[Federal Register Volume 60, Number 31 (Wednesday, February 15, 1995)]
[Notices]
[Pages 8741-8766]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-3629]
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NUCLEAR REGULATORY COMMISSION
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations; Biweekly Notice
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 20, 1995, through February 3, 1995.
The last biweekly notice was published on February 1, 1995 (60 FR
6296).
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 17, 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 [[Page 8742]] 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.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: January 19, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Surveillance Requirement 4.0.3 and
its associated bases to provide for a delay period of up to 24 hours in
which to perform a surveillance which has been discovered not to have
been performed within its specified frequency. This change would adopt
the requirements of NUREG-1431, ``Standard Technical Specifications,
Westinghouse Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The proposed change will reduce the requirement to unnecessarily
manipulate and challenge plant systems and equipment. The most probable
result of performing a surveillance during the delay period will be to
verify its conformance with Technical Specification requirements. Since
this change does not affect plant design, operation, or the manner in
which testing is performed, the consequences of accident scenarios
postulated in the Final Safety Analysis Report will not increase.
Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed change does not introduce any new equipment, nor does
it require existing systems to perform a different type of function
than they are currently designed to perform. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant reduction
in the margin of safety.
The margin of safety is neither described or prescribed for this
specification. The proposed change simply provides additional time to
perform a surveillance and verify that the operability of equipment is
in conformance with the Technical Specification requirements.
Therefore, the proposed 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
NRC Project Director: William H. Bateman.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: December 23, 1994.
Description of amendment request: The proposed amendments would
increase the allowable enrichment of new fuel stored in the new fuel
storage vault (NFSV), revise the enrichment description of fuel in the
reactor core, and include references to documents previously approved
by the staff in the [[Page 8743]] Technical Specifications that provide
analytical methods used to determine core operating 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:
A.1. The proposed change does not involve a significant increase in
the probability of occurrence or consequences of any accident
previously evaluated.
The Updated Final Safety Analysis Report (UFSAR) does not consider
any accidents involving the NFSV. The Fuel Handling Accidents that are
analyzed (Section 15.7.4) include dropping of a spent fuel assembly
onto the spent fuel pool floor and breaking of all fuel rods, and
dropping of a fuel assembly inside containment onto the top of the
core.
The proposed change to increase the NFSV fuel enrichment limit from
4.0 to 4.65 weight percent U-235 does not affect any of the initiators
or precursors of any accident previously evaluated. The proposed change
will not increase the likelihood that a transient initiating event will
occur because transients are initiated by equipment malfunction and/or
catastrophic system failure. Since the proposed change does not involve
the introduction of new or redesigned plant equipment, failure
mechanisms are not affected. As a result, the probability of occurrence
of accidents previously evaluated is not significantly increased.
A new criticality analysis for the proposed change to increase the
NFSV fuel enrichment limit from 4.0 to 4.65 weight percent U-235 was
performed for the NFSV. It was determined that even in worst case
conditions the acceptance criteria was met since the maximum Keff
was determined to be well below the 0.95 limit with a 95/95
probability/confidence level. The consequences of any accident,
including a fuel handling accident involving the NFSV, are not
significantly increased.
A.2. The proposed change does not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed change to the Technical Specifications does not
involve the addition of any new or different types of safety related
equipment, nor does it involve the operation of equipment required for
safe operation of the facility in a manner different from those
addressed in the safety analysis. No safety related equipment or
function will be altered as a result of the proposed changes. Also, the
procedures governing normal plant operation and recovery from an
accident are not changed by the proposed Technical Specification
changes. Since no new failure modes or mechanisms are added by the
proposed changes, the possibility of a new or different kind of
accident is not created.
A.3. The proposed change does not involve a significant reduction
in a margin of safety.
Plant safety margins are established through LCOs, limiting safety
system settings, and safety limits specified in the Technical
Specifications. There will be no changes to either the physical design
of the plant or to any of these settings and limits as a result of
increasing the NFSV fuel enrichment limit. The change does not involve
a significant increase in the probability of occurrence or consequences
of any accident previously evaluated or create the possibility of a new
or different kind of accident from any previously analyzed.
Additionally, the revised criticality analysis demonstrates that the
maximum Keff under all postulated conditions remains below the
acceptance value of 0.95. Therefore, the change will not result in a
significant reduction in a margin of safety.
B.1. The proposed change does not involve a significant increase in
the probability of occurrence or consequences of any accident
previously evaluated.
The proposed change to increase the reactor core fuel enrichment
range discussed in the Design Features section of Technical
Specifications from ``between 2.2 to 4.0'' to ``up to 4.65'' weight
percent U-235 is administrative in nature and does not affect any of
the initiators or precursors of any accident previously evaluated. The
proposed change will not increase the likelihood that a transient
initiating event will occur because transients are initiated by
equipment malfunction and/or catastrophic system failure. Since the
proposed change does not involve the introduction of new or redesigned
plant equipment, failure mechanisms are not affected. As a result, the
probability of occurrence of accidents previously evaluated is not
significantly increased.
The fuel enrichment limit of each core is determined by the core
specific design and is determined to be acceptable with respect to the
accident analysis by the reload analysis and is not impacted by the
value specified in the description in the Design Features section of
Technical Specifications. This value is only provided as the highest
expected core fuel enrichment in the Design Features section discussion
of the reactor core. This change is administrative in nature and does
not affect the consequences of any accident previously evaluated.
B.2. The proposed change does not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed change in the reactor core fuel enrichment description
contained in the Design Features section of Technical Specifications
does not involve the addition of any new or different types of safety
related equipment, nor does it involve the operation of equipment
required for safe operation of the facility in a manner different from
those addressed in the safety analysis. No safety related equipment or
function will be altered as a result of the proposed change. Also, the
procedures governing normal plant operation and recovery from an
accident are not changed by the proposed Technical Specification
change. Since no new failure modes or mechanisms are added by the
proposed change, the possibility of a new or different kind of accident
is not created.
B.3. The proposed change does not involve a significant reduction
in a margin of safety.
Plant safety margins are established through LCOs, limiting safety
system settings, and safety limits specified in the Technical
Specifications. There will be no changes to either the physical design
of the plant or to any of these settings and limits as a result of
increasing reactor core fuel enrichment value given in the Design
Features section of Technical Specifications. The change does not
involve a significant increase in the probability of occurrence or
consequences of any accident previously evaluated or create the
possibility of a new or different kind of accident from any previously
analyzed.
Based on the above discussion, the ability to safely shutdown the
operating unit and mitigate the consequences of all accidents
previously evaluated will be maintained. Therefore, the margin of
safety is not significantly affected.
C.1. The proposed change does not involve a significant increase in
the probability of occurrence or consequences of any accident
previously evaluated.
The proposed change to add three documents to the list of documents
that provide the analytical methods to determine core operating limits
is administrative in nature and does not affect any of the initiators
or precursors of any accident previously evaluated. The proposed change
will not increase the likelihood that a transient initiating event will
occur because transients are initiated by equipment malfunction
[[Page 8744]] and/or catastrophic system failure. Since the proposed
change does not involve the introduction of new or redesigned plant
equipment, failure mechanisms are not affected.
The documents have been previously reviewed and approved by the NRC
and it was determined that they provide an acceptable means to
determine core operating limits. As a result, the probability of
occurrence of accidents previously evaluated is not significantly
increased. Since the documents provide NRC approved methodologies for
determining core operating limits, the addition of the documents to
Technical Specifications or use of the documents to determine core
operating limits will not significantly increase the consequences of
any accident previously evaluated.
C.2. The proposed change does not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed change to add three documents to the list of documents
that provide the analytical methods to determine core operating limits
is administrative in nature and does not involve the addition of any
new or different types of safety related equipment, nor does it involve
the operation of equipment required for safe operation of the facility
in a manner different from those addressed in the safety analysis. No
safety related equipment or function will be altered as a result of the
proposed changes. Also, the procedures governing normal plant operation
and recovery from an accident are not changed by the proposed Technical
Specification changes. Since no new failure modes or mechanisms are
added by the proposed changes, the possibility of a new or different
kind of accident is not created.
C.3. The proposed change does not involve a significant reduction
in a margin of safety.
Plant safety margins are established through LCOs, limiting safety
system settings, and safety limits specified in the Technical
Specifications. There will be no changes to either the physical design
of the plant or to any of these settings and limits as a result of
adding references to the new documents. The ability to mitigate the
consequences of all accidents previously evaluated will be maintained.
Therefore, the margin of safety is not significantly affected.
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: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
NRC Project Director: Robert A. Capra.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York Date
of amendment request: September 19, 1994.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 4.4.A.3 to reference the
testing frequency requirements of 10 CFR Part 50, Appendix J, and to
state that NRC approved exemptions to the applicable regulatory
requirements are permitted. This proposed administrative revision
simply deletes the paraphrased language and directly references
Appendix J.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated
The proposed change will provide a one-time exemption from the 10
CFR [Part] 50, Appendix J Section III.D.1.(a) leak rate test schedule
requirement. This change will allow for a one-time test interval for
Type A Integrated Leak Rate Tests (ILRTs) of approximately 70 months.
Leak rate testing is not an initiating event in any accident,
therefore this proposed change does not involve a significant increase
in the probability of a previously evaluated accident.
Type A tests are capable of detecting both local leak paths and
gross containment failure paths. The history at IP-2 [Indian Point 2]
demonstrates that Type B and C Local Leak Rate Tests (LLRTs) have
consistently detected any excessive local leakages.
Administrative controls govern the maintenance and testing of
containment penetrations such that the probability of excessive
penetration leakage due to improper maintenance or valve misalignment
is very low. Following maintenance on any containment penetration, an
LLRT is performed to ensure acceptable leakage levels, following any
LLRT on a containment isolation valve, an independent valve alignment
check is performed. Therefore, Type A testing is not necessary to
ensure acceptable leakage rates through containment penetrations.
While Type A testing is not necessary to ensure acceptable leakage
rates through containment penetrations, Type A testing is necessary to
demonstrate that there are no gross containment failures. Structural
failure of the containment is considered to be a very unlikely event,
and in fact, since IP-2 has been in operation it has never failed a
Type A ILRT. Therefore, a one-time exemption increasing the interval
for performing an ILRT should not result in a significant decrease in
the confidence in the leak tightness of the containment structure.
The proposed change also revises Technical Specification 4.4.A.3 to
reference the testing frequency requirements of 10 CFR [Part] 50,
Appendix J, and to state that NRC approved exemptions to the applicable
regulatory requirements are permitted. The current language of TS
4.4.A.3 paraphrases the requirements of Section III.D.1.(a) of Appendix
J. The proposed administrative revision simply deletes the paraphrased
language and directly references Appendix J. No new requirements are
added, nor are any existing requirements deleted. Any specific changes
to the requirements of Section III.D.1.(a) will require a submittal
from Consolidated Edison 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 TS changes if
different exemptions are approved in the future.
The proposed change, in itself, does not affect reactor operations
or accident analysis 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.
Therefore, the 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 exemption request does not affect normal plant
operations or configuration, nor does it affect leak rate test methods.
The proposed change allows a one-time test interval of
[[Page 8745]] approximately 70 months for the ILRT. Given the test
history of IP-2 of no Type A test failures during plant lifetime, the
relaxation in schedule should not significantly decrease the confidence
in the leak tightness of the containment.
The proposed Technical Specification amendment provides
clarification to a specification that paraphrases a codified
requirement.
Since the proposed change would not change the design,
configuration or method of operation of the plant, it 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 purpose of the existing schedule for ILRTs is to ensure that
the release of radioactive materials will be restricted to those leak
paths and leak rates assumed in accident analyses. The relaxed schedule
for ILRTs does not allow for relaxation of Type B and C LLRTs.
Therefore, methods for detecting local containment leak paths and leak
rates are unaffected by this proposed change. Given that the test
history for ILRTs shows no failure during plant life, a one-time
increase of the test interval does not lead to a significant
probability of creating a new leakage path or increased leakage rates,
and the margin of safety inherent in existing accident analyses is
maintained.
The proposed Technical Specification change is administrative and
clarifies the relationship between the requirements of TS 4.4.A.3,
Appendix J and any approved exemptions to Appendix J. It does not, in
itself, change a safety limit, an LCO [limiting condition for
operation], or a surveillance requirement on equipment required to
operate the plant. The NRC will directly approve any proposed change or
exemption to [Section] III.D.1.(a) of Appendix J prior to
implementation.
Therefore, this change does not involve a significant reduction in
the margin of safety.
Based on the Safety Analysis, it is concluded that: (1) The
proposed change does not constitute a significant hazards consideration
as defined by 10 CFR 50.92 and (2) there is reasonable assurance that
the health and safety of the public will not be endangered by the
proposed change. Moreover, because this action does not involve a
significant hazards consideration, it will also not result in a
condition which significantly alters the impact of the station on the
environment as described in the NRC Final Environmental Statement.
Although the licensee has included an evaluation of a proposed
exemption to 10 CFR part 50, Appendix J requirements in the above
determination of no significant hazards consideration, only the part
related to the amendment is pertinent to this notice of proposed
amendment. The exemption request will be considered as a separate
matter on its own merits. The NRC staff has reviewed the licensee's
analysis and, based on this review, it appears that the three standards
of 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Ledyard B. Marsh
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: October 31, 1994
Description of amendment request: The requested amendments would
remove the stroke times for the steam generator power operated relief
valves (PORVs) from Technical Specification (TS) Tables 3.6-2a and 3.6-
2b. The PORVs are part of the main steam vent to atmosphere system. The
PORV actuators have difficulty developing enough closing thrust to
adequately overcome all of the friction loads within the valves;
therefore, difficulty exists in consistently meeting the present 5-
second closing stroke time requirement. The licensee requests the
proposed change on the basis that the PORVs do not receive an actual
containment isolation signal; therefore, it is justified to remove the
stroke times from TS Tables 3.6-2a and 3.6-2b.
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:
In 48 FR 14870, the Commission has set forth examples of amendments
that are considered not likely to involve significant hazards
considerations. Example (vi) describes a change which either may result
in some increase to the probability or consequences of a previously-
analyzed accident or may reduce in some way a safety margin, but where
the results of the change are clearly within all acceptable criteria
with respect to the system or component specified in the Standard
Review Plan. In this case, the proposed amendment is similar to example
(vi) in that it removes the required isolation time of the steam
generator PORVs from TS Tables 3.6-2a and 3.6-2b; however, no adverse
impact upon accident analyses is created as a result.
Criterion 1
The requested amendments will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The effects of the delays in isolation times on the various transients
affected have been analyzed and found to be acceptable. Since these
valves do not receive a containment isolation signal, and no credit is
taken for operation of these valves in the dose analysis for a
containment isolation function, a maximum stroke time does not apply
for containment isolation.
Criterion 2
The requested amendments will not create the possibility of a new
or different kind of accident from any accident previously evaluated.
SV PORV closure (provided the valves are not already closed at the
start of the transient) is a response to a transient already in
progress. The possibility of a spurious SV PORV opening will not be
affected by the requested amendments. No equipment or component
reconfiguration will occur as a result of this change. Finally, no
changes to plant procedures are being made which would affect any
accident causal mechanisms.
Criterion 3
The requested amendments will not involve a significant reduction
in a margin of safety. The isolation times which are applicable to
these valves are specified in TS Table 3.3-5, Engineered Safety
Features Response Times. The effects of the isolation of these valves
were evaluated based on their ESF function, not a containment isolation
function, and determined to be acceptable.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 8746]] 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
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: June 13, 1994, as supplemented August
15, 1994.
Description of amendment request: The proposed changes would
increase the initial fuel enrichment limit from a current maximum of
4.0 weight % to 4.75 weight % and establish new loading patterns for
new and irradiated fuel in the spent fuel pool to accommodate this
increase. These changes would also increase the efficiency of fuel
storage cell use in the spent fuel pools and provide additional
flexibility to the reload design efforts at Duke Power Company, while
at the same time maintaining sufficient criticality safety margin and
decay heat removal capabilities.
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.
There is no increase in the probability or consequences of an
accident in the new fuel vault since the only credible accidents for
this area are criticality accidents and it has been shown that
calculated, worst case Keff for this area is 0.95 under
all conditions.
There is no increase in the probability of a fuel drop accident in
the Spent Fuel Storage Pool since the mass of an assembly will not be
affected by the increase in fuel enrichment. The likelihood of other
accidents, previously evaluated and described in Section 9.1.2 of the
FSAR [Final Safety Analysis Report], is also not affected by the
proposed changes. In fact, it could be postulated that since the
increase in fuel enrichment will allow for extended fuel cycles, there
will be a decrease in fuel movement and the probability of an accident
may likewise be decreased. There is also no increase in the
consequences of a fuel drop accident in the Spent Fuel Pool since the
fission product inventory of individual fuel assemblies will not change
significantly as a result of increased initial enrichment. In addition,
no change to safety related systems is being made. Therefore, the
consequences of a fuel rupture accident remain unchanged. Also, it has
been shown that keff is 0.95, under all conditions
therefore, the consequences of a criticality accident remain unchanged
as well.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident since fuel handling accidents (fuel drop and
misplacement) are not new or different kinds of accidents. Fuel
handling accidents are already discussed in the FSAR for fuel with
enrichments up to 4.1 weight %. As described in Section VI.9 of
Attachment IV, additional analyses have been performed for fuel with
enrichment up to 4.75 weight %. Worst case misloading accidents
associated with the new loading patterns were evaluated. For all
possible misloading accidents the negative reactivity provided by
soluble boron maintains keff 0.95. of safety.
3. The proposed changes do not involve a significant reduction in
the margin of safety.
The proposed change does not involve a significant reduction in the
margin of safety since, in all cases, a keff 0.95 is
being maintained. Criticality analyses have been performed which show
that the new fuel storage vault will remain subcritical under a variety
of moderation conditions, from fully flooded to optimum moderation. As
discussed above, the Spent Fuel Pool will remain sufficiently
subcritical during any fuel misplacement accident.
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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: November 11, 1994, as supplemented
January 30, 1995.
Description of amendment request: The amendments would revise the
Technical Specifications Design Features section to establish
restricted loading patterns and associated burnup criteria for placing
fuel in the Oconee Spent Fuel Pools. These changes are necessary to
address two new fuel designs which have increased initial fuel
enrichment and therefore cannot be stored in the spent fuel pools under
existing Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Standard 1. The proposed amendments will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
Each accident analysis addressed in the Oconee Final Safety
Analysis Report (FSAR) has been examined with respect to changes in
Cycle 15 parameters to determine the effect of the Cycle 16 reload and
to ensure that the acceptance criteria of the FSAR safety analyses
remain satisfied. The transient evaluation of Cycle 16 is considered to
be bounded by previously accepted analyses. Section 7 of the Reload
Report addresses ``Accident and Transient Analysis'' for this core
reload.
There is no increase in the probability or consequences of an
accident due to the spent fuel storage restrictions proposed in this
amendment request. It has been shown that the calculated, worst case
keff for this area is [less than or equal to] 0.95 under all
conditions. There is no increase in the probability of a fuel drop
accident in the SFP [spent fuel pool] since the mass of the new
assemblies is not significantly different from the mass of the old
assemblies. The likelihood of other accidents, previously evaluated and
described in the FSAR, is also not affected by the proposed changes. In
fact, it could be postulated that since the increase in fuel enrichment
will allow for extended fuel cycle lengths, there will be a decrease in
fuel movement and the probability of an accident may actually be
reduced. There is also no increase in the consequences of a fuel rod
drop accident in the SFP since the fission product inventory of
[[Page 8747]] individual fuel assemblies will not change significantly
as a result of increasing the initial enrichment. In addition, no
change to safety related systems is being made. Therefore, the
consequences of a fuel rupture accident remain unchanged. In addition,
it has been shown that keff is [less than or equal to] 0.95 under
all conditions. Therefore, the consequences of a criticality accident
in the SFP remain unchanged as well. The above analysis ensures that
the proposed reload amendment request 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 kind of accident from any accident previously evaluated.
The analyses performed in support of this reload are in accordance
with the NRC approved methods delineated in Specification 6.9.2. The
predicted operating characteristics of Oconee 3 Cycle 16 are similar to
previously licensed designs. The Mark B10T and Mark B11 fuel assembly
designs remain mechanically compatible with all fuel handling
equipment. Therefore, no new or different kind of fuel handling
accident is created by the proposed amendment request.
Section 15.11 of the Oconee FSAR states that the refueling boron
concentration is maintained such that a criticality accident during
refueling is not considered credible. The proposed amendment request
continues to assure that a criticality accident in the SFP or during
refueling is not credible. The double contingency principle discussed
in ANSI N-16.1-1975 and the April 1978 NRC letter allows credit for
soluble boron under other abnormal or accident conditions, since only a
single accident need be considered at one time. Thus, by requiring a
minimum boron concentration in the SFP, a criticality accident caused
by violating the SFP storage restrictions is not considered credible.
Therefore, the proposed amendment request does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction in
the margin of safety.
The Oconee 3 Cycle 16 design was performed using the NRC approved
methods given in Specification 6.9.2. The safety limits for Oconee 3
Cycle 16 are unchanged from previous cycles. The limits and margins
summarized in the Oconee 3 Cycle 16 Reload Report are well within the
allowable limits and requirements, and reflect no reductions to any
margins of safety.
The proposed change does not involve a significant reduction in the
margin of safety related to SFP criticality. In all cases, a keff
[less than or equal to] 0.95 is maintained. Criticality analyses have
been performed which show that the SFP will remain sufficiently
subcritical during any fuel misplacement accident. In summary the
proposed changes do not involve a significant reduction in the margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036.
NRC Project Director: Herbert N. Berkow.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: January 6, 1995
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 3/4.8.1.1, ``AC Sources-
Operating,'' and 3/4.8.1.2, ``AC Sources-Shutdown,'' to (1) revise the
minimum quantity of fuel oil required in the day tanks and the storage
tanks, (2) add specific actions to be taken if the storage tank levels
fall below minimum requirements, (3) revise and relocate to the
associated Bases the fuel oil sampling and testing criteria, and (4)
add specific actions to be taken if the fuel oil properties do not meet
specified limits. The proposed amendment would also revise TS 6.8.4,
``Programs,'' to add a requirement for a diesel fuel oil testing
program. The licensee stated that the proposed changes are consistent
with the NRC's Improved Standard Technical Specifications (NUREG-1434).
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 operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The diesel generators are not initiators or precursors to an
accident previously evaluated. The diesel generators are required to
provide onsite power to safe shutdown loads as assumed in the accident
analysis. Therefore, the proposed changes to the diesel generator fuel
oil specifications cannot significantly affect the probability of a
previously evaluated accident.
The proposed change to the minimum required diesel generator fuel
oil levels is based on updated calculations of fuel consumption rates.
Because the updated calculations assume a lower consumption rate, the
new minimum fuel oil levels are lower but still assure that a seven-day
fuel oil capacity is available. Accordingly, the proposed change has no
effect on the operation of the diesel generator. The proposed change to
allow 48-hours to restore diesel generator fuel oil to the minimum
required level does not affect short-term diesel generator operability
and is acceptable based on the remaining fuel oil capacity (>6 days),
initiating the process for procuring additional fuel and the low
probability of an event requiring a diesel generator during this
interval. Also, the proposed allowance of a limited time to restore
diesel fuel oil properties to required limits will not affect the
short-term operability of the diesel generator. Even with minor
degradation of the fuel oil properties, the diesels will start and
perform their intended function. Relocation of the testing requirements
to the bases and adding a description of the Diesel Fuel Oil Testing
Program to the Administrative Control section are administrative
changes. The diesel fuel oil will continue to be sampled and tested in
a manner to assure its quality. In summary, the changes will not
adversely affect the performance or the ability of the diesel
generators to perform their intended function. Therefore, the proposed
changes will not significantly increase the consequences of an accident
previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes will revise the minimum required diesel
generator fuel oil levels and requirements associated with diesel
generator fuel oil properties. [[Page 8748]] The changes do not
introduce any new accident precursors and do not involve any
alterations to plant configurations which could initiate a new or
different kind of accident. The proposed changes do not affect the
short-term operability of the diesel generator. In addition, the
operability of the diesel generators is assured by periodic testing and
preventive maintenance. Therefore, the proposed changes will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
Safety margins are established through safety analyses. These
analyses assume that at least one diesel generator will start and load
whenever offsite power is lost. The proposed change to the minimum
required diesel generator fuel oil levels is based on updated
calculations of fuel consumption rates. The updated calculations use
the guidance delineated in Regulatory Guide 1.137 which is based on
time-dependent loads of the diesel-generators during design basis
events. Calculations based on time dependent loads result in new
minimum fuel oil levels which are lower. This change has no effect on
the operation of the diesel generator or on a margin of safety. The
allowance of a limited time to restore the fuel oil levels, or to
analyze and restore fuel oil properties to required limits, is
justified since the short term operability of the diesel generators is
not affected. Relocation of the fuel oil testing requirements to the
Bases does not affect the quality of the fuel oil. The 10CFR50.59
process will assure that future changes to the Bases will maintain the
current margins of safety, and that the diesel fuel oil will continue
to be sampled and tested in such a manner as to assure its quality.
Adding a description of the Diesel Fuel Oil Testing Program to the
Administrative Control section of Technical Specifications are
administrative. Therefore, the diesel generator will continue to
operate as analyzed and there will not be a significant reduction in a
margin of safety.
The proposed changes are further justified in that they are
consistent with the requirements of the Improved Standard Technical
Specifications (NUREG-1434).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Ledyard B. Marsh.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: January 6, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 3/4.3.7.5, ``Accident Monitoring
Instrumentation,'' and TS 3/4.4.2, ``Safety/Relief Valves.'' TS 3/
4.3.7.5 would be revised to delete certain instruments not classified
as Category 1 (Type A or non-Type A) as defined in Regulatory Guide
1.97 and to delete the requirement that accident monitoring
instrumentation be operable in Operational Condition 3. The ACTIONS of
TS Table 3.3.7.5-1 would be revised to allow 30 days to restore one
inoperable channel and 7 days to restore two inoperable channels. TS
3.3.7.5 would be revised to add an exception to the requirements of TS
3.0.4. In addition, editorial changes would be made to TS Tables
3.3.7.5-1 and 4.3.7.5-1 for consistency and clarity.
The proposed amendment would also revise TS 3/4.4.2 to remove
requirements related to safety/relief valve acoustic monitors to be
consistent with the proposed changes to TS Tables 3.3.7.5-1 and
4.3.7.5-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of NMP2 [Nine Mile Point Nuclear Station Unit 2] in
accordance with the proposed amendment, will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
PAM [Post-Accident Monitoring] instruments are used to help guide
operator response to postulated accidents. Thus, the status or
operability of PAM instrumentation does not affect the probability of
previously analyzed accidents. The non-Category 1 PAM instruments being
removed from the Technical Specifications do not meet any of the
Commission's screening criteria and are not of controlling importance
to safety or necessary to obviate the possibility of an abnormal
situation or event giving rise to an immediate threat to public health
and safety. The operability of critical parameters necessary to assure
proper response to previously analyzed accidents (i.e., Category 1
instruments) is still controlled by the Technical Specifications. Thus,
deleting non-Category 1 instruments will not increase the consequences
of any accident previously evaluated.
PAM instruments are related to the diagnosis and preplanned actions
required to mitigate DBAs [Design Basis Accidents] assumed to occur in
Operational Conditions 1 and 2. A DBA during Operational Condition 3 is
extremely unlikely. The requirement to maintain the Reactor Water
Level, Suppression Pool Water Level and Drywell High Range Radiation
Monitor instrumentation operable in Operational Condition 3 will be
deleted. Because Suppression Pool Water Level indication will no longer
be required in Operational Condition 3, its ACTION requirement was
revised to delete the requirement to place the plant in COLD SHUTDOWN,
Operational Condition 4. This is consistent with ITS [Improved Standard
Technical Specifications] which requires that the plant be brought to
an operational condition in which the LCO [Limiting Condition for
Operation] does not apply if a required action cannot be met.
Therefore, deleting the requirement that PAM instruments be operable
during Operational Condition 3 and changing the ACTION requirement for
Suppression Pool Water Level Monitoring does not affect the probability
or consequences of an accident.
The passive nature of the Category 1 PAM instruments (i.e., those
instruments that initiate no critical automatic action) and the
alternate means available to obtain the required information assure an
acceptable level of safety is maintained during operation with
instrument channels out of service. Since an acceptable level of safety
is maintained with inoperable channels, plant startup or operation with
inoperable channels will not alter plant response to analyzed
accidents. Thus, the proposed changes to the required ACTIONS and the
proposed exemption to Specification 3.0.4 will not increase the
consequences of analyzed events.
The proposed changes to the requirements for PCIV [Primary
[[Page 8749]] Containment Isolation Valve] indication are consistent
with the proposed required ACTIONS. Position indication will still be
required for each operable PCIV and penetrations without adequate PCIV
indication status will be isolated, thus assuring containment integrity
in the event of an accident. Deletion of the ``Minimum Required
Actions'' column in Table 3.3.7.5-1 is consistent with the proposed
ACTIONS for LCO 3.3.7.5, since compensatory actions are based on
compliance with the ``Required Number of Channels.'' Deleting the
``Applicable Operating Conditions'' column is consistent with the
proposed changes and other NMP2 Technical Specifications sections.
Finally, referencing Specification 4.0.5 is an administrative change
which does not alter any existing surveillance requirements for the
safety relief valves.
In aggregate, the proposed changes do not affect the plant in a way
that could directly contribute to causing or mitigating the effects of
an accident. Therefore, the operation of NMP2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The operation of NMP2, in accordance with the proposed amendment,
will not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not represent a physical change to the
plant as described in the NMP2 USAR [Updated Safety Analysis Report].
The proposed changes do not modify any plant equipment and the initial
conditions used for the design basis accident analysis are still valid.
Thus, no potential initiating events are created which would cause any
new or different kinds of accidents. PAM instrumentation is used to
guide operator response during postulated accidents. Those PAM
instruments considered of controlling importance to safety are retained
in the Technical Specifications. Thus, plant response to previously
analyzed events is not altered so as to create any new or different
kinds of accidents. Therefore, operation of Nine Mile Point Unit 2 in
accordance with the proposed change will not create the possibility of
a new or different kind of accident from any previously assessed.
The operation of NMP2, in accordance with the proposed amendment,
will not involve a significant reduction in a margin of safety.
The non-Category 1 PAM instruments being removed from the Technical
Specifications do not meet any of the Commission's screening criteria.
That is, the instruments being proposed for removal are not of
controlling importance to safety or necessary to obviate the
possibility of an abnormal situation or event giving rise to an
immediate threat to public health and safety. Thus, they are not
critical to any margin of safety.
PAM instruments are related to the diagnosis and preplanned actions
required to mitigate DBAs assumed to occur in Operational Conditions 1
and 2. A DBA during Operational Condition 3 is extremely unlikely. The
requirement to maintain the Reactor Water Level, Suppression Pool Water
Level and Drywell High Range Radiation Monitor instrumentation operable
in Operational Condition 3 will be deleted. Because Suppression Pool
Water Level indication will no longer be required in Operational
Condition 3, its ACTION requirement was revised to delete the
requirement to place the plant in COLD SHUTDOWN, Operational Condition
4. This is consistent with the ITS, which requires that the plant be
brought to an operational condition in which the LCO does not apply if
a required action cannot be met. Therefore, deleting the requirement
that PAM instruments be operable during Operational Condition 3 and
changing the ACTION requirement for Suppression Pool Water Level
Monitoring does not significantly reduce a margin of safety.
Since the Category 1 PAM instruments are passive in nature (i.e.,
no critical automatic action is assumed to occur from these
instruments) and alternate means exist to obtain the required
information, an acceptable level of safety is assured when instrument
channels are out of service. Also, the probability of an event
requiring PAM instrumentation is low. Continued operation with one
channel out of service, and limited plant operation with two channels
out of service, does not compromise plant safety margins. An acceptable
level of safety is maintained during plant startups and operation with
instrument channels out of service. Thus, the proposed changes to the
required ACTIONS and the proposed exemption to Specification 3.0.4 will
not significantly reduce a margin of safety.
The proposed changes to PCIV indication will assure correct
implementation of the ACTIONS discussed above. Isolating the flow path
associated with one or two inoperable PCIV indication channels is
conservative since the subject valve will be positioned as required to
assure primary containment integrity. The remaining editorial changes
are administrative in nature and by definition do not affect safety
margins. Deleting the ``Minimum Operable Channels'' and ``Applicable
Operating Conditions'' columns is consistent with the proposed changes.
Finally, referencing the requirements of Specification 4.0.5 is an
administrative change and by definition does not reduce the margin of
safety.
Therefore, the operation of NMP2 in accordance with 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Ledyard B. Marsh.
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: December 9, 1994.
Description of amendment request: The proposed changes incorporate
NRC recommendations contained in Generic Letter 93-05 related to the
diesel generator (DG) surveillance requirements and other DG
surveillance requirements related to the cold starts. The proposed
changes to the DG operability testing surveillance requirements are
consistent with the intent of GL 93-05 however vary in some
particulars, because of circumstances specific to Millstone 3. The
proposed changes will modify the requirement for the DG operability
testing when the other DG is inoperable, delete the requirement for DG
operability testing when one or both offsite AC sources are inoperable,
eliminate fast loading of DGs except for the 18-month test, and modify
the hot restart test from the 24-hour loaded test run for the DGs.
Basis for proposed no significant hazards consideration
determination: [[Page 8750]] 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 a SHC because the changes
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed changes revise the action requirements regarding
operability testing of a non-affected DG when the other DG is
inoperable, delete the requirement for operability testing of the DGs
when one or both offsite AC sources are inoperable and eliminate the
fast loading of DGs except for the 18-month test. These changes will
improve DGs performance by reducing the number of unnecessary quick
starts and by requiring more appropriate testing of the DGs when there
is a potential for common mode failure. The proposed change, to revise
the method of verifying DG hot restart capability after a 24-hour run
without loading the DG with LOP/SI [loss of offsite power/safety
injection] load, meets an intent of Regulatory Guide 1.108, Position
C.2.a.5, which states the purpose of the test as to ``demonstrate
functional capability at full load temperature conditions.'' Functional
capability of the DG can be adequately demonstrated by manually or
automatically restarting the DG within five minutes after a 24-hour
test run without loading it with LOP/SI loads, provided that a full
load temperature condition is maintained prior to restart. The proposed
DG restart method does not reduce the effectiveness of the test. The
proposed revisions of the DG surveillance requirements will not
increase the probability of an accident and it will not change the
response of the DG to a LOP as described in the Millstone Unit No. 3
FSAR. Since the plant response to an accident will not change, there is
no change in the potential for an increase in the consequences of an
accident previously analyzed.
2. Create the possibility of a new or different kind of an accident
previously evaluated.
The proposed changes of the DG surveillance requirements and
operability testing requirements do not affect the operation or
response of any plant equipment or introduce any new failure
mechanisms. The proposed changes do not affect the test results and the
DGs will be verified to be operable and their response to a loss of
voltage will be unchanged. The plant equipment will respond per the
design and analyses and there will not be a malfunction of a new or any
type introduced by the revision to the DG surveillance requirements. As
such, the changes do not create the possibility of a new or different
kind of accident previously evaluated.
3. Involve a significant reduction in the margin of safety.
The bases of Technical Specification 3/4.8, ``Electrical Power
Systems,'' state that the operability of the AC and DC power systems
and associated distribution systems ensure that sufficient power will
be available to supply the safety-related equipment required for safe
shut down and mitigation and control of accident conditions. The bases
also state that the surveillance requirements for determining the
operability of the DGs are in accordance with the recommendations of
Regulatory Guide 1.108, Revision 1. The revisions of the surveillance
requirements establishes tests that will continue to verify that the
DGs are operable and the testing will still meet the intent of
Regulatory Guide 1.108, Revision 1. Operable DGs ensure that the
assumptions in the bases of the Technical Specifications are not
affected and ensure that the margin of safety is not reduced.
Therefore, the assumptions in the bases of the technical specifications
are not affected and these changes do not result in 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: 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.
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: December 14, 1994.
Description of amendment request: The proposed amendment would
revise the Millstone Unit No. 3 Technical Specifications by:
1. Increasing the upper bound of the overall containment integrated
leakage rate required by Technical Specification 3.6.1.2.a from 0.3 wt.
% per day to 0.65 wt. % per day of the containment air per 24 hours at
design basis pressure.
2. Revising Technical Specification 4.6.6.1.d.3 by providing more
margin with respect to the drawdown time for secondary containment
vacuum.
3. Revising Bases Section 3/4.7.9 to reflect the above 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 (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
consequences of an accident previously evaluated.
* * * There is a reasonable assurance that the modified criteria for
the negative pressure in the secondary containment boundary proposed
via the proposed change (i.e., a negative pressure of 0.1 inches in one
minute and a negative pressure of 0.4 inches within the next two
minutes), can be accomplished in the prescribed time.
Extension of the time allowed to achieve the final drawdown of
secondary containment from 120 seconds to 180 seconds (these times
include the diesel generator start and load time of approximately 11
seconds) will have a negligible impact on heating and cooling. Plant
experience has shown that heatup and cooldown of thick-walled concrete
structures, such as the Millstone Unit No. 3 auxiliary building, is a
relatively slow process. Also, natural convection within the auxiliary
building tends to stabilize temperatures. Following an accident signal,
ventilation equipment is restarted promptly. Therefore, heatup or
cooldown, during short periods while ventilation fans and/or heaters
are inactive, is insignificant and can be neglected.
The proposed change to reinstate the containment integrated leakage
rate at the design basis pressure from 0.3 wt % per day to 0.65 wt %
per day has been evaluated to determine the impact to the Appendix J
requirements for Type A, B and C Testing. In addition, the radiological
consequence evaluation also addressed the increase in La (i.e.,
from 0.3 wt % per day to 0.65 wt % per day).
On October 12, 1993, Millstone Unit No. 3 successfully conducted
the second [[Page 8751]] Type A test in the first 10-year service
period. Test results indicated that the ``As-Found'' and ``As-Left''
ILRTs [integrated leakage rate tests] passed the technical
specification acceptance criteria. The ``As-Found'' value was 0.1327
weight percent per day and the ``As-Left'' value was 0.1313 weight
percent per day. These values represent 27.2% and 26.9% of the
technical specification criterion of 0.4875 wt % per day (0.75
La), based on La equal to 0.65 wt % per day, respectively. In
addition, as of October 9, 1993, the total Type B and C ``As-Found''
and ``As-Left'' leakage results were 0.099 wt % per day, and 0.084 wt %
per day, respectively. These values represent approximately 25.3% and
21.5% of the technical specification limit of 0.39% wt % per day (0.6
La), based on La equal of 0.65 wt % per day, respectively.
Correspondingly, the 1993 Type A, B, and C test results indicate that
the ``As-Found'' and ``As-Left'' result in each test case was below the
existing Technical Specification limit of 0.3 wt % per day. This
further demonstrates the overall leakage integrity of the containment
and its boundaries.
Based on the relatively low ``As-Left'' ILRT leakage rate (i.e.,
0.1313 wt % per day is well below the existing technical specification
limit of 0.225 wt % per day (0.75 La), based on La equal to
0.3 wt % per day), which represents the overall containment integrated
leakage rate for the containment prior to start-up, there is reasonable
assurance that containment integrity will be maintained below the
allowable leakage rate limit of 0.65 wt % per day. In addition, the
total Type B and C ``As-Left'' leakage result of 0.084 wt % per day
(this is well below the existing technical specification limit of 0.18
wt % per day (0.6 La), based on La equal to 0.3 wt % per
day), provides further assurance that leakage, based on individual
penetration, will be maintained within sufficient margin of the leakage
limits.
Because the last Type A, B, and C tests were performed under the
technical specification limit of 0.65 wt % per day, the proposed change
to restore La to 0.65 wt % per day has no impact to these systems
from a leakage allowance perspective. As indicated above, the previous
test results met the technical specification leakage limits (based on
0.65 wt % per day) within sufficient margin and, therefore, would not
present any challenge to these leakage limits.
NNECO has evaluated the proposed changes to Surveillance
Requirement 4.6.6.1.d.3 that increase the time to draw a final required
negative pressure as measured at the 24'-6'' elevation of the auxiliary
building in conjunction with the proposed change to reinstate the
containment integrated leakage rate of 0.65 wt % per day to determine
the impact on the offsite doses following a LOCA. The calculated
radiological doses are, in most cases, less than the previously
calculated doses (i.e., EAB [exclusion area boundary] and LPZ [low-
population zone] doses) and are within the 10CFR100 limits. Previously,
the EAB thyroid and whole body doses as documented in the November 4,
1993, submittal were calculated to be 141 REM and 9.4 REM respectively,
while the previously docketed (i.e., the November 4, 1993, submittal)
LPZ doses to the thyroid and whole body were calculated to be 29.8 REM
and 1.7 REM respectively. Utilizing the revised application of
containment recirculation spray DF, the EAB thyroid and whole body
doses were calculated to be 61 REM and 16.7 REM, respectively, and the
LPZ thyroid and whole body doses were calculated to be 10.9 REM and 2.8
REM respectively. The assumptions used in the above radiological dose
calculations are provided in Attachment 1. It is noted that a LOCA at
Millstone Unit No. 3 is also one of the bounding accidents for the
Millstone Unit No. 3 control room, Millstone Unit No. 2 control room,
and the Millstone Technical Support Center habitability analysis.
Therefore, the doses for these areas were recalculated and are
presented in the Safety Assessment section above. The Millstone Unit
No. 1 control room and the Emergency Operating Facility doses are
bounded by the Millstone Unit No. 1 LOCA calculations.
The Millstone Unit Nos. 2 and 3 control rooms and Millstone
Technical Support Center doses were not recalculated in 1993 (i.e.,
November 4, 1993, submittal) since EAB/LPZ doses proved that the
releases were less than the 1990 submittal. In summary, all control
room and Technical Support Center doses are within the guidelines of
GDC 19. Therefore, the proposed changes do not result in an increase in
consequences of an accident (i.e., a LOCA) previously analyzed.
The proposed changes to Bases Section 3/4.6.6 do not have any
safety impact since they only reflect the changes proposed to
Surveillance Requirement 4.6.6.1.d.3.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not compromise the ability of the SLCRS
[supplementary leak collection and release system] and ABFS [auxiliary
building filter system] to mitigate the consequences of an accident.
The proposed changes do not make any physical or operational changes to
existing plant structures, systems or components. The proposed changes
do not introduce any new or unique operational modes or accident
precursors. Therefore, the proposed changes do not create 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.
NNECO has evaluated the proposed changes to Surveillance
Requirement 4.6.6.1.d.3 that increase the time to draw a final required
negative pressure as measured at the 24'-6'' elevation of the auxiliary
building in conjunction with the proposed change to reinstate the
containment integrated leakage rate of 0.65 wt % per day to determine
the impact on the offsite doses following a LOCA. The calculated
radiological doses are, in most cases, less than the previously
calculated doses and these doses are within the 10CFR100 limits. All
control rooms and technical support center doses are within the
guidelines of GDC 19. Therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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.
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: December 23, 1994.
Description of amendment request: The proposed amendment would
change the acceptance criteria for the peak transient generator voltage
from 4784 volts to 5000 volts during full load rejection tests of the
diesel generator (DG), and delete the 10-year surveillance requirement
to perform a [[Page 8752]] 110% pressure test of the DG fuel oil
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 (SHC), which is presented below:
* * * The proposed changes do not involve a SHC because the changes
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
DG Full-Load Rejection Test
NNECO is proposing to modify Surveillance Requirement 4.8.1.1.2.g.3
of the Millstone Unit No. 3 Technical Specifications by changing the
acceptable transient voltage to 5000 volts from 4784 volts. This change
will permit the DG full load rejection tests to be performed at
realistic plant conditions using a power factor that will envelope the
calculated power factor during the worst kW loading conditions. The
transient voltage of 5000 volts is within the normal design limits of
the DGs.
The proposed change does not alter the intent of the surveillance,
does not involve any physical changes to the plant, does not alter the
way any structure, system, or component functions, and does not modify
the manner in which the plant is operated. As such, the proposed change
to Surveillance Requirement 4.8.1.1.2.g.3 will not degrade the
capability of the DGs to perform their intended safety function, and
will not reduce the availability of the DGs. Actually, the proposed
change will increase the effectiveness of the full load rejection
tests, because the DGs will be tested in a configuration that is closer
to the design basis conditions.
Pressure Test of the DG Fuel Oil System
The DG fuel oil system is classified as an ASME Code Class 3 system
in accordance with the guidance of Regulatory Guide 1.26, ``Quality
Group Classification and Standards for
Water-, Steam-, and Radioactive-waste Components of Nuclear Power
Plants.'' Surveillance Requirement 4.0.5 requires the testing of ASME
Class 1, 2, and 3 components in accordance with Section XI of the ASME
Code. Surveillance Requirement 4.8.1.1.2.i.2 is redundant to the ASME
Section XI pressure test requirements of Surveillance Requirement
4.0.5. Additionally, the DG fuel oil tank cannot be tested in the
configuration required by Surveillance Requirement 4.8.1.1.2.i.2,
because the tanks are vented to the atmosphere and the vent cannot be
isolated. Therefore, NNECO is proposing to delete Surveillance
Requirement 4.8.1.1.2.i.2.
The proposed change does not modify the manner in which the DGs
respond to an accident. Also, the proposed change does not reduce the
reliability of the DGs.
Conclusion
Based on the above, the proposed changes to Surveillance
Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No.
3 Technical Specifications do not involve a significant increase in the
probability or consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
DG Full-Load Rejection Test
The DGs are required to operate in response to a loss of offsite
power. Their failure cannot initiate an accident. Additionally, the
proposed change to Surveillance Requirement 4.8.1.1.2.g.3 does not
affect the operation or response of any plant structure, system, or
component, and it does not introduce any new failure mechanisms.
Pressure Test of the DG Fuel Oil System
The proposed change to Surveillance Requirement 4.8.1.1.2.i.2 does
not affect the design or function of the DG fuel oil system. Failure of
the DG fuel oil system would not initiate an accident.
Conclusion
Based on the above, the proposed changes to Surveillance
Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No.
3 Technical Specifications will not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Involve a significant reduction in the margin of safety.
DG Full-Load Rejection Test
NNECO is proposing to modify Surveillance Requirement 4.8.1.1.2.g.3
of the Millstone Unit No. 3 Technical Specifications by changing the
acceptable transient voltage to 5000 volts from 4784 volts. The intent
of the proposal is to permit the DG full load rejection tests to be
conducted at conditions which simulate design basis conditions.
The proposed change does not alter the intent of the surveillance,
does not involve any physical changes to the plant, does not alter the
way any structure, system, or component functions, and does not modify
the manner in which the plant is operated. As such, the proposed change
to Surveillance Requirement 4.8.1.1.2.g.3 will not degrade the ability
of the DGs to perform their intended safety function, and will not
reduce the availability of the DGs.
The bases of Technical Specification 3/4.8, ``Electrical Power
Systems,'' state that the operability of the AC and DC power systems
and associated distribution systems ensure that sufficient power will
be available to supply the safety related equipment required for safe
shutdown and for the mitigation of transients. The proposed change to
the surveillance requirement will increase the effectiveness of the
full load rejection tests.
This will ensure the operability of the DGs. Operable DGs ensure
that the assumptions for the bases of the Millstone Unit No. 3
Technical Specifications are not affected.
Pressure Test of the DG Fuel Oil System
NNECO is proposing to delete Surveillance Requirement 4.8.1.1.2.i.2
from the Millstone Unit No. 3 Technical Specifications. This
surveillance requirement is redundant to the requirements of
Surveillance Requirement 4.0.5 which invokes ASME Section XI.
Additionally, the fuel oil system cannot be tested to the requirements
of Surveillance Requirement 4.8.1.1.2.i.2 because the DG fuel oil tanks
are vented to the atmosphere and this vent path cannot be isolated.
Millstone Unit No. 3 will include the DG fuel oil system pressure
test as an augmented inspection within the Inservice Inspection
program. Inspections will be performed in compliance with the
requirement of the 1983 Edition of ASME Section XI, Table IWD-2500-1,
``Test and Examination Categories.'' Testing (i.e., a system
hydrostatic test) in accordance with ASME Section XI will provide
equivalent assurance of tank and piping integrity.
Conclusion
Based on the above, the proposed changes to Surveillance
Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No.
3 Technical Specifications do not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 8753]] amendment request involves no significant hazards
consideration.
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.
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 18, 1995.
Description of amendment request: The proposed changes to the
technical specifications will increase the minimum required boron
concentration in the boric acid tank (BAT) from 6300 ppm to 6600 ppm.
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
consequences of an accident previously evaluated.
The change affects the minimum required boron concentration in the
BAT. Changes in the tank's boron concentration will not affect the
probability of any plant accident.
An increase in the minimum BAT concentration of 6600 ppm was
recommended by Westinghouse based on their Cycle 6 BORDER evaluation.
The BORDER evaluation conservatively determines the ability to maintain
shutdown margin when the plant is taken from an initial operating
condition of Mode 1 or 2 to a final condition of Mode 5 or 6 using an
assumed minimum BAT concentration. Therefore, the ability to maintain
shutdown margin is assured and the change will not adversely affect the
consequences of any plant accident.
2. Create the possibility of a new or different kind of accident
from any Previously Analyzed.
The change conservatively increases the minimum required boron
concentration in the BAT from 6300 ppm to 6600 ppm. There is no impact
on the operability of plant systems or equipment. Therefore, the change
does not create a malfunction that is different from those previously
evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed increase in the minimum boron concentration in the BAT
provides conservatism in the calculated shutdown margin for Millstone
Unit No. 3. The change does not adversely affect any equipment credited
in the safety analysis. Also, the change does not adversely affect the
probability or consequences of any plant accident, including the
calculated PCT [peak clad temperature] or offsite doses. Therefore,
there is no impact on 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 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.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of amendment requests: January 10, 1995.
Description of amendment requests: The proposed amendments would
revise the Prairie Island Event Monitoring Instrumentation Technical
Specifications and associated Bases to conform to Standard Technical
Specifications for post-accident monitoring.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The primary purpose of post accident monitoring instrumentation is
to display plant variables that provide information to the control room
operators during accident situations. Plant instrumentation was
evaluated for importance for this function when Regulatory Guide 1.97
[''Instrumentation for Light Water Cooled Nuclear Power Plants to
Assess Plant Conditions During and Following an Accident'']
classifications were determined. The Prairie Island Regulatory Guide
1.97 classification of instruments was previously approved by the NRC
on October 18, 1985. This amendment request proposes to base Prairie
Island Technical Specifications on the results of the Regulatory Guide
1.97 evaluation in accordance with the guidance of the industry
standard.
Revising the allowed outage time for these instruments will not
significantly increase the probability or consequences of an accident
since these instruments do not initiate automatic actions, there are
available backup indications and the probability of an event requiring
these instruments to be operable is very low.
Therefore, the probability or consequences of an accident
previously evaluated are not affected by any of the proposed
amendments.
2. The proposed amendment[s] will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The license amendment request proposes to add instruments to the
Technical Specifications which have been previously determined to be
important for post accident monitoring, and to remove instruments from
Technical Specifications which have been previously determined to be
less important for post accident monitoring. This amendment ensures the
control room operators are provided with the instrumentation required
to properly manage an accident situation.
Therefore, based on the above considerations, the possibility of a
new or different kind of accident from any accident previously
evaluated would not be created.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The post accident monitoring functions do not initiate any
automatic actions. The instrumentation to be added to the Event
Monitoring Instrumentation Table was previously recognized through the
Regulatory Guide 1.97 evaluation process as important for post accident
monitoring and would be relied upon if there were an event without this
license amendment. Instrumentation to be removed from Technical
Specifications was previously recognized to be less
[[Page 8754]] important and would not be relied upon very much in an
event. Overall, with the trade-off of adding and deleting
instrumentation, the margin of safety will not be significantly
affected.
The proposed license amendment will increase the allowed outage
time for most of the instruments. Again, these instruments do not
provide automatic actions, they provide indications for monitoring post
accident conditions. All of the instruments have backup or
corroborating indications which could be relied upon if the Technical
Specifications instruments were inoperable. Also, an event requiring
use of these instruments has a very low probability. For these reasons
the proposed changes in allowed outage time will not result in a
significant reduction in the margin of safety.
For these same reasons, the proposed changes in radiation
instrument surveillance requirements will not significantly reduce the
margin of safety.
Overall, a significant reduction in the margin of safety would not
result from this license amendment.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: John N. Hannon.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: January 25, 1995.
Description of amendment request: The proposed Technical
Specification change would replace a specific requirement for the
frequency of Type A tests with a general requirement to perform Type A
tests. The proposed amendment would change Surveillance Requirement
4.6.1.2.a. Specifically, the change would require the performance of
Type A tests (overall containment integrated leak rate tests (ILRTs))
at intervals as specified in 10 CFR 50, Appendix J, instead of on a
specific schedule for performance of ILRTs of ``40 plus or minus 10
months.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated (10 CFR
50.92(c)(1)) because the proposed change merely replaces a prescriptive
schedule for performing ILRTs with a requirement to conduct the ILRTs
on a schedule consistent with the Commission's regulations. The change
does not alter the methodology, frequency, or acceptance criteria for
ILRTs, does not affect the design basis of the containment, and does
not change the post-accident response of the containment.
B. The change does not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because the change does not affect the manner by which the
facility is operated and does not make any changes to existing plant
structures, systems, or components. The proposed change merely replaces
a prescriptive schedule for performing ILRTs with a requirement to
conduct the ILRTs on a schedule consistent with the Commission's
regulations.
C. The change does not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)) because the proposed change does not
affect the manner by which the facility is operated or involve changes
to equipment or features which affect the operational characteristics
of the facility.
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: Exeter Public Library, 47
Front Street, Exeter, NH 03833.
Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One
International Place, Boston, MA 02110-2624.
NRC Project Director: Phillip F. McKee.
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: December 7, 1994.
Description of amendments request: The amendments would provide a
permanent voltage-based steam generator tube repair criteria for both
units. This criteria is based on the guidance contained in the NRC
Proposed Generic Communication (Generic Letter 94-XX), ``Voltage-Based
Repair Criteria for the Repair of Westinghouse Steam Generator Tubes
Affected by Outside Diameter Stress Corrosion Cracking,'' that was
issued for public comment in the Federal Register (59 FR 41520) on
August 12, 1994. The licensee's submittal also includes responses to
and identifies exceptions taken to the draft Generic Letter. The
significant exceptions are: (1) The requirement to reinspect all tubes
if bobbin probe wear exceeds 15%; (2) the 1 x 10-2 limit on the
calculated conditional burst probability; and (3) the need to pull
additional steam generator tubes to evaluate the current condition of
the steam generator tubes. In addition, the operational leakage
requirement for Unit 2 will be modified to reduce the total allowable
primary-to-secondary leakage for any steam generator from 500 gallons
per day (gpd) to 150 gallons per day.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of Farley units in accordance with the proposed
license amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Testing of model boiler specimens for free standing tubes at room
temperature conditions shows burst pressures as high as approximately
5000 psi for indications of outer diameter stress corrosion cracking
with voltage measurements as high as 26.5 volts. Burst testing
performed on pulled tubes with up to 7.5 volt indications show burst
pressures in excess of 5900 psi at room temperature. As stated earlier,
tube burst criteria are inherently satisfied during normal operating
conditions by the presence of the tube support plate. Furthermore,
correcting for the effects of temperature on material properties and
minimum strength levels (as the burst testing was [[Page 8755]] done at
room temperature), tube burst capability significantly exceeds the R.G.
[Regulatory Guide] 1.121 criterion requiring the maintenance of a
margin of 1.43 times the steam line break pressure differential on tube
burst if through-wall cracks are present without regard to the presence
of the tube support plate. Considering the existing data base, this
criterion is satisfied with bobbin coil indications with signal
amplitudes over twice the 2.0 volt voltage-based repair criteria,
regardless of the indicated depth measurement. This structural limit is
based on a lower 95% confidence level limit of the data. The 2.0 volt
criterion provides an extremely conservative margin of safety to the
structural limit considering expected growth rates of outside diameter
stress corrosion cracking at Farley. Alternate crack morphologies can
correspond to a voltage so that a unique crack length is not defined by
a burst pressure to voltage correlation. However, relative to expected
leakage during normal operating conditions, no field leakage has been
reported from tubes with indications with a voltage level of under 7.7
volts for 3/4 inch tube which correlates to 10 volts for 7/8 inch
tubing (as compared to the 2.0 volt proposed voltage-based tube repair
limit). Thus, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident.
Relative to the expected leakage during accidents (sic) condition
loadings, the accidents that are affected by primary-to-secondary
leakage and steam release to the environment are Loss of External
Electrical Load and/or Turbine Trip, Loss of All AC Power to Station
Auxiliaries, Major Secondary System Pipe Failure, Steam Generator Tube
Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control
Rod Drive Mechanism Housing. Of these, the Major Secondary System Pipe
Failure is the most limiting for Farley in considering the potential
for off-site doses. The offsite doses analyses for the other events
which model primary-to-secondary leakage and steam releases from the
secondary side to the environment assume that the secondary side
remains intact. The steam generator tubes are not subjected to a
sustained increase in differential pressure, as is the case following a
steam line break event. This increase in differential pressure is
responsible for the postulated increase in leakage and associated
offsite doses following a steam line break event. In addition, the
steam line break event results in a bypass of containment for steam
generator leakage. Upon implementation of the voltage-based repair
criteria, it must be verified that the expected distributions of
cracking indications at the tube support plate intersections are such
that primary-to-secondary leakage would result in site boundary dose
within the current licensing basis. Data indicate that a threshold
voltage of 2.8 volts could result in through-wall cracks long enough to
leak at steam line break conditions. Applications of the proposed
repair criteria requires that the current distribution of a number of
indications versus voltage be obtained during the refueling outages.
The current voltage is then combined with the rate of change in voltage
measurement and a voltage measurement uncertainty to establish an end
of cycle voltage distribution and, thus, leak rate during steam line
break pressure differential. The leak rate during a steam line break is
further increased by a factor related to the probability of detection
of the flaws. If it is found that the potential steam line break
leakage for degraded intersections planned to be left in service
coupled with the reduced specific activity levels allowed result in
radiological consequences outside the current licensing basis, then
additional tubes will be plugged or repaired to reduce steam line break
leakage potential to within the acceptance limit. Thus, the
consequences of the most limiting design basis accident are constrained
to present licensing basis limits.
(2) The proposed license amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Implementation of the proposed voltage-based tube support plate
elevation steam generator tube repair criteria does not introduce any
significant changes to the plant design basis. Use of the criteria does
not provide a mechanism that could result in an accident outside of the
region of the tube support plate elevations. Neither a single or
multiple tube rupture event would be expected in steam generator in
which the repair criteria have been applied during all plant
conditions. The bobbin probe signal amplitude repair criteria are
established such that operational leakage or excessive leakage during a
postulate steam line break condition is not anticipated. Southern
Nuclear has previously implemented a maximum leakage limit of 140/150
gpd (Unit 1/Unit 2) per steam generator. The R.G. 1.121 criterion for
establishing operational leakage limits that require plant shutdown are
based upon leak-before-break considerations to detect a free span crack
before potential tube rupture. The 140/150 gpd limit provides for
leakage detection and plant shutdown in the event of the occurrence of
an unexpected single crack resulting in leakage that is associated with
the longest permissible crack length. R.G. 1.121 acceptance criteria
for establishing operating leakage limits are based on leak-before-
break considerations such that plant shutdown is initiated if the
leakage associated with the longest permissible crack is exceeded. The
longest permissible crack is the length that provides a factor safety
of 1.43 against bursting at steam line break pressure differential. A
voltage amplitude of approximately 9 volts for typical outside diameter
stress corrosion cracking corresponds to meeting this tube burst
requirement at the 95% prediction interval on the burst correlation.
Alternate crack morphologies can correspond to a voltage so that a
unique crack length is not defined by the burst pressure versus voltage
correlation. Consequently, typical burst pressure versus throughwall
crack length correlations is used below to define the ``longest
permissible crack'' for evaluating operating leakage limits.
The single through-wall crack lengths that results in tube burst at
1.43 times steam line break pressure differential and steam line break
conditions are about 0.53 inch and 0.84 inch, respectively. Normal
leakage for these crack lengths would range from about 0.4 gallons per
minute to 4.5 gallons per minute, respectively, while lower 95%
confidence level leak rates would range from about 0.06 gallons per
minute to 0.6 gallons per minute, respectively.
An operating leak rate of 140/150 gpd per steam generator has been
implemented. This leakage limit provides for detection of 0.4 inch long
cracks at nominal leak rates and 0.6 inch long cracks at the lower 95%
confidence level leak rates. Thus, the 140/150 gpd limit provides for
plant shutdown prior to reaching critical crack lengths for steam line
break conditions at leak rates less than 95% confidence level and for
three times normal operating pressure differential at less than nominal
leak rates.
Considering the above, the implementation of voltage-based plugging
criteria will not create possibility of a new or different kind of
accident from any previously evaluated.
(3) The proposed license amendment does not involve a significant
reduction in margin of safety.
The use of the voltage-based tube support plate elevation repair
criteria is demonstrated to maintain steam [[Page 8756]] generator tube
integrity commensurate with the requirements of R.G. 1.121. R.G. 1.121
describes a method acceptable to the NRC staff for meeting GDCs
[General Design Criteria] 2, 14, 15, 31, and 32 by reducing the
probability of the consequences of steam generator tube rupture. This
is accomplished by determining the limiting conditions of degradation
of steam generator tubing, as established by inservice inspection, for
which tubes with unacceptable cracking should be removed from service.
Upon implementation of the criteria, even under the worst case
conditions, the occurrence of outside diameter stress corrosion
cracking at the tube support plant elevations is not expected to lead
to a steam generator tube rupture event during normal or faulted plant
conditions. The most limiting effect would be a possible increase in
leakage during a steam line break event. Excessive leakage during a
steam line break event, however, is precluded by verifying that, once
the criteria are applied, the expected end of cycle distribution of
crack indications at the tube support plate elevations would result in
minimal, and acceptable primary to secondary leakage during the event
and, hence, help to demonstrate radiological conditions are less than
an appropriate fraction of the 10 CFR 100 guideline.
The margin to burst for the tubes using the voltage-based repair
criteria is comparable to that currently provided by existing technical
specifications.
In addressing the combined effects of LOCA [loss-of-coolant
accident] + SSE [safe shutdown earthquake] on the steam generator
component (as required by GDC 2), it has been determined that tube
collapse may occur in the steam generators at some plants. This is the
case as the tube support plates may become deformed as a result of
lateral loads at the wedge supports at the periphery of the plate due
to either the LOCA rarefaction wave and/or SSE loadings. Then, the
resulting pressure differential on the deformed tubes may cause some of
the tubes to collapse.
There are two issues associated with steam generator tube collapse.
First, the collapse of steam generator tubing reduces the RCS [reactor
coolant system] flow area through the tubes. The reduction in flow area
increases the resistance to flow of steam from the core during a LOCA
which, in turn, may potentially increase Peak Clad Temperature (PCT).
Second, there is a potential the partial through-wall cracks in tubes
could progress to through-wall cracks during tube deformation or
collapse or that short through-wall indications would leak at
significantly higher leak rates than included in the leak rate
assessments.
Consequently, a detailed leak-before-break analysis was performed
and it was concluded that the leak-before-break methodology (as
permitted by GDC 4) is applicable to the Farley reactor coolant system
primary loops and, thus, the probability of breaks in the primary loop
piping is sufficiently low that they need not be considered in the
structural design basis of the plant. Excluding breaks in the RCS
primary loops, the LOCA loads from the large branch line breaks were
analyzed at Farley and were found to be of insufficient magnitude to
result in steam generator tube collapse or significant deformation.
Regardless of whether or not leak-before-break is applied to the
primary loop piping at Farley, any flow area reduction is expected to
be minimal (much less than 1%) and PCT margin is available to account
for this potential effect. Based on analyses' results, no tubes near
wedge locations are expected to collapse or deform to the degree that
secondary to primary in-leakage would be increased over current
expected levels. For all other steam generator tubes, the possibility
of secondary-to-primary leakage in the event of a LOCA + SSE event is
not significant. In actuality, the amount of secondary-to-primary
leakage in the event of a LOCA + SSE is expected to be less than that
previously allowed, i.e., 500 gpd per steam generator. Furthermore,
secondary-to-primary in-leakage would be less than primary-to-secondary
leakage for the same pressure differential since the cracks would tend
to tighten under a secondary-to-primary pressure differential. Also,
the presence of the tube support plate is expected to reduce the amount
of in-leakage.
Addressing the R.G. 1.83 considerations, implementation of the tube
repair criteria is supplemented by 100% inspection requirements at the
tube support plate elevations having outside diameter stress corrosion
cracking indications, reduced operating leakage limits, eddy current
inspection guidelines to provide consistency in voltage normalization,
and rotating pancake coil inspection requirements for the larger
indications left in service to characterize the principle degradation
mechanism as outside diameter stress corrosion cracking.
As noted previously, implementation of the tube support plate
elevation repair criteria will decrease the number of tubes that must
be taken out of service with tube plugs or repaired. The installation
of steam generator tube plugs or tube sleeves would reduce the RCS flow
margin, thus implementation of the voltage-based repair criteria will
maintain the margin of flow that would otherwise be reduced through
increased tube plugging or sleeving.
Considering the above, it is concluded that the proposed change
does not result in a significant reduction in margin with respect to
plant safety as defined in the Final Safety Analysis Report or any
bases of the plant 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: Houston-Love Memorial 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.
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.
Description of amendments request: The requested changes to the
Technical Specifications (TS) would implement the recommended changes
from Generic Letter 93-05, ``Line Item Technical Specification
Improvements to Reduce Surveillance Requirements for Testing During
Power Operation.'' Specifically, the amendments would implement TS
changes corresponding to the following GL 93-05 line-item improvement
issues: Control Rod Movement Test for Pressurized Water Reactors,
Radiation Monitors, Surveillance of Boron Concentration in the
Accumulator/Safety Injection/Core Flood Tank, Containment Spray System,
Hydrogen Recombiner, and Special Test Exemptions.
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 changes do not involve any change to the configuration or
method [[Page 8757]] of operation of any plant equipment used to
mitigate the consequences of an accident. The changes to the
surveillance requirements will result in an overall improvement in
plant safety by reducing the likelihood of plant trips and subsequent
challenges to safety systems, decreasing equipment degradation due to
excessive testing, reducing radiation exposure to plant personnel,
increasing the availability of safety related equipment, and
eliminating an unnecessary burden on plant personnel. Therefore, the
proposed changes do 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 kind of accident from any accident previously evaluated. The
proposed changes do not involve any change to the configuration or
method of operation of any plant equipment used to mitigate the
consequences of an accident. The relaxation of surveillance tests
curtails the excessive amount of testing that increases wear on the
equipment and reduces the likelihood of plant trips and subsequent
challenges to safety systems. The relaxation also increases the
availability of safety related equipment. Accordingly, no new failure
modes have been defined for any plant system or component important to
safety nor has any new limiting failure been identified as a result of
the proposed changes. 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 changes do not involve a significant reduction in a
margin of safety. The proposed changes eliminate an unnecessary burden
without compromising protection for public health and safety. The
proposed changes were generically analyzed by the NRC as part of a
comprehensive study and presented in NUREG-1366 ``Improvement to
Technical specifications (sic) Surveillance Requirements.'' The NRC
concluded that while some testing at power is essential to verify
equipment and system operability, safety can be improved, equipment
degradation decreased, and unnecessary personnel burden relaxed by
reducing the amount of testing at power. SNC has analyzed plant
operations and made a comparison with the criteria stated in NUREG-1366
for the line-item improvements contained in this request and has found
the NUREG-1366 basis to be consistent with the Farley design and
operation experience. 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 location: Houston-Love Memorial 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.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: December 6, 1994.
Description of amendment request: The proposed change to Technical
Specification 3/4.1.3.2 will delete Surveillance Requirement (SR)
4.1.3.2.2, that presently requires, every 31 days, the movement of at
least 2% of its height for each Axial Power Shaping Rod not fully
withdrawn. The proposed amendment would also change the surveillance
intervals for the following Technical Specifications (TS) in accordance
with the guidance of Generic Letter 93-05, ``Line Item Technical
Specifications Improvements to Reduce Surveillance Requirements For
Testing During Power Operation,'' and NUREG-1366, ``Improvements to
Technical Specifications Surveillance Requirements:''
1. TS 4.1.3.2 for the Movable Control Assemblies ``Group Height--
Safety and Regulating Rod Groups,'' will relax testing requirements
from at least once every 31 days to every 92 days.
2. TS 4.4.6.2, for ``Operational Leakage,'' relaxes the requirement
to leakage test RCS pressure isolation valves prior to MODE 2 whenever
the plant has been in COLD SHUTDOWN for 72 hours to whenever the plant
has been in COLD SHUTDOWN for 7 days.
3. SR 4.5.2.c.2 for TS 4.5.2, ``ECCS Subsystems--Tavg equal to or
greater than 280 deg. F,'' relaxes the inspection requirements for
ensuring no debris in containment from ``at the completion of each
containment entry'' to ``at least once daily.''
4. TS 4.6.2.1.d, for the ``Containment Spray System,'' relaxes the
SR to perform an air or smoke flow test through the spray header and
nozzles from once per 5 years to once per 10 years.
5. TS 4.10.4.2 for ``Special Test Exceptions Shutdown Margin''
relaxes the SR interval for testing rod insertion capability prior to
reducing shutdown margin from 24 hours to 7 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the NRC 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 change does not involve a significant increase in the
probability of an accident previously evaluated nor does it involve a
significant increase in the consequences of an accident previously
evaluated because no change is being made to any accident initiator and
no accident conditions or assumptions used in evaluating the
radiological consequences of an accident are changed. Relaxation of
surveillance requirements is in accordance with GL 93-05, NUREG-1366,
and is compatible with plant operating experience. Deletion of SR
4.1.3.2 is consistent with NUREG-1430, ``Improved Standard Technical
Specifications for B&W Plants.'' No credit is taken in any accident
analysis or mitigation requirements for the Axial Power Shaping Rod
Group.
(2) The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of any new or
different kind of accident from any accident previously evaluated
because no new accident initiators or assumptions are introduced by
these proposed changes. Relaxation of SRs as discussed in GL 93-05 was
evaluated as reducing equipment degradation with no increase in safety
consequences consistent with the maintenance of plant specific
reliability of the equipment and systems affected. Deletion of the SR
to move the Axial Power Shaping Rod Group does not affect the
requirement to verify rod position, and there is no credit taken for
movement of these rods to mitigate an accident.
(3) The proposed changes do not result in a significant reduction
in the margin of safety. [[Page 8758]]
The changes do not involve a significant reduction in the margin of
safety, because the proposed changes affect only surveillance
requirements, do not affect the function of the components and systems
involved, and do not decrease the estimated equipment or system
reliability.
Based on the NRC staff analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Project Director: Leif J. Norrholm.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: December 6, 1994.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 4.0.5, ``Applicability'' and its
associated Bases; TS 3/4.1.2.3, ``Reactivity Control Systems--Makeup
Pump--Shutdown; TS 3/4.1.2.4, ``Reactivity Control Systems--Makeup
Pump--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.'' The proposed change would replace the specific
monthly surveillance requirements associated with the makeup pumps and
boric acid pumps with a surveillance requirement referencing TS 4.0.5,
which references Section XI of the American Society of Mechanical
Engineers Boiler and Pressure Vessel Code for quarterly pump testing
requirements. The proposed change to TS 4.0.5 and its associated Bases
would revise the requirement regarding the NRC's approval of relief
requests to be in accordance with the NRC Staff's recommendation
contained in NUREG-1482, ``Guidelines for Inservice Testing at Nuclear
Power Plants.'' Additionally, TS 4.0.5.a.2 which describes historical
requirements for inservice inspection and testing would be deleted and
TS 4.0.5.a.1 would be renumbered as TS 4.0.5.a.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the NRC Staff has
performed an 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.
Operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in
accordance with these changes, would not involve a significant increase
in the probability of an accident previously evaluated because no
accident initiators, conditions, or assumptions are affected by the
proposed changes to replace the specific monthly surveillance
requirements for the makeup and boric acid pumps with surveillance
requirements referencing TS 4.0.5 (ASME Boiler and Pressure Vessel Code
Section XI requirements) and to delete wording regarding NRC approval
of relief requests. The changes do not involve a significant increase
in the consequences of an accident previously evaluated, because no
accident conditions or assumptions are affected that would increase the
radiological consequences of a previously evaluated accident.
(2) The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not result in any new accident initiators
nor do they alter any accident scenarios. The changes do not create the
possibility of a different kind of accident from any accident
previously evaluated, because the surveillance requirements for the
makeup and boric acid pumps only affect the testing of existing
components, systems, and functions, and do not introduce any new
requirements.
(3) The proposed changes do not result in a significant reduction
in the margin of safety.
The proposed changes do not reduce or adversely affect the
capabilities or reliability of any plant structures, systems or
components. Relaxation of the surveillance testing interval for the
boric acid and makeup pumps and modifying the testing requirements is
consistent with previous NRC guidance.
Based on this NRC staff evaluation, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Project Director: Leif J. Norrholm.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: January 13, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) by relocating Tables 3.3-2,
``Reactor Trip System Instrumentation Response Times,'' and 3.3-5,
``Engineered Safety Features Response Times,'' to FSAR Chapter 16,
Section 16.3. The Bases discussion specific to Table 3.3-5 would also
be relocated to FSAR Section 16.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed revision does not involve a significant hazards
consideration because operation of Callaway Plant with this change
would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Overall protection system performance will remain within the bounds
of the accident analyses documented in FSAR Chapter 15, WCAP-10961-P,
and WCAP-11883 since no changes to the response times or measurement
interval are proposed.
The RTS and ESFAS will continue to function in a manner consistent
with the above analysis assumptions and the plant design basis. As
such, there will be no degradation in the performance of nor an
increase in the number of challenges to equipment assumed to function
during an accident situation.
These Technical Specification revisions do not involve any hardware
changes nor do they affect the probability of any event initiators.
There will be no change to normal plant operating parameters or
accident mitigation capabilities. Therefore, there will be no increase
in the probability or consequences of any accident occurring due to
these changes.
(2) Create the possibility of a new or different kind of accident
from any previously evaluated. [[Page 8759]]
As discussed above, there are no hardware changes associated with
these Technical Specification revisions nor are there any changes in
the method by which any safety-related plant system performs its safety
function. The normal manner of plant operation is unaffected.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result of
these changes. There will be no adverse effect or challenges imposed on
any safety-related system as a result of these changes. Therefore, the
possibility of a new or different type of accident is not created.
(3) Involve a significant reduction in a margin of safety.
No response time changes are proposed in this amendment
application; only the document where these limits are listed will be
changed. There will be no effect on the manner in which safety limits
or limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment of
protection functions. There will be no impact on DNBR limits, FQ,
F-delta-H, LOCA PCT, peak local power density, or any other margin of
safety.
Based upon the preceding information, it has been determined that
the proposed changes to the Technical Specifications do not involve a
significant increase in the probability or consequences of an accident
previously evaluated, create the possibility of a new or different kind
of accident from any accident previously evaluated, or involve a
significant reduction in a margin of safety. Therefore, it is concluded
that the proposed changes meet the requirements of 10CFR50.92(C) [sic]
and do not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Project Director: Leif J. Norrholm.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: December 8, 1994.
Description of amendment request: The proposed amendment would
change Standby Gas Treatment Power Supply Requirements during refueling
operations.
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:
SGTS [Standby Gas Treatment System] DURING REFUELING OPERATIONS
(Specification 3.7.B.1, 3.7.B.3)
1. The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The Standby Gas Treatment System (SGTS) is not the initiator of any
accident. SGTS may be required to operate for a design basis loss of
coolant accident or for a refueling accident in order to mitigate the
consequences of said accident by providing a filtered exhaust path to
minimize the potential release of radioactive material to the environs.
The proposed amendment does not reduce or change the operational
requirements for the SGTS for an accident. The proposed amendment now
clearly defines the operability requirements during refueling
conditions. The proposed amendment further requires the availability of
a second auxiliary power supply in the event that an Emergency Diesel
Generator (EDG) is out of service during refueling operations, not
currently required. We conclude, therefore, that the proposed amendment
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The SGTS is not an accident initiator, therefore, the proposed
amendment will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment will not involve a significant reduction
in a margin of safety. The proposed amendment requires the availability
of a second auxiliary power supply in the event that an EDG is out of
service during refueling operations, not currently required.
Maintaining availability of a specific reliable auxiliary electrical
power source as an alternative to an EDG in this mode provides
assurance that SGTS can, if required, be operated without placing undue
constraints on EDG availability and represents an enhancement that
increases a margin of safety. We conclude, therefore, that the proposed
amendment does not involve a significant reduction in a margin of
safety.
Based on the above discussion, we have determined that this change
does not constitute a significant hazards consideration as defined in
10CFR50.92(c).
LABORATORY CARBON SAMPLE ANALYSIS (Specification 3.7.B.2.b)
1. The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The Standby Gas Treatment System (SGTS) is not the initiator of any
accident. SGTS may be required to operate for a design basis loss of
coolant accident or for a refueling accident in order to mitigate the
consequences of said accident by providing a filtered exhaust path to
minimize the potential release of radioactive material to the environs.
The proposed amendment does not reduce or change the operational
requirements for the SGTS for an accident. The proposed amendment now
clearly defines the operability requirements during the interval
between sample removal and completion of laboratory analysis.
2. The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The SGTS is not an accident initiator, therefore, the proposed
amendment will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment will not involve a significant reduction
in a margin of safety. The proposed change does not reduce the
requirements or acceptance criteria for sampling, testing or analysis.
The proposed change only incorporates into the specification an
existing clarification which addresses the determination of operability
during the time between sample removal and completion of laboratory
analysis. The change provides an explicit time limit consistent with
current regulatory criteria for completion of analyses.
Based on the above discussion, we have determined that this change
does not constitute a significant hazards [[Page 8760]] consideration
as defined in 10CFR50.92(c).
TORUS VENT MODE (Specification 4.7 B.2.c)
1. The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The Standby Gas Treatment System (SGTS) is not the initiator of any
accident. SGTS may be required to operate for a design basis loss of
coolant accident or for a refueling accident in order to mitigate the
consequences of said accident by providing a filtered exhaust path to
minimize the potential release of radioactive material to the environs.
The proposed amendment does not reduce or change the operational
requirements for the SGTS for an accident.
2. The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The SGTS is not an accident initiator, therefore, the proposed
amendment will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment will not involve a significant reduction
in a margin of safety. The proposed change will incorporate into the
specification an existing clarification. Use of the SGTS filters during
Torus venting results in an insignificant flow through the filters.
Further, maintaining humidity control prevents any adsorber
degradation. Past sample testing on a six month calendar interval when
720 hours operating time has not accumulated has shown no detectable
impact.
Based on the above discussion, we have determined that this change
does not constitute a significant hazards consideration as defined in
10CFR50.92(c).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301.
Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,
One International Place, Boston, Massachusetts 02110-2624.
NRC Project Director: Walter R. Butler.
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.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: November 19, 1994.
Brief description of amendment: The proposed amendment would revise
Section 3.10.8 and the associated Bases of the Indian Point Nuclear
Generating Unit No. 3 Technical Specifications. Specifically, the
proposed revision would reduce the maximum allowable control rod drop
time from 2.4 to 1.8 seconds. The change would remove, for testing
purposes, the allowance for a seismic event (0.6 seconds), which had
been integral to the 2.4 second safety analysis basis. Since a seismic
event cannot be simulated during the rod drop time test, the more
conservative testing acceptance criteria value of 1.8 seconds is needed
to ensure that the plant is within its design basis. This proposed
revision will support control rod testing which is required during
startup from the current outage.
Date of publication of individual notice in Federal Register:
January 20, 1995 (60 FR 4203).
Expiration date of individual notice: February 21, 1995.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of application for amendments: November 30, 1994.
Brief description of amendments: These amendments relocate Table
3.3-2, ``Reactor Protective Instrumentation Response Times,'' and Table
3.3-5, ``Engineered Safety Features Response Times,'' of TS 3/4.3.1 and
3/4.3.2, respectively, to the Palo Verde Updated Final Safety Analysis
Report (UFSAR) in accordance with the guidance provided in Generic
Letter 93-08. In addition, the amendments make administrative changes
to two previous TS amendment requests to maintain consistency with the
deletion of Tables 3.3-2 and 3.3-5. The amendments also delete an
obsolete footnote on page 3/4 3-17 of the Palo Verde Unit 2's TS.
[[Page 8761]]
Date of issuance: February 3, 1995.
Effective date: February 3, 1995.
Amendment Nos.: 88, 75 and 59.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
496) The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
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: November 16, 1994.
Brief description of amendments: The proposed amendments change the
Technical Specifications to revise the wording for the containment
integrated leakage rate testing in Section 3/4.6.1.2 to make it
consistent with the requirements of the BWR-4 Improved Standard
Technical Specifications (NUREG-1433).
Date of issuance: January 26, 1995.
Effective date: January 26, 1995.
Amendment Nos.: 173 and 204.
Facility Operating License Nos. DPR-71 and DPR-62.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65810).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 26, 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.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. STN
50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: June 13, 1994, as supplemented
on October 7, 1994.
Brief description of amendments: The amendments revise the
administrative controls in Section 6 of the technical specifications
(TS). The changes include: (1) a change to the submittal frequency of
the Radiological Effluent Release Report from semiannually to annually;
(2) changes to the Shift Technical Advisor (STA) description; (3) a
clarification of the Shift Engineer responsibilities; and (4) several
editorial changes.
Date of issuance: February 2, 1995.
Effective date: February 2, 1995.
Amendment Nos.: 69, 69, 59 and 59.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53839).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 2, 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.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of application for amendment: May 17, 1993 as supplemented
October 12, 1994.
Brief description of amendment: The amendment replaces License
Condition 2.C.4, relating to the implementation and maintenance of the
approved Fire Protection Program, in its entirety with a new License
Condition. In conjunction, with this change, and in accordance with GL
86-10, Technical Specification provisions related to the Fire
Protection Program are being deleted and placed in the Updated Final
Safety Analysis Report.
Date of Issuance: February 1, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 179.
Facility Operating License No. DPR-61. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36432).
The October 12, 1994, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 1, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457.
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
application for amendments: August 25, 1994, as supplemented November
16, 1994.
Brief description of amendments: The amendments revise Technical
Specification Table 3.3-4, by revising the ``Trip Setpoint'' and
``Allowable Value'' for the 4 kV bus undervoltage grid degraded voltage
relays and the ``Allowable Value'' for the 4 kV undervoltage loss of
voltage/loss of offsite power relays. This revision was submitted in
response to a concern identified by the licensee in their Self-
Initiated Technical Audit and during the electrical distribution system
functional inspection team findings.
Date of issuance: January 20, 1995.
Effective date: To be implemented within 30 days from the date of
issuance.
Amendment Nos.: 127 and 121.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 1994 (59 FR
51619).
The November 16, 1994, letter provided clarifying information that
did not change the scope of the August 25, 1994, 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 20, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of application for amendment: July 28, 1994.
Brief description of amendment: This amendment revises Technical
Specifications 3/4.4.13 to incorporate Low Temperature Overpressure
Protection requirements similar to those recommended by the NRC staff
via Generic Letter 90-06. [[Page 8762]]
Date of Issuance: January 27, 1995.
Effective Date: January 27, 1995.
Amendment No.: 132.
Facility Operating License No. DPR-67: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42341).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 27, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
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: February 3, 1994.
Brief description of amendments: The amendments relocate the
requirements of Technical Specification 3/4.7.10, Area Temperature
Monitoring, to section 16.3 of the VEGP Final Safety Analysis Report
(FSAR). With this relocation to the FSAR, GPC plans to clarify the
basis for areas to be monitored and modify these surveillance
requirements. This change is in accordance with NUREG-1431, ``Standard
Technical Specifications, Westinghouse Plants.''
Date of issuance: January 23, 1995.
Effective date: To be implemented within 30 days from the date of
issuance.
Amendment Nos.: 83 and 61.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 2, 1994 (59
FR 45735).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 23, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: May 13, 1992.
Brief description of amendment: The amendment changes the allowable
primary-to-secondary leakage rate, as specified in License Condition
2.c.(8)2, from 0.1 gallons per minute (gpm) to 0.2 gpm.
Date of Issuance: January 31, 1995.
Effective date: January 31, 1995.
Amendment No.: 193.
Facility Operating License No. DPR-50. Amendment revises a License
Condition.
Date of initial notice in Federal Register: October 14, 1992 (57 FR
47137).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated January 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 26, 1994.
Brief description of amendment: The amendment revised Technical
Specification 3.5.C.1 and 3.5.C.4 to increase the minimum pressure at
which the high pressure coolant injection system is required to be
operable from 113 psig to 150 psig.
Date of issuance: January 25, 1995.
Effective date: January 25, 1995.
Amendment No.: 166.
Facility Operating License No. DPR-46. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53841). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 25, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 22, 1994.
Brief description of amendment: The amendment revised Technical
Specification 1.0.J, definition of limiting conditions for operation,
consistent with the guidance provided in NRC Generic Letter 87-09,
``Sections 3.0 and 4.0 of the Standard Technical Specifications on the
Applicability of Limiting Conditions for Operation and Surveillance
Requirements.''
Date of issuance: February 3, 1995.
Effective date: February 3, 1995.
Amendment No.: 168.
Facility Operating License No. DPR-46. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 3, 1995 (60 FR
153).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305.
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: July 21, 1994.
Brief description of amendment: The amendment revises Technical
Specifications 2.2.2, 3.2.8, 4.2.8, and the associated Bases to reduce
the number of reactor head safety valves required operable from 16
valves to 9 valves. The setpoints of the valve groups are unchanged by
this amendment. The amendment requires testing of the safety valves in
accordance with the approved NMP-1 Inservice Test Program.
Date of issuance: January 25, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 152.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45027).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 25, 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.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York.
Date of application for amendment: October 28, 1994.
Brief description of amendment: The amendment revises Technical
[[Page 8763]] Specification (TS) 1.7, ``CORE ALTERATION,'' to state
that movement or replacement of incore instrumentation is not
considered to be a CORE ALTERATION and that movement of control rods is
not considered a CORE ALTERATION provided there are no fuel assemblies
in the associated core cell. This amendment includes changes to TS 3/
4.9.3, ``Control Rod Position,'' and associated Bases to be consistent
with the revision to TS 1.7. TS 3/4.9.3 is being revised to require
that all control rods be inserted only during loading of fuel
assemblies into the core rather than during CORE ALTERATIONS. These
changes are consistent with the NRC's, ``Improved Standard Technical
Specifications,'' (NUREG-1434).
This amendment also revises Item 1.i.3) of TS Tables 3.3.2-1 and
4.3.2.1-1 to delete the requirement for Reactor Water Cleanup isolation
due to actuation of the Standby Liquid Control System (SLCS) in
OPERATIONAL CONDITION 5. License Amendment No. 48 issued on September
30, 1993, deleted the requirement for the SLCS to be OPERABLE in
OPERATIONAL CONDITION 5; however, due to an oversight, Item 1.i.3) and
associated notations were not deleted from TS Tables 3.3.2-1 and
4.3.2.1-1 as part of License Amendment No. 48. This amendment corrects
that oversight.
Date of issuance: January 20, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 61.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 23, 1994 (59
FR 60382).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 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.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: November 14, 1994.
Brief description of amendment: The amendment revises Technical
Specification 4.5.1.e.2.e) to reduce the leak rate test pressure for
the Automatic Depressurization System (ADS) nitrogen receiving tanks
from 385 psig to 365 psig.
Date of issuance: January 31, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 62.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65817).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 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.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: January 14, 1994, as modified by letter
dated October 17, 1994.
Description of amendment request: The amendment revises the
Appendix A Technical Specifications (TS) to specify the composition of
the Station Operation Review Committee (SORC) based on experience and
expertise vice organizational position, to implement a Station
Qualified Reviewer Program (SQRP), and to revise the time within which
the Nuclear Safety Audit Review Committee (NSARC) must issue reports
and minutes.
The amendment also incorporated a number of editorial changes to
delete certain items that are no longer applicable; remove
inconsistencies involving the names of systems, equipment and NSARC
function, composition, and use of alternates; and correct the value for
the reactor coolant system volume. Other editorial changes have been
incorporated for document format consistency. The amendment affects the
following: TS Sections 1.31, 3.3.3.6, 3.4.1.2, 4.6.3.2, 3.7.1.2, 3/4
10.6, 5.4.2, 6.3.1, 6.4, 6.7, and 6.8.1.4, and Table 4.3-1.
Date of issuance: January 26, 1995.
Effective date: January 26, 1995.
Amendment No.: 34.
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27057) The licensee's letter dated October 17, 1994, provided
clarification and minor revision to the application but does 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 26, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library, 47
Front Street, Exeter, NH 03833.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: July 22, 1994.
Brief description of amendment: The amendment revises the Technical
Specifications to incorporate a different setpoint and transient
methodology for determining the maximum allowable power range neutron
flux setpoint. These changes allow Millstone Unit 3 to operate with a
reduced number of main steam-line safety valves at a reduced power
level, as determined by the high flux setpoint.
Date of issuance: January 31, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 102.
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 14, 1994 (59
FR 47171).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 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.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: October 25, 1994.
Brief description of amendments: These amendments add to the
Susquehanna, Units 1 and 2, Technical Specifications, isolation signals
to Table 3.6.3-1 for the containment isolation valves on the sample
lines for the containment radiation monitoring and
[[Page 8764]] wetwell sample lines. This change is based on the
licensee's design change for installation of a new CRM and wetwell
sample system.
Date of issuance: January 31, 1995.
Effective date: January 31, 1995.
Amendment Nos.: 141 and 111.
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 7, 1994 (59 FR
63126). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Philadelphia Electric Company, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania.
Date of application for amendment: June 10, 1994, as supplemented
by letter dated December 19, 1994.
Brief description of amendment: This amendment involves a one-time
change affecting the Allowed Outage Time (AOT) for the Emergency Sevice
Water (ESW) system, Residual Heat Removal Service Water (RHRSW) System,
the Suppression Pool Cooling, the Suppression Pool Spray, and Low
Pressure Coolant Injection modes of the Residual Heat Removal System,
and Core Spray System to be extended from 3 and 7 days to 14 days
during the Unit 2 refueling outage scheduled to begin in January 1995.
This proposed extended AOT allows adequate time to install isolation
valves and cross-ties on the ESW and RHRSW Systems to facilitate future
inspections or maintenance.
Date of issuance: January 27, 1995.
Effective date: January 27, 1995.
Amendment No. 86.
Facility Operating License No. NPF-39. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37077). The December 19, 1994 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 27, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania
Date of application for amendment: June 30, 1994.
Brief description of amendment: This amendment removes the controls
for a remote shutdown system control valve and the primary containment
isolation valves from TS Tables 3.3.7.4-1 and 3.6.3-1 respectively, as
a result of eliminating the steam condensing mode of the Residual Heat
Removal system.
Date of issuance: January 27, 1995.
Effective date: January 27, 1995.
Amendment No. 47.
Facility Operating License No. NPF-85. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42343).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 27, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania
Date of application for amendment: August 27, 1993, supplemented by
letter dated November 17, 1993.
Brief description of amendment: The amendment allows an expanded
operating domain for the Limerick Generating Station (LGS), Unit 2,
resulting from the implementation of the Average Power Range Monitor--
Rod Block Monitor Technical Specifications/Maximum Extended Load Line
Limit Analysis. These improvements are a prerequisite for Power Rerate
Program implementation at Limerick Generating Station, Unit 2.
Date of issuance: January 27, 1995.
Effective date: January 27, 1995.
Amendment No. 48.
Facility Operating License No. NPF-85. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 13, 1993 (58 FR
52992). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 27, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: August 22, 1994.
Brief description of amendments: These amendments revise TS 3.1.5,
``Standby Liquid Control System,'' to remove the requirement for the
standby liquid control system to be operable in OPERATIONAL CONDITION
5, Refueling, when any control rod is withdrawn and the TS definition
of CORE ALTERATION to exclude control rod movement in a control cell
that contains no fuel assemblies.
Date of issuance: January 27, 1995.
Effective date: January 27, 1995.
Amendment Nos. 87/49.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55881).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 27, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: September 28, 1993.
Brief description of amendment: The amendment revised Technical
Specification (TS) Section 4.11.D to change the surveillance
requirements for the Emergency Service Water System pumps. The change
added pump flow rate requirements and tests the pumps in accordance
with the licensee's Inservice Testing Program. The respective TS Bases
were also revised.
Date of issuance: January 30, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 223.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 24, 1993 (58
FR 62156). [[Page 8765]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 30, 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.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: January 3, 1995 (TS 95-01).
Brief description of amendments: The amendments add a permissive
statement to Surveillance Requirement 4.9.7.1 that will allow the
auxiliary building bridge crane interlocks and physical stops to be
defeated during implementation of the spent fuel pool storage capacity
increase modification.
Date of issuance: January 24, 1995.
Effective date: January 24, 1995.
Amendment Nos.: 194 and 185.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: January 9, 1995 (60 FR
2404) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 24, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration amd
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 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 March 17, 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 [[Page 8766]] with particular
reference to the following factors: (1) the nature of the petitioner's
right under the Act to be made a party to the proceeding; (2) the
nature and extent of the petitioner's property, financial, or other
interest in the proceeding; and (3) the possible effect of any order
which may be entered in the proceeding on the petitioner's interest.
The petition should also identify the specific aspect(s) of the subject
matter of the proceeding as to which petitioner wishes to intervene.
Any person who has filed a petition for leave to intervene or who has
been admitted as a party may amend the petition without requesting
leave of the Board up to 15 days prior to the first prehearing
conference scheduled in the proceeding, but such an amended petition
must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, 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).
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 26, 1994, as supplemented by
letters dated December 27, 1994, and January 27, 1995.
Brief description of amendment: The amendment changed the Technical
Specification Section 3/4.12.A to allow for increased flow capacity of
the control room emergency filter system. By increasing the maximum
allowed makeup capacity of this system, additional margin is provided
for the positive pressurization of the control room envelope.
Date of issuance: January 27, 1995.
Effective date: January 27, 1995.
Amendment No.: 167.
Facility Operating License No. DPR-46. 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
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305.
Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499.
NRC Project Director: William D. Beckner.
Dated at Rockville, Maryland, this 8th day of February 1995.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 95-3629 Filed 2-14-95; 8:45 am]
BILLING CODE 7590-01-P
