also, Technical Specification 1.7e,
``Containment Integrity,'' would be revised to reference Technical
Specification 4.6.1.1.c. These proposed changes will implement the
performance-based testing of Option B to Appendix J, for Type A, B, and
C containment leak testing by referring to Regulatory Guide 1.163,
``Performance-Based Containment Leakage-Test 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 proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to Technical Specifications 3/4.6.1.1 and
3/4.6.1.3, and the addition of Technical Specification 6.8.4i to
implement the new performance based Containment Leakage Rate Testing
Program, have no effect on plant operation. The proposed changes
only provide mechanisms within the technical specifications for
implementing a performance-based methodology, for determining the
frequency of leak rate testing, which has been approved by the NRC.
The test type and test method used for testing would not be changed.
The test acceptance criteria would not be changed, and containment
leakage will continue to be maintained within the required limits.
Directly referencing the Containment Leakage Rate Testing
Program for containment integrated leak rate test and local leak
rate test requirements does not involve any modification to plant
equipment or affect the operation or design basis of the
containment. Leakage rate testing is not a precursor to or an
initiating event for any accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes only allow for implementation of 10 CFR 50,
Appendix J, Option B, testing frequencies and do not involve any
modifications to any plant equipment or affect the operation or
design basis of the containment. The proposed changes do not affect
the response of the containment during a design basis accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect or change a Safety Limit, any
limiting condition for operation or affect plant operations. The
changes only implement the allowed Option B testing frequencies that
have been determined by the NRC not to involve a safety concern. The
testing method, acceptance criteria, and bases are not changed and
still provide assurance that the containment will provide its
intended function.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
[[Page 3505]]
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: November 14, 1995, as
supplemented January 4, 1996.
Brief description of amendments: The amendments revise the
Technical Specifications to incorporate 10 CFR Part 50, Appendix J,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' Option B. Technical Specification changes for the LaSalle
facility will be addressed under separate correspondence.
Date of issuance: January 11, 1996
Effective date: January 11, 1996
Amendment Nos.: 148, 142, 169, and 165
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 7, 1995 (60 FR
62896). The January 4, 1996, supplement provided a specific
implementation date for the requested amenement. This information was
within the scope of the original application and did not change the
staff's initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 11, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of application for amendment: November 14, 1995
Brief description of amendment: The amendment revises the Haddam
Neck Technical Specifications (TS) to provide an one-time exception to
TS 3.9.12, '' Fuel Building Storage Air Cleanup System,'' to allow the
fuel storage building air cleanup system to be inoperable for a limited
duration during intervals in which new fuel rack modules will be moved
into and old fuel rack modules will be moved out of the fuel storage
building.
Date of Issuance: January 17, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 187
Facility Operating License No. DPR-61. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 28, 1995 (60
FR 58688) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated January 17, 1996 No significant
hazards consideration comments received: No.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457.
Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan Date of application for amendment:
November 8, 1995, as supplemented November 17, 1995
Brief description of amendment: The amendment removes the
prescriptive Type A containment leakage test rate frequency of 40 plus
or minus 10 months and adds a reference to perform containment leakage
rate tests in accordance with the criteria specified in Appendix J of
10 CFR Part 50 as modified by approved exemptions. In addition, the
amendment revises the test pressure for Type B and C testing to correct
a typographical error.
Date of issuance: January 16, 1996
Effective date: January 16, 1996
Amendment No.: 117
Facility Operating License No. DPR-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62489) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 16, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: March 29, 1995, as supplemented
by letters dated September 18 and November 16, 1995
Brief description of amendments: The amendments revise Technical
Specification requirements for the Low Temperature Overpressure
Protection system and update the heatup and cooldown curves for both
units.
Date of issuance: January 11, 1996
Effective date: As of the date of issuance to be implemented within
60 days
Amendment Nos.: Unit 1 - 162; Unit 2 - 144
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications. Date of initial notice in Federal
Register: September 27, 1995 (60 FR 49933) The September 18 and
November 16, 1995, letters provided clarifying information that did not
change the scope of the March 29, 1995, application and the initial
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated January 11, 1996. No significant hazards
consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: December 2, 1994
Brief description of amendments: The amendments replace Appendix B,
``Environmental Technical Specifications,'' with an Environmental
Protection Plan (Nonradiological) and revise the Operating Licenses to
reflect these changes.
Date of issuance: December 19, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: Unit 1 - 199 - Unit 2 - 140
Facility Operating License Nos. DPR-57 and NPF-5. Amendments
revised the Technical Specifications and Operating Licenses.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
502) The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 19, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
[[Page 3506]]
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: March 17, 1995, as supplemented
by letter dated July 6, 1995
Brief description of amendments: The amendments revise Technical
Specification 3/4.9.4, Containment Building Penetrations, to allow the
personnel airlock to be open during core alterations or movement of
irradiated fuel within the containment.
Date of issuance: November 30, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 92 and 70
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35077) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 30, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 17, 1995, as supplemented by
letters dated November 22, and December 18, 20, and 27, 1995
Brief description of amendment: The amendment revised the primary
containment air lock technical specifications to allow the air locks to
be open in Mode 5 (refueling) during core alterations except for
movement of recently irradiated fuel. All other provisions of the
August 17, 1995, requests are defered.
Date of issuance: January 11, 1996
Effective date: January 11, 1996
Amendment No.: 85
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47619) The additional information contained in the supplemental
letters dated November 22, and December 18, 20, and 27, 1995, was
clarifying in nature and thus, within the scope of the initial notice
and did not affect the staff's proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated Janaury 11, 1996.
No significant hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 20, 1995
Brief description of amendment: The proposed amendment revised the
technical specifications to eliminate the response time testing
requirements for selected Reactor Protection System Instrumentation.
Date of issuance: January 11, 1996
Effective date: January 11, 1996
Amendment No.: 86
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62492) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated Janaury 11, 1996. No significant
hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: May 25, 1995 (AEP:NRC:1071T)
Brief description of amendments: The amendments incorporate a
cycle- and burnup-dependent peaking factor penalty in the Core
Operating Limits Report and add an appropriate reference to the COLR
and update the topical report reference in the Technical
Specifications.
Date of issuance: January 4, 1996
Effective date: January 4, 1996, with full implementation within 45
days
Amendment Nos.: Unit 1, 206, Unit 2, 190
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated January 4, 1996. No significant hazards consideration
comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: October 3, 1995
Brief description of amendment: The amendment removes the Limiting
Condition for Operation (LCO) and Surveillance Requirements for the
loss-of-normal power (LNP) trip function from Tables 3.2.2 and 4.2.1
and inserts new LCO 3.2.F and Surveillance Requirement 4.2.F. In
addition, the amendment adds a new table to specify the required LNP
instrumentation for each bus, updates the Table of Contents, makes some
editorial changes, and revises the associated Bases section.
Date of issuance: January 17, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 92
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 4, 1995 (60 FR
62111) The Commission's related evaluation of the amendment is
contained in a Safety evaluation dated January 17, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
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: September 11, 1995, as
supplemented November 15, 1995.
Brief description of amendment: The amendment changes Technical
Specification Sections 3.4.8 and 3.9.9, Tables 2.2-1, 3.3-3, 3.3-5 and
3.3-8, and Bases Sections 3/4.2.1, 3/4.4.8 and 3/4.11.2.1. These
changes combine several different administrative changes which will
correct typographical errors, provide clarifications, or make editorial
changes.
Date of issuance: January 17, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 194
[[Page 3507]]
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52933) The November 15, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated January 17, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
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: September 29, 1995, as
supplemented November 9, 1995.
Brief description of amendment: The amendment provides three
changes to the Technical Specifications (TS) relating to the
pressurizer safety valves (PSV) and the main steam safety valves
(MSSV).
The first change is to TS 3.4.2.1 and 3.4.2.2 and involves relaxing
the as-found setpoint tolerance for the pressurizer safety valves
(PSVs) and the main steam safety valves (MSSVs) from the current value
of plus or minus 1% to plus or minus 3%. Table 4.7-1 is also modified
to correct the as-found tolerance for the MSSV from plus or minus 1% to
plus or minus 3%. Notes are added to TS 3.4.2.2 and Table 4.7-1 which
specify that the lift setting should be determined at nominal operating
conditions and should be set at plus or minus 1% of the lift setting.
For the second change, Surveillance Requirement 4.7.1.1 and Table
4.7-1 are modified to eliminate the need to verify the orifice size of
each MSSV.
The third change modifies the statement for TS 3.7.1.1 so that if a
MSSV is inoperable and compensating action cannot be taken, the plant
must be brought to hot shutdown (Mode 4) within 12 hours instead of
cold shutdown (Mode 5) in 30 hours.
Date of issuance: January 18, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 195
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 25, 1995 (60 FR
54723) The November 9, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated January 18, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
PECO Energy Company, Public Service Electric and Gas Company
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: November 30, 1995
Brief description of amendments: The amendments change the
technical specification requirements for control rod drive scram
accumulator and charging water header minimum pressure.
Date of issuance: January 11, 1996
Effective date: Unit 2, as of date of issuance, to be implemented
concurrently with Amendment 210, issued August 30, 1995; Unit 3, as of
date of issuance, to be implemented concurrently with Amendment 214,
issued August 30, 1995.
Amendments Nos.: 211 and 216
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 8, 1995 (60 FR
63073) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 11, 1996 No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas Company
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: November 21, 1995
Brief description of amendments: The amendments change the test
pressure requirements for the high pressure coolant injection system
and the reactor core isolation cooling system surveillance tests. The
amendments also change Section 5.5.7 of the technical specifications to
eliminate reference to a section which was previously eliminated.
Date of issuance: January 11, 1996
Effective date: As of date of issuance.
Amendments Nos.: 212 and 217
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 5, 1995 (60 FR
62271) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 11, 1996 No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas Company
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: December 19, 1995
Brief description of amendments: These amendments change the
ventilation filter test program bypass and penetration leakage test
acceptance criteria from less than 0.05 percent to less than 1.0
percent. The change corrects an administrative error that occurred
during the development of the Peach Bottom Improved Technical
Specifications which were issued as Amendments 210 and 214 to the Peach
Bottom licenses on August 30, 1995.
Date of issuance: January 16, 1996
Effective date: Unit 2, effective as of date of issuance, to be
implemented concurrently with Amendment 210, issued August 30, 1995;
Unit 3, effective as of date of issuance, to be implemented
concurrently with Amendment 214, issued August 30, 1995.
Amendments Nos.: 213 and 218
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications. Public comments requested as to
proposed no significant hazards consideration: Yes (60 FR 66997,
December 27, 1995). That notice provided an opportunity to submit
comments on the Commission's proposed no significant hazards
[[Page 3508]]
consideration determination. No comments have been received. The notice
also provided for an opportunity to request a hearing by January 26,
1996, but indicated that if the Commission makes a final no significant
hazards consideration determination any such hearing would take place
after issuance of the amendment. The Commission's related evaluation of
the amendments, finding of exigent circumstances, and final
determination of no significant hazards consideration are contained in
a Safety Evaluation dated January 16, 1996
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Dated at Rockville, Maryland, this 23rd day of January 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation.
[Doc. 96-1683 Filed 1-30-96; 8:45 am]
BILLING CODE 7590-01-F
