[Federal Register Volume 62, Number 250 (Wednesday, December 31, 1997)]
[Notices]
[Pages 68303-68323]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-33968]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Pub. L. 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 December 6, 1997, through December 18, 1997.
The last biweekly notice was published on December 17, 1997 (62 FR
66133).
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 Chief, Rules and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, 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 January 30, 1998, 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
[[Page 68304]]
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, 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 Nos. 50-325 and 50-324]
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: November 26, 1997.
Description of amendments request: Carolina Power & Light Company
(CP&L) has proposed amendments to the Technical Specifications (TS) for
the Brunswick Steam Electric Plant Units 1 and 2 (BSEP 1 & 2) to revise
certain instrumentation allowable values. The revised values were
calculated using a methodology and format consistent with that provided
in NUREG-1433, Revision 1, ``Standard Technical Specifications General
Electric Plants, BWR/4.'' The current TS are based on the uncertainty
associated with the trip unit portion of the instrumentation circuitry.
The proposed values are based on the uncertainty associated with the
entire instrumentation loop (sensor and trip unit). The NRC has
previously approved this methodology for BSEP 1 & 2 as part of a 5
percent power uprate amendment dated November 1, 1996.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes affect accident mitigation instrumentation
allowable values. The changes will not affect the accident
mitigation instrumentation functions. No changes will occur in the
way in which equipment is operated. Therefore, the probability of a
previously evaluated accident can not be affected.
The proposed changes establish the allowable values for certain
functions in accordance with the CP&L setpoint methodology, which
has been approved, by the NRC, for use at the BSEP. The proposed
changes do not affect the actual instrument setpoints. The proposed
allowable values were calculated by applying calibration based
errors to the trip setpoint values; thereby establishing an
operability limit associated with the entire loop of an
instrumentation function to ensure sufficient margin to protect
analytical limits. The changes do not affect the analytical limits
associated with the involved instrumentation functions. The involved
instrumentation will continue to perform its accident mitigation
functions as designed. Therefore, the consequences of a previously
evaluated accident are not increased.
2. The proposed amendments would not create the possibility of a
new or different
[[Page 68305]]
kind of accident from any accident previously evaluated.
The proposed changes do not affect the actual instrument
setpoints nor do they affect the accident mitigation instrumentation
functions. No changes will occur in the way in which equipment is
operated. The involved instrumentation will continue to perform its
accident mitigation functions as designed. Therefore, the proposed
license amendments can not create the possibility of a new or
different kind of accident.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The proposed changes affect accident mitigation instrumentation
allowable values. The changes will not affect the accident
mitigation instrumentation functions. No changes will occur in the
way in which equipment is operated. The proposed changes establish
the allowable values for certain functions in accordance with the
CP&L setpoint methodology which has been approved, by the NRC, for
use at the BSEP. The proposed allowable values were calculated by
applying calibration based errors to the trip setpoint values;
thereby establishing an operability limit associated with the entire
loop of an instrumentation function to ensure sufficient margin to
protect analytical limits. The changes do not affect the analytical
limits associated with the involved instrumentation functions. The
involved instrumentation will continue to perform its accident
mitigation functions as designed. Therefore, the proposed license
amendments 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: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: James E. Lyons.
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: October 29, 1997.
Description of amendment request: Technical Specifications (TS)
3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2 presently require a plant
shutdown and declaring the redundant required feature inoperable, when
the required feature powered from the operable A.C. source is
inoperable. The proposed change clarifies the intent of this TS to
permit the applicable redundant required feature TS to direct a plant
shutdown when required. The proposed amendment changes the existing TS
3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2 to eliminate the separate
requirement for plant shutdown and instead allows the applicable
required redundant feature TS to direct the plant shutdown when
required.
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:
This change does not involve a significant hazards consideration
for the following reasons:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment will not introduce any new equipment or
require existing equipment to function different from that
previously evaluated in the Final Safety Analysis Report (FSAR) or
TS. The changes are consistent with NUREG-1431 and the Commission's
Final Policy Statement on Technical Specification improvements.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment will not introduce any new equipment or
require existing equipment to function different from that
previously evaluated in the Final Safety Analysis Report (FSAR) or
TS. The changes are consistent with NUREG-1431 and the Commission's
Final Policy Statement on Technical Specification improvements. The
proposed amendment will not create any new accident scenarios,
because the change does not introduce any new single failures,
adverse equipment or material interactions, or release paths.
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.
Margin of safety for acceptable TS action times have been
determined for each TS related system. The proposed change will not
alter individual system TS action times. HNP [the Harris Nuclear
Plant] proposes to change the requirement to shutdown after
expiration of the completion time of an inoperable A.C. source
concurrent with an inoperable required feature. Instead of requiring
a shutdown, the required feature on the inoperable A.C. source will
be declared inoperable and the individual TS will be implemented.
In most cases with both redundant features inoperable, a plant
shutdown will be required by TS 3.0.3. In the few instances where
additional time is allowed by the individual TS for both redundant
required features being inoperable, then an immediate plant shutdown
would not be required. The allowed out of service time for loss of
individual safety functions has been previously analyzed for HNP TS
and NUREG-1431, Revision 1.
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: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: James E. Lyons.
Florida Power and Light Company, et al.
[Docket Nos. 50-335 and 50-389]
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: December 1, 1997.
Description of amendment request: The proposed amendment revises
the Unit 1 and Unit 2 Environmental Protection Plans (EPP) Section 4,
``Environmental Conditions,'' and Section 5, ``Administrative
Procedures,'' to incorporate the proposed terms and conditions of the
Incidental Take Statement included in the Biological Opinion issued by
the National Marine Fisheries Service (NMFS) on February 7, 1997. The
proposed amendment also revises the wording in the Unit 1 EPP to make
it consistent with the Unit 2 EPP.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
[[Page 68306]]
probability or consequences of an accident previously evaluated.
The changes are administrative in nature and would in no way
affect the initial conditions, assumptions, or conclusions of the
St. Lucie Unit 1 or Unit 2, accident analyses. In addition, the
proposed changes would not affect the operation or performance of
any equipment assumed in the accident analyses.
Based on the above information, we conclude that the proposed
changes would not significantly increase the probability or
consequences of an accident previously evaluated.
(2) Use of the modified specification would not create the
possibility of a new or different kind of accident from any
previously evaluated.
The changes are administrative in nature and would in no way
impact or alter the configuration or operation of the facilities and
would create no new modes of operation. We conclude that the
proposed changes would not create the possibility of a new or
different kind of accident.
(3) Use of the modified specification would not involve a
significant reduction in a margin of safety.
As indicated in the discussion of Criterion 1, the changes are
administrative in nature and would in no way affect plant or
equipment operation or the accident analysis. We conclude that the
proposed changes would not result in 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: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
IES Utilities Inc.
[Docket No. 50-331]
Duane Arnold Energy Center, Linn County, Iowa
Date of amendment request: October 30, 1996.
Description of amendment request: The proposed amendment, included
as part of the proposed conversion from current Technical
Specifications (CTS) to improved Technical Specifications (ITS), would
modify the Surveillance Requirements (SRs) recommended in NUREG-1433
LOC 3.5.1 by revising the combinations (Conditions C, D, G, and I of
ITS 3.5.1) of emergency core cooling systems/subsystems that may be out
of service. The combinations are supported by the Duane Arnold Energy
Center (DAEC) Loss-of-Coolant Accident (LOCA) analysis.
Condition C
ITS 3.5.1 Action C establishes Required Actions and Completion
Times for the situation when one core spray (CS) subsystem and one or
two residual heat removal (RHR) pump(s) are inoperable. The proposed
specification is less restrictive than CTS 3.5.A.4, which allows one
RHR pump to be inoperable for 30 days, and CTS 3.5.A.5, which allows
two RHR pumps (i.e., the low pressure coolant injection (LPCI)
subsystem) to be inoperable for up to 7 days, provided the remaining
RHR (i.e., LPCI) active components, both CS subsystems, the containment
spray subsystem, and the diesel generators are verified to be operable.
The CTS does not allow one CS subsystem and one or two RHR pump(s) to
be inoperable at the same time. The LOCA analysis presented in NEDC-
31310P, (Duane Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant
Accident Analysis), indicates that an adequate level of protection is
provided by the remaining operable ECCS subsystems. The accident
analysis also demonstrates that in this condition, the peak clad
temperature remains below the regulatory limit. However, another single
failure may place the plant in a condition where adequate core cooling
may not be available during a DBA-LOCA. Therefore, a Completion Time of
72 hours has been proposed to either restore the inoperable CS
subsystem or the inoperable RHR pump(s).
Condition D
ITS 3.5.1 Action D establishes Required Actions and Completion
Times for the situation when two CS subsystems are inoperable. The
proposed specification is less restrictive than CTS 3.5.A.2, which
allows only one CS subsystem to be inoperable. CTS 3.5.A.6 would
require the plant to be in Hot Shutdown within 12 hours and Cold
Shutdown within the following 24 hours if both CS subsystems were
inoperable. With two CS subsystems inoperable, the LOCA analysis
presented in NEDC-31310P, (Duane Arnold Energy Center SAFER/GESTR-LOCA
Loss-of-Coolant Accident Analysis), indicates that the remaining
operable low pressure ECCS subsystem consisting of LPCI with four RHR
pumps operable (only 3 pumps required), provides adequate protection.
However, another single failure may place the plant in a condition
where adequate core cooling may not be available during a Design Basis
Accident LOCA. Therefore, a Completion Time of 72 hours has been
proposed to restore one CS subsystem to operable status.
Condition G
ITS 3.5.1 Action G establishes Required Actions and Completion
Times for the situation when HPCI and one RHR pump are inoperable. The
proposed specification is less restrictive than CTS 3.5.D.2, which
allows continued operation if HPCI is inoperable only if both CSs,
LPCI, ADS, and RCIC are verified to be operable. While the LPCI
subsystem is technically operable with only 3 of 4 RHR pumps operable,
the CTS is currently interpreted by DAEC to require all 4 RHR pumps to
be operable for the requirements of CTS 3.5.D.2 to be met, as a single
RHR pump has more makeup capability than the HPCI System. Thus for
mitigating small and intermediate break LOCAs, one LPCI pump, in
combination with ADS, is more than adequate core cooling. The condition
of when HPCI and one RHR pump are inoperable is bounded by the analysis
in NEDC-31310P, Duane Arnold Energy Center, SAFER/GESTR-LOCA Loss-of-
Coolant Accident Analysis. Since the remaining operable low pressure
ECCS subsystems are more than capable of performing their intended
function, and RCIC and ADS are Operable, the proposed Action G
maintains LOCA analysis assumptions for ECCS Operability. The proposed
ITS condition allows 7 days to restore the HPCI System or the RHR pump
to operable status. The licensee considers the 7 day Completion Time
reasonable in that the LOCA analysis demonstrates that in this
condition, the peak clad temperature remains below the regulatory
limit. The 7 day Completion Time also provides the benefit of
potentially avoiding an unnecessary plant shutdown while the safety
functions are still capable of being performed.
Condition I
ITS 3.5.1 Action I establishes Required Actions and Completion
Times for the situation when HPCI and one ADS valve are inoperable. The
proposed Specification is less restrictive than CTS 3.5.D.2, which
allows continued operation if HPCI is inoperable only if both CSs,
LPCI, ADS, and RCIC are verified to be operable. While ADS is capable
of performing its design function with only 3 of 4 valves operable, per
NEDC-31310P, Duane Arnold Energy Center, SAFER/GESTR-
[[Page 68307]]
LOCA Loss-of-Coolant Accident Analysis, the CTS requires all 4 ADS
valves to be operable for the requirements of CTS 3.5.D.2 to be met.
The proposed specification is less restrictive than CTS 3.5.F.2, which
allows continued operation when one ADS valve is inoperable only if
HPCI is verified to be operable. Since all low pressure ECCS subsystems
remain capable of performing their design function and ADS is still
capable of performing its design function, ITS 3.5.1 Action I maintains
LOCA assumptions to ensure an adequate level of protection is
maintained. The proposed condition allows 72 hours to restore the HPCI
system or the ADS valve to operable status, since another single
failure (i.e., loss of another ADS valve), may place the plant in a
condition where adequate core cooling may not be available during a
small or intermediate break LOCA.
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:
For Condition C
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change will allow one Core Spray subsystem and one
or two RHR pump(s) to be inoperable for up to 72 hours. The ECCS
subsystems affected by this change are not assumed to be initiators
of analyzed events. Therefore, the proposed change does not increase
the probability of any accident. The role of these ECCS subsystems
is in the mitigation of accident consequences. The proposed change
does not allow unlimited continuous operation with the plant in a
condition where an additional single failure could result in a loss
of ECCS function. The proposed change does not increase the
consequences of an accident because accident analysis presented in
NEDC-31310P, Duane Arnold Energy Center SAFER/GESTR-LOCA Loss-of-
Coolant Accident Analysis, indicates that an adequate level of
protection is maintained by the ADS System and the remaining
Operable ECCS subsystems when one Core Spray subsystem and one or
two RHR pump(s) are inoperable. Therefore, this change will not
involve a significant increase in the consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change will not involve any physical changes to
plant systems, structures, or components (SSCs), or the manner in
which these SSCs are operated, maintained, modified, tested or
inspected. The change ensures the remaining ECCS capability is
adequate to mitigate the consequences of accidents. Therefore, this
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change does not significantly reduce the margin of
safety because accident analysis presented in NEDC-31310P, Duane
Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident
Analysis, indicates that the plant is protected by the ADS System
and the remaining ECCS subsystems when one Core Spray subsystem and
one or two RHR pump(s) are inoperable. The accident analysis
demonstrates that in this condition, the peak clad temperature
remains below the regulatory limit. However, with one Core Spray
subsystem and one or two RHR pump(s) inoperable, another single
failure may place the plant in a condition where adequate core
cooling may not be available during a DBA-LOCA. Therefore, a
Completion Time of 72 hours has been assigned to either restore the
inoperable Core Spray subsystem or the RHR pump. In addition, this
change provides the benefit of potentially avoiding an unnecessary
plant shutdown (due to a Completion Time being provided for one Core
Spray subsystem and one or two RHR pump(s)) when the remaining ECCS
subsystems and the ADS are capable of mitigating potential events.
Therefore, this change does not involve a significant reduction in a
martin safety.
For Condition D
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change will allow both Core Spray subsystems to be
inoperable for up to 72 hours. The ECCS subsystems affected by this
change are not assumed to be initiators of analyzed events.
Therefore, the proposed change does not increase the probability of
any accident. The role of these ECCS subsystems is in the mitigation
of accident consequences. The proposed change does not allow
unlimited continuous operation with the plant in a condition where
an additional single failure could result in a loss of ECCS
function. The proposed change does not increase the consequences of
an accident because accident analysis presented in NEDC-3131OP,
Duane Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident
Analysis, indicates that an adequate level of protection is
maintained by the ADS System and remaining Operable ECCS subsystem
when two Core Spray subsystems or inoperable. Therefore, this change
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change will not involve any physical changes to
plant systems, structures, or components (SSCs), or the manner in
which these SSCs are operated, maintained, modified, tested, or
inspected. The change ensures the remaining ECCS capability is
adequate to mitigate the consequences of accidents. Therefore, this
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change does not significantly reduce the margin of
safety because accident analysis presented in NEDC-31310P, Duane
Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident
Analysis, indicates that the plant is protected by the ADS System
and the remaining ECCS subsystem when two Core Spray subsystems are
inoperable. The accident analysis demonstrates that in this
condition, the peak clad temperature remains below the regulatory
limit. However, with both Core Spray subsystems inoperable, another
single failure may place the plant in a condition where adequate
core cooling may not be available during a DBA-LOCA. Therefore, a
Completion Time of 72 hours has been assigned to restore one
inoperable Core Spray subsystem. In addition this change provides
the benefit of potentially avoiding an unnecessary plant shutdown
(due to a Completion Time being provided for both Core Spray
subsystems inoperable) when the remaining ECCS subsystem and the ADS
are capable of mitigating potential events. Therefore, this change
does not involve a significant reduction in a margin of safety.
Condition G
1. Does the change involve a significant increase in the
probability or consequences or an accident previously evaluated?
The proposed change will allow the HPCI System and one RHR pump
to be inoperable for up to 7 days. The ECCS subsystems affected by
this change are not assumed to be initiators of analyzed events.
Therefore, the proposed change does not increase the probability of
any accident. The role of these ECCS subsystems is in the mitigation
of accident consequences. The proposed change does not allow
unlimited continuous operation with the plant in a condition where
an additional single failure could result in a loss of ECCS
function. The proposed change does not increase the consequences of
an accident because accident analysis presented in NEDC-31310P,
Duane Arnold Energy Center SAFER/GESTRA-LOCA Loss-of-Coolant
Accident Analysis, indicated that an adequate level of protection is
maintained by the ADS System and the remaining Operable ECCS
subsystems when HPCI and one RHR pump are inoperable. Therefore,
this change will not involve a significant increase in the
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change will not involve any physical changes to
plant systems, structures, or components (SSCs), or the manner in
which these SSCs are operated, maintained, modified, tested, or
inspected. The change ensures the remaining ECCS capability is
adequate to mitigate the consequences of accidents. Therefore, this
change will not create the possibility of a new or different kind of
accident from any accident previously evaluate.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change does not significantly reduce the margin of
safety because accident
[[Page 68308]]
analysis presented in NEDC-31310P, Duane Arnold Energy Center SAFER/
GESTR-LOCA Loss-of-Coolant Accident Analysis, indicates that the
plant is protected by the ADS System and the remaining ECCS
subsystems when HPCI and one RHR pump are inoperable. The accident
analysis demonstrates that in this condition, the peak clad
temperature remains below the regulatory limit. However, with both
HPCI and one RHR pump inoperable, another single failure may place
the plant in a condition where adequate core cooling may not be
available during an accident. Therefore, a Completion Time of 7 days
has been assigned to either restore the inoperable HPCI System or
the RHR pump. In addition, this change provides the benefit of
potentially avoiding an unnecessary plant shutdown (due to a
Completion Time being provided for the HPCI System and one RHR pump
inoperable) when the remaining ECCS subsystems and the ADS are
capable of mitigating potential events. Therefore, this change does
not involve a significant reduction in a margin of safety.
Condtion I
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change will allow the HPCI system and one ADS valve
to be inoperable for up to 72 hours. The ECCS subsystems affected by
this change are not assumed to be initiators or analyzed events.
Therefore, the proposed change does not increase the probability of
any accident. The role of these ECCS subsystems is in the mitigation
of accident consequences. The proposed change does not allow
unlimited continuous operation with the plant in a condition where
an additional single failure could result in a loss of ECCS
function. The proposed change does not increase the consequences of
an accident because accident analysis presented in NEDC-31310P,
Duane Arnold Energy Center SAFER/GESTER-LOCA Loss-of-Coolant
Accident Analysis, indicates that an adequate level of protection is
maintained by the remaining ADS valves (the ADS design function is
maintained) in combination with the remaining Operable ECCS
subsystems when HPCI and one ADS valve are inoperable. Therefore,
this change will not involve a significant increase in the
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or difference
kind of accident form any accident previously evaluated?
The proposed change will not involve any physical changes to
plant systems, structures, or components (SSCs) or the manner in
which these SSCs are operated, maintained, modified, tested, or
inspected. The change ensures the remaining ECCS capability in
adequate to mitigate the consequences of accidents. Therefore, this
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change does not significantly reduce the margin of
safety because accident analysis presented in NEDC-31310P, Duane
Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident
Analysis, indicates that the plant is protected by the remaining ADS
valves and the low pressure ECCS subsystems when HPCI and one ADS
valve are inoperable. The accident analysis demonstrates that in
this condition, the peak clad temperature remains below the
regulatory limit. However, with both HPCI and one ADS valve
inoperable, another single failure (i.e., of an ADS valve) may place
the plant in a condition where adequate core cooling may not be
available during a small or intermediate break LOCA. Therefore, a
Completion Time of 72 hours has been assigned to either restore the
inoperable HPCI System or the ADS valve. In addition, this change
provides the benefit of potentially avoiding an unnecessary plant
shutdown (due to a Completion Time being provided for the HPCI
System and one ADS valve inoperable) when the remaining ECCS
subsystems and ADS valves are capable of mitigating potential
events. Therefore, this change does 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: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
Acting NRC Project Director: Richard P. Savio.
Indiana Michigan Power Company
[Docket Nos. 50-315 and 50-316]
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: August 1, 1997 (AEP:NRC:0906H).
Description of amendment requests: The proposed amendments would
revise Technical Specification surveillance 4.7.1.2.b. to delete the
requirement that the test be performed at a specified secondary steam
supply pressure.
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
The proposed changes will not significantly increase the
probability or consequences of an accident previously evaluated.
This is an administrative change intended to clarify the
technical specification. There will be no change to the test
procedure as a result of this clarification. The proposed change
better correlates with the accident requirements for which TDAFP
[turbine driven auxiliary feed pump] flow is required, and the
change is consistent with the present requirement of testing the
TDAFP at a secondary side pressure greater than 310 psig.
Criterion 2
The proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does not physically modify the plant, nor does
it result in the installation of equipment which could introduce a
new failure mechanism.
Criterion 3
The proposed change does not involve a significant reduction in
a margin of safety. The proposed change does not affect the
performance of the TDAFP. Thus, the TDAFP remains capable of
providing the required flow under accident conditions, and no safety
margins are reduced.
This is an administrative change intended to clarify the
technical specification. There will be no change to the test
procedure as a result of this clarification
.The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Richard P. Savio, Acting.
Indiana Michigan Power Company
[Docket Nos. 50-315 and 50-316]
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: August 11, 1997 (AEP:NRC:1265).
Description of amendment requests: The proposed amendments would
revise the Technical Specifications (TS) to allow the filling of the
emergency core cooling system (ECCS) accumulators without declaring
ECCS equipment inoperable.
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:
[[Page 68309]]
Criterion 1
This amendment request does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the proposed changes to the T/S represent the
possibility of an event that has such a low probability as to not be
considered credible. A calculation was performed that demonstrated
the CDF resulting from the accumulator fill line operation with all
of the conditions assumed above is approximately 3 x
10--10 per year. This is well below the NEI guidelines of
1 x 10-6 for acceptable risk for a given evolution.
Therefore, based on probabilistic considerations and the robust
design of the pumps, we conclude the risk associated with this
proposed change will not result in a significant increase in the
probability or consequences of a previously evaluated accident.
Criterion 2
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The change does not involve a physical change to the plant, but does
involve a change in the plant operating configuration. The
possibility of a LBLOCA [large break loss of coolant accident]
occurring during the accumulation fill evolution has been evaluated
and determined to not be credible. Westinghouse has confirmed the
accumulator fill line was not modeled in the accident analyses due
to the extremely short duration of the fill operation and the
extremely small amount of flow that the fill line is capable of
passing. The overall effect this configuration would have on the
capability of the SI [safety injection] pump to perform its design
function, should a LBLOCA occur during the extremely brief window of
opportunity, is negligible and would not create a new type of
accident.
Criterion 3
This proposed change does not involve a significant reduction in
a margin of safety, as the risk from the postulated sequence of
events is insignificant. Additionally, engineering evaluation has
determined that the real response of an SI pump under the postulated
conditions would not be severe. The rugged construction of the
pumps, and the design margin built into them, are factors that
support the engineering judgment that the affected pump would
continue to operate for some time, at some capacity beyond the
manufacturer's design limit. As a result of exceeding the limit, the
pump may experience some cavitation and require additional
corrective maintenance, but would be expected to deliver a
significant fraction of its design flow.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Richard P. Savio, Acting.
Niagara Mohawk Power Corporation
[Docket No. 50-410]
Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: October 7, 1997.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) to change the setpoints of
Surveillance Requirements (SRs) 4.9.6.a, 4.9.6.f, and 4.9.6.g for the
refueling platform main hoist. Specifically, each refueling platform
crane or hoist used for handling control rods or fuel assemblies within
the reactor pressure vessel would be demonstrated operable by:
a. Demonstrating operation of the overload cutoff on the main hoist
when the load exceeds 1600 +100/-0 pounds (rather than 1200 +50/-50
pounds).
f. Demonstrating operation of the loaded interlock on the main
hoist when the load exceeds 700 +50/-0 pounds (rather than 485 +50/-50
pounds).
g. Demonstrating operation of the redundant loaded interlock on the
main hoist when the load exceeds 700 +50/-0 pounds (rather than 550
+50/-50 pounds).
The proposed amendment, in effect, would authorize replacement of
the existing triangular refueling platform mast with a round, heavier
mast (General Electric Model NF-500) which includes an installed
camera/TV 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, 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 proposed change revises the setpoints for three TS SRs based
on modifications to the refueling platform mast. The new mast is
essentially a direct replacement for the existing mast, with the
exception that the new mast is approximately 400 lbs. heavier, which
directly affects the setpoints. No change in the frequency or manner
in which the surveillances are performed is proposed. Refueling
interlocks will continue to function as designed. No changes to the
methods in which plant systems are operated are required. The same
design criteria and standards were applied to the new mast,
including the seismic capability of the refueling platform with the
heavier mast. Therefore, none of the precursors of previously
evaluated accidents are affected, and no new failure modes are
introduced.
Based on the additional weight of the new mast and camera/TV
system, the revised GESTAR [General Electric GESTAR II document
NEDE-24011-P-A-11-U5] criteria for fuel rod damage (more
conservative threshold level), the use of GE11 [9x9] fuel for the
bundle drop analysis, the number of damaged fuel rods has increased
slightly for the potential fuel handling accident. The results of
this increase were evaluated and dispositioned against the bounding
calculation to show that the current USAR [updated safety analysis
report] analysis bounds the revised radiological consequences which
remain well within the GDC [General Design Criterion] 19 and
10CFR[part]100 limits. The systems that are available to mitigate
the consequences of any accident have not been affected and are
still capable of performing their required functions. Therefore,
this change will not involve a significant increase in the
probability or 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 change revises the setpoints for three TS SRs based
on installation of a new refueling platform which is heavier than
the current mast. No change in the frequency or manner in which the
surveillances are performed has occurred. Refueling interlocks will
continue to function as designed. No changes to the methods in which
plant systems are operated are required. The same design criteria
and standards were applied to the new mast, including the seismic
capability of the refueling platform with the heavier mast. The
basic function and operation of the refueling platform is unchanged.
The uptravel stop and downtravel mechanical cutoff setpoints are not
being changed and will continue to ensure that adequate water
shielding is maintained. As such, the change does not introduce any
new failure modes or conditions that may create a new or different
kind of accident. Therefore, this change 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.
The proposed change revises three TS SR setpoints based on
installation of a new refueling platform mast. No change in the
frequency or manner in which the surveillances are performed has
occurred. Refueling interlocks will continue to function as
designed. No changes to the methods in which plant systems are
operated are required. The same design criteria and standards were
applied to the new mast, including the seismic capability of the
[[Page 68310]]
refueling platform with the heavier mast. The addition of a camera/
TV system will provide enhanced visibility for fuel handling
activities and additional assurance that the grapple is oriented
over the correct fuel bundle.
The additional weight of the new mast has been evaluated and the
operability requirements as described in the TS and TS Bases are
unchanged. The modification and revised setpoints do not change the
function of the refueling platform main hoist. The revised setpoints
will continue to assure the lifting capacity of the main hoist will
not be sufficient to result in damage to core internals or the
reactor pressure vessel in the event that they are accidentally
engaged.
The necessary systems are still available to mitigate any
potential radiological consequences of the increased number of
damaged fuel rods. The radiological consequences remain within the
bounds of the current safety analysis and well below the GDC 19 and
10CFR[Part]100 limits. Therefore, the change does not involve any
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: S. Singh Bajwa.
Niagara Mohawk Power Corporation
[Docket No. 50-410]
Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: October 31, 1997.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) to support installation of the
General Electric Nuclear Measurement Analysis and Control (NUMAC) Power
Range Neutron Monitor (PRNM) System. The TS changes apply to Sections
2.2, ``Limiting Safety System Settings''; 3/4.3.1, ``Reactor Protection
System Instrumentation'' and its corresponding Bases; and 3/4.3.6,
``Control Rod Block Instrumentation.''
Basis for proposed no significant hazards consideration
determination: The NUMAC-PRNM will monitor groups of Local Power Range
Monitor (LPRM) signals and, together with the Oscillation Power Range
Monitor (OPRM), initiate a reactor scram upon identifying neutron flux
oscillations characteristic of a thermal-hydraulic instability. The
NUMAC-PRNM will replace the existing Average Power Range Monitor (APRM)
System and will ultimately support the activation of the OPRM. The
proposed modification is in response to Generic Letter 94-02, ``Long-
Term Solutions and Upgrade of Interim Operating Recommendations for
Thermal-Hydraulic Instabilities in Boiling Water Reactor.'' Except for
minor deviations, the proposed TS changes are consistent with General
Electric Licensing Topical Report (LTR), NEDC-32410P-A, ``Nuclear
Measurement Analysis and Control Power Range Neutron Monitor (NUMAC-
PRNM) Retrofit Plus Option III Stability Trip Function,'' which was
approved by the NRC staff September 5, 1995. Changes with respect to
response time testing requirements would be based on Supplement 1 to
NEDC-32410P-A, approved by the NRC staff December 26, 1996.
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.
As discussed in NEDC-3241OP-A, the NUMAC-PRNM modification and
associated changes to the TS involve systems that are intended to
detect the symptoms of certain events or accidents mitigating
actions. The worst case failure of the systems involved would be a
failure to initiate mitigative actions (i.e., scram or rod block),
but no failure can cause an accident and therefore the probability
of precursors of any accidents previously evaluated is not
increased. The NUMAC-PRNM system performs the same operations as the
existing equipment, reduces the need for tedious operator action
during normal conditions and allows the operator to focus more on
overall plant conditions. Automatic self-test and increased operator
information available with the NUMARC-PRNM system is likely to
reduce the burden during off-normal conditions as well. The NUMAC-
PRNM system is compatible with the environmental conditions at the
mounting location (e.g., temperature, humidity, seismic,
electromagnetic fields) such that system performance will not be
degraded when compared to the system being replaced. Therefore, the
proposed change will not result in a significant increase in the
probability of any accidents previously evaluated.
The proposed changes to the RPS [reactor protection system] and
Control Rod Block instrumentation TSs are necessitated by the NUMAC-
PRNM replacement. As discussed in the evaluation, in the 4 APRM
channel configuration, any two of the four APRM channels and one 2-
out-of-4 voter channel in each RPS trip system are required to
function for the APRM safety trip function to be accomplished.
Therefore, the proposed TS change requires that 3 of the 4 APRM
channels be operable. This assures at least two APRM channels to
each of the 2-out-of-4 voter channels are available in the event of
a single APRM channel failure and one APRM is bypassed. Also, the
proposed TS requires a minimum of two 2-out-of-4 voter channels per
RPS trip system (i.e., all four voter channels). This assures that
at least one voter channel per trip system is available even in the
event of a single voter channel failure. Surveillance testing
requirements were revised to take advantage of certain features of
the NUMAC-PRNM (digital) replacement of the existing analog APRM
system. These advantages included improved accuracy, stability ,
self-testing, reduced drift, and constant time for digital
processing. Testing of the RPS and Control Rod Block instrumentation
will continue to be performed as described in the evaluation to
assure that the reliability and performance of these systems will
not be adversely affected.
The proposed NUMAC-PRNM replacement system has been specifically
designed to assure that the system response times meet the current
acceptance limits (worst case). As a result, due to statistical
variations resulting from the sampling and update cycles, the
response time is typically faster than required in order to assure
the required response time is always met. The architecture of the
NUMAC-PRNM system has reduced segmentation compared to the existing
PRM system. Examples of the reduced segmentation are combining
previously separate functions, several input channels sharing an
input board, and a central loop processor for many channels. The
replacement equipment includes up to 5 LPRM inputs on a single
module compared to one per module on the current system. Up to 17
LPRM signals are processed through one preprocessor. The
recirculation flow signals are processed in the same hardware as the
LPRM processing. The net effect of these architectural aspects is
that there are some single failures that cause a greater loss of
``sub-functionality'' than in the current system. However, other
architectural and functional aspects have an offsetting effect.
Redundant power supplies are used so that a single failure of AC
power has no effect on the overall NUMAC-PRNM system functions while
still resulting in a half scram, as does the current system.
Continuous automatic self-test also assures that if a single failure
does occur, it is much more likely to be detected immediately. The
net effect is that from a total system level, there is no increased
risk of loss of critical functionality or reduction in safety
margins due to the architecture of the replacement system.
Failure analysis indicates that a software common cause failure
is not a significant contributor to the unavailability of the NUMAC-
PRNM. However, in spite of that conclusion, means are provided
within the system to mitigate the effects of such a failure and
alert an operator. Therefore, such a failure, even if it occurred,
will not increase
[[Page 68311]]
the consequences of a previously evaluated accident. To reduce the
likelihood of common cause failures of software controlled
functions, thorough and careful verification and validation (V&V)
activities are performed both for the requirements and the
implementing software design. In addition, the software is designed
to limit the loading that external systems or equipment can place on
the system, thus significantly reducing the risk that some abnormal
dynamic condition external to the system can cause an overload. For
conservatism, however, despite, these V&V activities, common cause
failures of software controlled functions due to residual software
design faults are assumed to occur. Both the software and hardware
are designed to manage the consequences of such failures. Safety
outputs are designed to be fail safe by requiring dynamic update of
output modules or data signals, where failure to update the
information is detected by simple receiving hardware, which in turn,
forces a trip. This aspect covers all but rather complex failures
where the hardware or software executes a portion of the overall
logic but fails to process some portion of the new information
(inputs ``freeze'') or some portion of the logic (outputs
``freeze''). To help reduce the likelihood of complex failures, a
watchdog timer is used which is updated by a very simple software
routine that in turn monitors the operational cycle time of all
tasks in the system. The software design is such that as long as all
tasks are updating at the design rate, it is likely that software
controlled functions are executing as intended. Conversely, if any
task fails too update at the design rate, that is a strong
indication of at least some unanticipated condition. If such a
condition occurs, its watchdog timer will not be updated, the
computer will be restarted, and the outputs will detect an abnormal
condition and provide an alarm.
It is very difficult to quantify a software common cause failure
rate. Analyses for the current system did consider common cause
failures and assessed them to be at a rate of about 0.3 times the
random failure rate. The reference analysis uses a field basis for
the random rates. The analysis for the replacement design uses
conservative estimates for failure rates of equipment that are
actually a little higher than those assumed for the current
equipment. The methodology being applied concludes that the common
mode failure rate for the replacement system is somewhat higher than
the current system. However, that is offset by more frequent
surveillance tests performed by the self-test that result in an
estimated slightly lower unavailability for the NUMAC-PRNM scram
function compared to the current PRM system. The USAR, in general,
considers the failure rate of the function, not that of sub-
components. On that basis, there will not be an increase, due to
software common cause failure, in the probability of a malfunction
analyzed in the USAR.I21The NUMAC-PRNM human-machine interface
design does not introduce an increased burden or constraints on the
operators' ability to adequately respond to an accident such that
there would be more severe consequential effects. The information
available to the operators is the same as with the current system.
No actions are required by the operator to obtain information
normally used and equivalent to that available with the current
equipment. However, the replacement system does provide more direct
accessible information regarding the condition of the equipment,
including automatic self-test, which can aid the operator in
diagnosing unusual situations beyond those defined in the licensing
basis.
The replacement system has a significantly lower power
requirement and is generally smaller, reducing somewhat the seismic
loading on the panels. The equipment qualification also includes EMI
[electro magnetic induction] emissions which, combined with the fact
that the replacement equipment is mounted in its own cabinet
(replaces all of the current equipment), minimized the likelihood of
significant impact on other existing equipment.
The replacement equipment makes increased use of qualified
optical methods to provide both safety and functional isolation
between safety-related and nonsafety-related systems. Where fiber
optic methods cannot be used, the isolation provided is comparable
to or better than that provided in the current system.
The net electrical and thermal load for the replacement system
is less than that for the current system. Accordingly, the
replacement system had adequate cabinet cooling and no forced
cooling is required.
The replacement system meets or exceeds all applicable
requirements for separation, independence and grounding. The use of
fiber optic connections between the APRM and RBM [rod block monitor]
improves the separation and reduces the dependence of the system on
common grounds. However, for noise rejection, the equipment design
and manufacturing requirements assure improved grounding of the
actual equipment.
No change in wiring or grounding external to the panels
containing the replacement equipment is necessary for correct
operation of the replacement equipment.
NEDC-3241OP-A, Section 3.2.3, discusses different plant
configurations for recirculation flow channels, including the case
where plants currently (before implementing the NUMAC PRNM system)
have four flow channels. Absence of any discussion in the LTR
related to separation for plants originally having four flow
channels implies that those plants are expected to meet full
separation requirements. The LTR includes a further statement that
``The criterion is to maintain equal or better protection against
single failures while allowing bypassing of the APRM channel that
processes the flow signal.''
The NMPC [Niagara Mohawk Power Corporation] NUMAC PRNM system
has four recirculation low channels, but the flow input circuits for
two of the four are not separated from each other outside the PRNM
panel. As a result, a single failure that causes both of these flow
signals to go high could, depending on the specific value, cause the
APRM flow biased trip setpoint in two channels to go to the clamped
setpoint. If, at the same time, a third channel is bypassed, the
APRM flow-biased trip setpoint for the APRM system could be non-
conservative. (NOTE: The flow signals are compared to one another.
Should the flow signals not be within specified limits, an alarm and
a control rod block would be initiated.)
Despite the fact that two of the four flow input circuits are
not separated from each other outside the PRNM panel, the
replacement system is judged to be adequate with the current field
routing of flow signals and meets the LTR criteria. This conclusion
is based on the fact that there is no credible fault in the circuits
within the duct, in which the flow signals are routed, that can
damage the other circuits. Also, there is no credible external fault
that can damage the circuits inside the duct. Therefore, it is
concluded that the separation between the two flow input circuits is
adequate to meet the system single failure requirements in that no
credible single failure will disable the flow inputs to more than
one APRM channel. Additionally, there are no reload licensing
transient analyses that take credit for the flow-biased simulated
thermal power scram setpoint.
The replacement design has been specifically designed to have
the same or more conservative ``fail safe'' failure modes as the
current system. For example, in the case of a single power bus
failure, the current system loses about one half of the LPRM
information and an output trip occurs. For the replacement system,
that failure still results in an output trip, but no LPRM
information is lost. In the current system, a static failure in
several areas in the system could result in a ``fail-as-is'' state
of the outputs. In the replacement system, dynamic coupling starting
in the main processor and going to the final output virtually
eliminates ``fail-as-is'' failure modes and replaces them with
``fail tripped'' modes.
The replacement system has the same loss of power failure mode
as the current system relative to the trip outputs and for loss of
AC [alternating current] power. For loss of DC [direct current]
power, the replacement system in most cases continues to operate
normally due to redundancy of the power supplies. Therefore, the
consequences are no different or improved compared to those
considered in the USAR.
Both the current system and the replacement system automatically
startup on application of power (or re-application). However, the
replacement system may take slightly longer to reach normal
operation due to initializing activities. However, no USAR
evaluations take credit for rapid start of the PRM. Therefore, the
slightly longer startup time from point of power application is
bounded by the USAR analysis. Upon application of power, once the
system is set up for the specific application, it automatically
returns to those settings upon application of power. All such setup
parameters are stored in non-volatile memory.
Human-machine interfaces (HMI) failures in the current system
could be related to misadjusted settings, incorrect reading of
meters, and failure to return the equipment to the normal operating
configuration. There are comparable failure modes for some of these
in the digital system where an
[[Page 68312]]
erroneous potentiometer adjustment in the current system is
equivalent to an erroneous digital entry in the replacement system.
Certain potential ``failure to reconfigure'' errors in the current
system have no counterpart in the replacement system because any
``reconfiguration'' is automatically returned to normal by the
system. Also, since parameters are available for review at any time,
even if an error such as a digital entry error occurs, it is more
likely that the error would be almost immediately detected by
recognition that the displayed value is not the correct one. Failure
analysis of the current system assumes certain rates of human error.
The rates for the replacement system will be lower, and hence are
bounded by the USAR analysis. The NUMAC-PRNM system has been
approved as an acceptable neutron monitoring replacement by the NRC.
Therefore, based on the above discussions, the proposed change
will not result in a significant increase in the consequences of any
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.
NMPC proposes to replace the existing RPS APRM system with the
NUMAC-PRNM system and make associated changes to the RPS and Control
Rod Block TS instrumentation sections. As discussed in NEDC-3241OP-
A, no new system level failure modes are created with the
replacement system. The NUMAC-PRNM modification and associated
changes to the TSs involve systems that are intended to detect the
symptoms of certain events or accidents and initiate mitigating
actions. The worst case failure of the systems involved would be a
failure to initiate mitigative actions (i.e., scram or rod block),
but no failure can cause an accident. This is unchanged from the
current system. The proposed changes do not modify the basic
functional requirements of the affected equipment, create any new
system interfaces or interactions nor create any new system failure
modes or sequence of events that could lead to an accident. The
replacement system is more tolerant of degraded power than the
current system. Software common cause failures can at most cause the
system to fail to perform its safety function. As with system level
failures, software failures could fail to initiate actions to
mitigate the consequences of an accident, but would not cause one.
Surveillance testing will continue to be performed to assure
reliability and maintain current performance levels.
The NUMAC-PRNM system is a digital system with software
(firmware) control. As such, it has ``central'' processing points
and software controlled digital processing where the current system
has analog and discrete component processing. The result is that the
specific failures of hardware and potentially common cause software
are different from the current system. Also, automatic self-test
results in some cases in a direct trip as a result of a hardware
failure where the current system may have remained ``as is.''
However, when these are evaluated at the system level, there are no
new effects. In general, the USAR assumes simplistic failure modes
(relays for example) but does not specifically evaluate effects
added by the NUMAC-PRNM such as self-test detection and automatic
trip or alarm. The effects of software common cause failures are
mitigated by hardware design and system architecture. The
replacement system is fully qualified to operate in its installed
location and will not affect other equipment. The NUMAC-PRNM system
has been approved as an acceptable neutron monitoring replacement by
the NRC. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any
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.
The proposed modification and associated TS changes will not
adversely affect the performance characteristics of the RPS and
Control Rod Block instrumentation nor will it affect the ability of
the subject instrumentation to perform its intended function. As
stated in NEDC-3241OP-A, the replacement system has improved channel
trip accuracy compared to the current system and meets or exceeds
system requirements assumed in setpoint analysis. Also, the channel
response time is within acceptable limits, the channel indicated
accuracy is improved over the current system, and the replacement
system does not cause a plant parameter for any analyzed event to
fall outside of acceptable limits. The surveillance testing and
frequencies proposed will assure reliability of the RPS and Control
Rod Block instrumentation. In addition, the subject equipment was
qualified, where appropriate, to assure its intended safety function
is performed. Therefore, the proposed changes do not involve
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: S. Singh Bajwa.
Pacific Gas and Electric Company
[Docket Nos. 50-275 and 50-323]
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo
County, California
Date of amendment requests: July 30, 1997.
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to add a limiting condition for operation
and surveillance requirements for a residual heat removal (RHR) pump
trip on low refueling water storage tank (RWST) level to TS 3/4.3.2,
``Engineered Safety Features Actuation System Instrumentation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This change assures the availability of the refueling water
storage tank (RWST) low-level trip of the residual heat removal
(RHR) pumps by establishing limits on the time that a channel can be
out of service to 72 hours and establishing surveillance criteria to
verify the operation of the logic. The RHR system is used to respond
to loss of coolant accidents (LOCAs) and other (e.g., secondary
side) accidents that could result in initiation of a safety
injection signal, and is not a precursor to any of these events as
evaluated in safety analyses. Under accident conditions the RWST
serves as the source of water for the emergency core cooling system
(ECCS) pumps and the containment spray pumps. The RWST and the RHR
pump trip are accident mitigation components and are not precursors
for any accident evaluated in the safety analyses.
The existing Technical Specification (TS) would allow one RWST
level indication channel to be inoperable indefinitely, and has an
allowed outage time (AOT) for two channels inoperable of up to seven
days. Additionally, the existing TS does not apply to the RWST low-
level RHR pump trip logic. The new TS provides controls that require
that all three RWST low-level trip channels be maintained operable
while the plant is in Modes 1 to 4, and provides for an AOT for one
channel inoperable for up to 72 hours, if the inoperable channel is
placed in the cut-out mode within 6 hours. By placing the inoperable
channel in the cut-out mode, the possibility of a channel failure
causing an RHR pump failure to start at the onset of an accident is
precluded even with a single active failure. This assures that the
consequences of an accident are not increased.
The change will have no affect on the probability of a physical
failure of an RHR pump because it only ensures the presence of a
pump trip signal when required. Therefore, there is no increase in
the probability of failure of an RHR train to function as designed.
This change will have no affect on the probability of any other ECCS
equipment failure as it only affects the presence of a trip signal
for the RHR pumps.
[[Page 68313]]
The new TS 3.3.2 item would provide controls that require that
all three RWST level channels be maintained operable while the plant
is in operating Modes 1 to 4 (power operation through hot shutdown).
By maintaining the three channels operable, the RHR pump actuation/
trip logic operability is assured so that the RHR and RWST can in
all cases perform their intended accident mitigation functions
following a design basis event as evaluated in the safety analyses.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The RHR system is used to respond to LOCAs and other (e.g.,
secondary side) accidents that could result in initiation of a
safety injection signal. Under accident conditions the RWST serves
as the initial source of water for injection by the RHR and other
ECCS pumps, and is the source of water for the containment spray
pumps. This change does not affect operation of the systems as it
relates to their response to accident conditions. It provides
additional assurance that the RHR pump trip logic will operate as
designed by establishing administrative controls on the time the
system is susceptible to a single failure. No new failure modes have
been introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The relevant margin of safety is based on the RHR pumps starting
and then automatically stopping at the correct RWST water level. The
new TS 3.3.2 item provides controls that require all three RWST
level channels be maintained operable while the plant is in Modes 1
to 4. By maintaining the three channels operable, the capability of
the RHR pump actuation/trip logic to survive a single active failure
is assured. Therefore, the trip logic operability is assured and the
margin is preserved. This change also provides additional assurances
that the remaining water in the RWST at the time of switchover is
consistent with that assumed in the Final Safety Analysis Report and
Safety Evaluation Reports.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
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 amendment request: June 25, 1997.
Description of amendment request: The amendments would modify the
Susquehanna Steam Electric Station, Units 1 and 2 Technical
Specifications to reflect an increase in the secondary containment
bypass leakage. Specifically, Section 3.6.1.2 is changed to replace the
leakage of 1.2 scf per hour for any one main steam line drain with
25.43 scfh for secondary containment bypass leakage from all sources;
Section 3.6.1.2 is changed to include the Main Steam Line Drain, high-
pressure coolant injection (HPCI) system drain, and reactor core
isolation cooling (RCIC) system drain leakages as part of the 300 scfh
leakage requirement; and Section 3/4.6.1.2 is changed to include a
discussion which related the secondary containment bypass leakage TS to
the radiological dose analyses.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Of the potential accidents described in FSAR [Final Safety
Analysis Report] Chapters 6 and 15, only a ``Decrease in Reactor
Coolant Inventory'' as described in FSAR Section 15.6.5 is affected
by the proposed action. The specific accident of concern is a design
basis LOCA [loss-of-coolant accident] concurrent with a LOOP [loss-
of-offsite power] which results in RPV [reactor pressure vessel]
depressurization and failure to recover RPV level above the FW
[feedwater] spargers. For this accident, the current licensing basis
offsite and control room dose analyses assume a secondary
containment bypass leakage rate of 9 scfh and primary containment
water (called ESF [engineered safety function]) leakage of 5 gpm.
The current licensing basis analyses do not attribute this leakage
to any specific pathway.
The proposed action does not increase the probability of a
previously analyzed accident in any way. The condition of concern is
the result of an accident and as such does not contribute to the
initiation of an accident as analyzed in the FSAR.
Of concern is whether or not the proposed action significantly
increases the consequences of an accident as previously evaluated.
Calculations of off-site dose assuming SCBL [secondary containment
bypass leakage] of 28 scfh, primary containment water leakage of 20
gpm, and crediting suppression pool scrubbing show decreases in
thyroid dose, but slight increases in whole body dose when compared
with dose calculations performed to support the removal of the MSIV-
LCS [main steam isolation valve-leakage control system]. This result
is expected because the effect of suppression pool scrubbing is
factored into the revised licensing basis analysis. Suppression pool
scrubbing is effective in reducing iodine release but has no assumed
effect on the removal of noble gases. Since the methodology/
assumptions for scrubbing are acceptable to the NRC [Nuclear
Regulatory Commission] per the guidance in SRP [Standard Review
Plan] Section 6.5.5 and the values for decontamination factors are
conservative, the judgment may be made that considerable margin is
preserved within the analysis.
Although the whole body dose with SCBL of 28 scfh and water
leakage of 20 gpm is increased from the previously approved MSIV-LCS
dose analysis, the increase is small (about 1 rem at the two hour
site boundary; less than 0.1 rem 30 day LPZ [low population zone]).
The total dose including the increase is still well below the
10CFR100 whole body regulatory limit of 25 rem to which SSES
[Susquehanna Steam Electric Station] was licensed. No change in
operating procedures is anticipated. Calculated post accident
control room thyroid dose decreases as a result of this change, and
the increase in control room whole body dose is less than 0.05 rem,
well below the 10CFR50, Appendix A, GDC [General Design Criterion]
19 dose limits outlined in NUREG-0800. Thus, no appreciable effect
on operator response will occur as a result of this change.
The addition of the HPCI and RCIC Steam Line Drains to the Tech
Spec for MSIV leakage is being performed as a result of the
modification which eliminated the MSIV Leakage Control System (MSIV
LCS). At the time this modification was performed, these lines were
not identified as potential SCBL pathways. However, because leakage
from the HPCI and RCIC drain lines are part of the same pathway to
the condenser which is now used by the main steam line drains (MSLD)
and included in the Technical Specifications, they must be combined
with the MSIV's and MSLD to be less than 300 scfh. This change only
affects the accounting of the various drain leakages in the valve
testing program. The justification for this change is the same
justification provided in the ITS [Improved Technical Specification]
submittal (PLA-4488, August 1, 1996) which adds the MSLD to this
Technical Specification. The test pressure change to allow testing
at Pa was previously proposed in PLA-4502, September 23, 1996. One
additional change to delete a footnote related to the removal of the
MSIV Leakage Control System is
[[Page 68314]]
included because this system has been removed from Susquehanna SES.
Since the increase in SCBL and primary containment water leakage
result in only a small increase in the doses previously evaluated by
the NRC and the other changes do not affect the dose analyses, the
proposed change does not result in a significant increase in the
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Because the FSAR analysis already assumes SCBL and ESF leakage
occur and the other changes do not affect the type of accident[s]
that are postulated to occur, the proposed change does not present
the possibility of an accident of a different type. Additionally,
the change in dose analysis methodology does not create an accident
or malfunction of a different type since it only involves the
analysis of the effects of such accidents or malfunctions.
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 change does not involve a significant reduction
in a margin of safety.
This question addresses changes in system parameters only. Dose
consequences are addressed in Section 1 above. The only Technical
Specification dealing with SCBL is T.S. 3.6.1.2 which requires the
leakage from any one Main Steam Line Drain (MSLD) Valve to be less
than or equal to 1.2 scfh when tested at Pa (45.0 psig). As noted
earlier, the current licensing basis accident dose analysis assumes
a total of 9 scfh for bypass leakage and 5 gpm for primary
containment water leakage but does not attribute them to any
particular source. The proposed action increases the assumed SCBL
from 9 to 28 scfh and water leakage from 5 gpm to 20 gpm. These
leakage rates are insignificant in terms of SGTS [standby gas
treatment system] flows or water loss from ECCS systems. These
leakage rates do not affect building temperatures or pressures so
that they become closer to acceptance limits. Likewise, no other
system parameter values become closer to limits as a result of these
changes in leakage. Consequently, the existing margin of safety
between the licensing basis analysis and system parameter acceptance
limits is not reduced. The changes to the HPCI, RCIC, and main steam
line drain leakage only affect the accounting for the various
leakages in the leakage testing program. The deletion of the
footnote is administrative because the MSIV Leakage Control System
has been removed from the Susquehanna SES. The change in test
pressure was previously evaluated in PLA-4502, September 23, 1996.
Thus, no decrease in margin of safety results.
The NRC staff has reviewed the licensee's analysis and notes that a
discussion of the administrative change to delete a footnote in Section
3.6.1.2 is in the third section of the no significant hazards
consideration. The staff finds that this administrative change also
does not involve a significant increase in the probability or
consequences of an accident previously evaluated and does not create
the possibility of a new or different kind of accident from any
accident previously evaluated. Based on this staff 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Pennsylvania Power and Light Company
[Docket No. 50-387]
Susquehanna Steam Electric Station, Unit 1, Luzerne County,
Pennsylvania
Date of amendment request: August 26, 1997.
Description of amendment request: The amendment would modify the
Susquehanna Steam Electric Station, Unit 1 Technical Specifications to
change the definitions in Section 1.0 to make them applicable to
ATRIUM-10 fuel (reflecting the new design), to include the Unit 1 Cycle
11 flow dependent minimum critical power ratio (MCPR) Safety Limits in
Sections 2.1.2 and 3.4.1.1.2, to change Section 5.3.1 to reflect the
ATRIUM-10 design, and to include Siemens Power Corporation methodology
topical reports and references to the methodology in Section 6.9.3.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The applicable sections of the FSAR [Final Safety Analysis
Report] are Chapters 5, 6.3, 9, and 15 of the FSAR. Chapter 5
discusses the results of the ASME [American Society of Mechanical
Engineers] overpressure analysis for the reactor pressure boundary.
Chapter 6.3 discusses the LOCA [loss-of-coolant accident]. Chapter 9
discusses fuel storage and handling. Chapter 15 describes the
transient and accident analyses, a majority of which have been
dispositioned to be non-limiting. A discussion of the impact of the
Technical Specification changes is provided below.
The change to Definitions 1.2 and 1.3 makes the definitions
applicable to ATRIUM TM-10. There are no effects on
safety functions from this change.
A cycle specific MCPR Safety Limit analysis was performed for
PP&L [Pennsylvania Power and Light Company] by SPC [Siemien Power
Corporation]. This analysis used NRC [Nuclear Regulatory Commission]
approved methods described in Technical Specification Reference 13
(ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2), as modified
by EMF-97-010(P), Rev. 1. The SAFETY LIMIT MCPR calculation
statistically combines uncertainties on feedwater flow, feedwater
temperature, core flow, core pressure, core power distribution, and
the uncertainty in the Critical Power Correlation. The SPC analysis
used cycle specific power distributions and calculated MCPR values
such that at least 99.9% of the fuel rods are expected to avoid
boiling transition during normal operation or anticipated
operational occurrences. The SAFETY LIMIT MCPRs are specified as a
function of core flow. The resulting two-loop and single-loop values
(Technical Specification Sections 2.1.2 and 3.4.1.1.2) are included
in the proposed change. Thus, the cladding integrity and its ability
to contain fission products are not adversely affected.
The MCPR methodology for ATRIUM TM-10 fuel (SPC
report EMF-97-010(P), Rev. 1), included in the revised Technical
Specifications via reference (Section 6.9.3.2) and previously
approved by the NRC for Unit 2 Cycle 9, describes conservative
methods for developing the MCPR Safety Limits and Operating Limits
for the U1C11 reload of ATRIUM TM-10 fuel in the
Susquehanna Steam Electric Station. This methodology conservatively
accounts for a flow dependence in the ATRIUM TM-10
critical power test data as well as an increased correlation
uncertainty for high local peaking factor rods. The results of using
this methodology are core flow dependent MCPR Safety Limits plus
conservative MCPR Operating Limits for Unit 1 Cycle 11. The
resulting MCPR Safety Limits and Operating Limits will continue to
assure that at least 99.9% of the fuel rods are expected to avoid
boiling transition during normal operation or anticipated
operational occurrences. Thus, the cladding integrity and its
ability to contain fission products are not adversely affected. The
proposed change in MCPR methodology does not physically affect the
plant or its systems.
Using the approach discussed in EMF-97-010(P), Rev. 1, analyses
of the Pump Seizure accident with the new MCPR methodology (SPC
report EMF-97-010(P), Rev. 1) will demonstrate that the NRC
acceptance criterion (i.e., small fraction of 10CFR100 dose limits)
is met.
The change to the Design Features (Section 5.3) increases the
maximum allowable lattice average enrichment. Analyses have
demonstrated that the ATRIUM TM-10 fuel will remain
subcritical (k-effective < 0.95)="" in="" both="" the="" spent="" fuel="" pool="" and="" the="" new="" fuel="" vault.="" thus,="" the="" change="" to="" maximum="" allowable="" lattice="" average="" enrichment="" has="" no="" impact="" on="" safety="" functions.="" the="" description="" [[page="" 68315]]="" of="" a="" fuel="" assembly="" (section="" 5.3)="" is="" also="" revised="" to="" reflect="" the="" atrium="">TM-10 central water channel, and reference to an
active fuel length of 150 inches was deleted. This change reflects
the physical characteristics of the ATRIUM TM-10 fuel and
has no impact on the probability or consequences of an event.
Included in the revised Technical Specifications via reference
(Section 6.9.3.2) are additional NRC approved methodology reports.
The NRC approved topical reports contain methodology which is used
to assure safe operation of Unit 1 with ATRIUM TM-10
fuel. These methodologies assure that the core meets appropriate
margins of safety for all expected plant operational conditions
ranging from refueling and cold shutdown of the reactor through
power operation. Thus, the results obtained from the analyses will
provide assurance that the reactor will perform its design safety
function during normal operation and design basis events.
The BASES changes for Section 2.1.1 (THERMAL POWER, Low Pressure
or Low Flow) reflect that the Safety Limit is valid for both 9x9-2
and ATRIUM TM-10. BASES for Section 2.1.2 were changed to
refer to Section 6.9.3.2 for applicable references.
Therefore, the proposed action does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The changes to the Unit 1 Technical Specifications (Definitions,
MCPR safety limits, Design Features, and inclusion of methodology
references) to allow use of ATRIUM TM-10 fuel do not
require any physical plant modifications, physically affect any
plant components, or entail significant changes in plant operation.
Thus, the proposed change does not create the possibility of a
previously unevaluated operator error or a new single failure. The
consequences of transients and accidents will remain within the
criteria approved by the NRC. 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 change does not involve a significant reduction
in a margin of safety.
The applicable Technical Specification Sections include 1.0,
2.0, 3/4.4, 5.3, and 6.9.3.2.
The changes to the Unit 1 Technical Specifications discussed in
Item 1 above do not require any physical plant modifications,
physically affect any plant components, or entail significant
changes in plant operation. Therefore, the proposed change will not
jeopardize or degrade the function or operation of any plant system
or component governed by Technical Specifications. The consequences
of transients and accidents will remain within the criteria approved
by the NRC. The proposed MCPR Safety Limits and the NRC approved
methods and revised MCPR methodology detailed in the references
added to Section 6.9.3.2 maintain an equivalent margin of safety as
defined in the BASES of the applicable Technical Specification
sections.
Therefore, the proposed change does 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Southern California Edison Company, et al.
[Docket Nos. 50-361 and 50-362]
San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego
County, California
Date of amendment requests: June 18, 1997.
Description of amendment requests: The licensee proposes to revise
Technical Specification (TS) 3.8.1, ``AC Sources--Operating'' and
applicable Bases. This change will more clearly reflect safety analysis
and testing conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change would revise Technical Specification (TS) TS
3.8.1, ``AC Sources--Operating,'' Surveillance Requirement (SRs)
3.8.1.1, 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13,
3.8.1.14, 3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and 3.8.1.20 and
applicable Bases to more clearly reflect surveillance test
conditions and system design requirements. Changes to the SRs
include more restrictive voltage and frequency acceptability limits.
The new requirements reflect the system design requirements in order
to ensure Class 1E system operability, meet the requirements of the
safety analysis, and to agree with the existing test surveillances.
In addition, the discussion regarding design basis reactive
power loading is eliminated since this cannot be readily controlled
during testing.
Operation of the facility would remain unchanged as a result of
the proposed change and no assumptions or results of any accident
analyses are affected. Therefore, the proposed change will not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change would revise Technical Specification (TS) TS
3.8.1, ``AC Sources--Operating,'' Surveillance Requirement (SRs)
3.8.1.1, 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13,
3.8.1.14, 3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and 3.8.1.20 and
applicable Bases to more clearly reflect surveillance test
conditions and system design requirements.
Operation of the facility would remain unchanged as a result of
the proposed change. Therefore, the proposed change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change would revise Technical Specification (TS) TS
3.8.1, ``AC Sources--Operating,'' Surveillance Requirement (SRs)
3.8.1.1, 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13,
3.8.1.14, 3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and 3.8.1.20 and
applicable Bases to more clearly reflect surveillance test
conditions and system design requirements. Changes to the SRs
include more restrictive voltage and frequency acceptability limits.
The new requirements reflect the system design requirements in order
to ensure Class 1E system operability, meet the requirements of the
safety analysis, and to agree with the existing test surveillances.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
Southern California Edison Company, et al.
[Docket Nos. 50-361 and 50-362]
San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego
County, California
Date of amendment requests: November 14, 1997 (supersedes February
1, 1994, amendment request).
[[Page 68316]]
Description of amendment requests: The licensee proposes to revise
the licensing basis as described in the Updated Final Safety Analysis
Report Section 3.5, ``Missile Protection,'' to allow the use of NUREG-
0800, ``Standard Review Plan'' methodology in evaluating tornado-
generated missiles. In particular, a probability based criteria is
proposed to evaluate missile barrier requirements consistent with
Section 3.5.1.4 of NUREG-0800.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
NUREG-0800, Standard Review Plan (SRP) Section 3.5.1.4, Revision
0 and Section 3.5.1.5 Revision 1 provide a conservatively acceptable
probability threshold for safety due to damage caused by postulated
missile strikes. Section 3.5.1.4, Revision 0 uses 10-7
per year for a tornado-generated missile strike, and Section 3.5.1.5
Revision 1 uses 10-7 per year for exceeding 10 CFR Part
100 limits.
The proposed criteria of probability of damage to critical
exposed equipment (as defined in San Onofre Updated Final Safety
Analysis Report proposed Table 3.5-13) of 10-7 per year
per unit is consistent with this guidance.
The probability of damage to exposed critical components due to
a postulated missile strike of 10-7 is so small as to be
negligible. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This amendment request establishes a conservative criteria for
tornado-generated missiles consistent with the SRP guidance and will
not create a new or different kind of accident from any accident
that has been previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This proposed change is consistent with the methodology and
acceptance criteria of the SRP, and the SRP criteria ensures that
there will be no undue risk to the health and safety of the public.
Therefore, there will be no significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Attorney for licensee: T.E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia
Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia
[Docket Nos. 50-424 and 50-425]
Date of amendments request: August 8, 1997, as supplemented October
10, 1997. This application and supplement supersedes the October 4,
1996, application, noticed in the Federal Register on November 19, 1996
(61 FR 58903), in its entirety.
Description of amendments request: The proposed amendments would
change the Technical Specifications to credit soluble boron in the
spent fuel pool for maintenance of subcriticality and increase the
allowable fuel enrichment to 5.0 percent U-235 as follows:
1. Revisions to the Table of Contents
The Table of Contents would be revised to include two additional
Technical Specifications 3.7.17, ``Fuel Storage Pool Boron
Concentration,'' and 3.7.18, ``Fuel Assembly Storage in the Fuel
Storage Pool'' and add Figures 3.7.18-1, 3.7.18-2, and 4.3.1-1
through 4.3.1-9 describing burnup credit, checkerboard
configurations and interface requirements. These changes would be
added to support crediting soluble boron in the fuel storage pool
criticality analyses.
2. Addition of Technical Specifications 3.7.17 and 3.7.18
Technical Specifications 3.7.17, ``Fuel Storage Pool Boron
Concentration,'' and 3.7.18, ``Fuel Assembly Storage in the Fuel
Storage Pool,'' would be added to credit soluble boron in the fuel
storage pool criticality analyses, and specify acceptable
enrichment-burnup combinations for storage of fuel in the fuel
storage pool.
3. Revision to Technical Specification 4.3.1.1
Design Features Section 4.3.1.1 would be revised to reflect the
increased maximum enrichment assumed in the fuel storage pool
criticality analyses, add a requirement to maintain Keff
less than 1.0 when fully flooded with unborated water, change the
0.95 Keff requirement from ``if fully flooded with
unborated water'' to ``when fully flooded with water borated to 450
ppm (Unit 1) or 500 ppm (Unit 2),'' and to add a reference to
Specification 3.7.18 for allowable enrichment-burnup combinations.
Requirements for fuel that do not meet the requirements of
Specification 3.7.18, would also be added to Section 4.3.1.1,
including Figures 4.3.1-1 through 4.3.1-9 depicting acceptable
enrichment-burnup requirements and checkerboard configurations.
4. Revisions to the Table of Contents (Bases)
The Table of Contents would be revised to include two additional
Technical Specification Bases Sections B 3.7.17 ``Fuel Storage Pool
Boron Concentration'' and B 3.7.18 ``Fuel Assembly Storage in the
Fuel Storage Pool.''
5. Addition of Bases for Technical Specifications 3.7.17 and 3.7.18
Two additional Technical Specification Bases Sections B 3.7.17,
``Fuel Storage Pool Boron Concentration'' and B 3.7.18, ``Fuel
Assembly Storage in the Fuel Storage Pool'' would be added to credit
soluble boron in the fuel storage pool criticality analyses.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The radiological consequences of 5.0 weight percent U-235 fuel
on accidents previously evaluated in the Vogtle FSAR [Final Safety
Analysis Report] are not significant. Increasing the enrichment up
to and including 5.0 weight percent U-235 has minor effects on the
radiological source terms and subsequently the potential releases
both normal and accidental are not significantly affected.
Evaluations performed (WCAP-12610-P-A, Reference 5 [of the
licensee's application]) considered the source term, gap fraction,
and the accident doses for a maximum fuel enrichment of 5.0 weight
percent U-235. It was concluded that operating with and storing fuel
with 5.0 weight percent U-235 enrichment may result in minor changes
in the normal annual releases of long half-life fission products
that are not significant. Also, the radiological consequences of
accidents are minimally affected due to the very small changes in
the core inventory and the fact that the currently assumed gap
fractions remain bounding.
The use of the slightly higher enrichment for VEGP [Vogtle
Electric Generating Plant] fuel will not result in burnups in excess
of those currently allowed for VEGP. The cycle design methods and
limits will remain the same as are currently licensed. Therefore,
the use of fuel with the higher enrichment will not result in
conditions outside those currently allowed for VEGP.
There is no increase in the probability of a fuel assembly drop
accident in the fuel storage pool when considering the presence of
soluble boron in the pool water for criticality control. The
handling of the fuel assemblies in the fuel storage pool has always
been performed in borated water.
Fuel assembly placement will be controlled pursuant to approved
fuel
[[Page 68317]]
handling procedures and will be in accordance with the spent fuel
rack storage configuration limitations in the Technical
Specifications. The consequences of a misplaced assembly have been
included in the analysis supporting this revision to the Technical
Specifications.
There is no increase in the consequences of the accidental
misloading of a fuel assembly into the fuel storage pool racks
because criticality analyses demonstrate that the pool will remain
subcritical following an accidental misloading of an assembly. There
are no credible dilution events that reduce the subcriticality
margin below the 5% margin recommended in NRC guidance (references
1, 2, and 3 [of the licensee's application]). Even if the fuel
storage pool were diluted to a boron concentration of 0 ppm the No
Soluble Boron 95/95 analysis demonstrates that the pool will remain
subcritical. The proposed Technical Specifications limitations will
ensure that an adequate fuel storage pool boron concentration will
be maintained.
There is no increase in the probability of the loss of normal
cooling to the fuel storage pool water due to the presence of
soluble boron in the pool water for subcriticality control, because
a concentration of soluble boron similar to the proposed limit has
always been maintained in the fuel storage pool water.
The loss of normal cooling to the fuel storage pool will cause
an increase in the temperature of the fuel storage pool water. This
will cause a decrease in water density which would normally result
in an addition of negative reactivity. However, since Boraflex is
not considered to be present, and the fuel storage pool water has a
high concentration of boron, a density decrease causes a positive
reactivity addition. The amount of soluble boron required to offset
this postulated accident was evaluated for the allowed storage
configurations. The amount of soluble boron necessary to mitigate
these accidents and ensure that the Keff will be
maintained less than or equal to 0.95 has been included in the fuel
storage pool boron concentration. Because adequate soluble boron
will be maintained in the pool water, the consequences of a loss of
normal cooling to the fuel storage pool will not be increased.
Therefore, based on the conclusions of the above analysis, the
proposed changes will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The potential for criticality accidents in the fuel storage pool
are not new or different types of concerns. The potential
criticality accidents have been reanalyzed in the Criticality
Analysis report (Enclosure 5 [of the licensee's application]) to
demonstrate that the pool remains subcritical.
Soluble boron has been maintained in the fuel storage pool water
since its initial operation. The possibility of a fuel storage pool
dilution is not affected by the proposed change to the Technical
Specifications. Therefore, the implementation of Technical
Specification controls for the soluble boron will not create the
possibility of a new or different kind of accidental pool dilution.
With credit for soluble boron now a major factor in controlling
subcriticality, an evaluation of fuel storage pool dilution events
was completed. The results of the evaluation concluded that no
credible events would result in a reduction of the criticality
margin below the 5% margin recommended by the NRC. In addition, the
No Soluble Boron 95/95 criticality analysis assures that dilution to
0 ppm will not result in criticality.
Proposed Technical Specifications 3.7.17, 3.7.18 and 4.3.1.1
which ensure the maintenance of the fuel storage pool boron
concentration and storage configuration, do not represent new
concepts. The actual boron concentration in the fuel storage pool
has been maintained at a higher value than the proposed limits for
the Unit 1 and 2 fuel storage pools for refueling purposes. The
criticality analysis (Enclosure 5 [of the licensee's application])
determined that a boron concentration of 450 ppm (Unit 1) and 500
ppm (Unit 2) results in a Keff [less than or equal to]
0.95.
There is no significant change in plant configuration, equipment
design, or usage of plant equipment. The safety analysis for
dilution accidents has been expanded; however, the criticality
analyses assure that the pool will remain subcritical with no credit
for soluble boron. Therefore, the proposed changes will not create
the possibility of a new or different kind of accident.
3. The proposed change does not result in a significant
reduction in the margin of safety.
Proposed Technical Specifications 3.7.17, 3.7.18, and 4.3.1.1
and the associated fuel storage pool boron concentration and storage
requirements will provide adequate margin to assure that the fuel
storage array will always remain subcritical by the 5% margin
recommended by the NRC. Those limits are based on the criticality
analysis (Enclosure 5 [of the licensee's application]) performed in
accordance with the Westinghouse fuel storage rack criticality
analysis methodology described in Reference 4 [of the licensee's
application].
While the criticality analysis utilized credit for soluble
boron, the storage configurations have been defined using
Keff calculations to ensure that the spent fuel rack
Keff will be less than 1.0 with no soluble boron.
Soluble boron credit is used to offset off-normal conditions
(such as a misplaced assembly) and to provide subcritical margin
such that the fuel storage pool Keff is maintained less
than or equal to 0.95.
The combination of the No Soluble Boron 95/95 Keff
calculation which shows that the Keff will remain less
than 1.0 when flooded with unborated water and the unavailability of
the large volumes of water which are necessary to dilute the fuel
storage pool to a Keff of > 0.95, provide a level of
safety comparable to the conservative criticality analysis
methodology required by References 1, 2, and 3 [of the licensee's
application].
Therefore, the proposed changes in this license amendment will
not result in a significant reduction in the plant's 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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Project Director: Herbert N. Berkow.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of
Georgia, City of Dalton, Georgia
Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia
[Docket Nos. 50-424 and 50-425]
Date of amendment request: September 4, 1997, as supplemented
November 20, 1997.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to change the capacity of the
Vogtle Unit 1 spent fuel storage pool from 288 to 1476 assemblies, and
would revise the design features description to reflect the criticality
analyses and storage cell spacing. Specifically, the changes would be
as follows:
1. Figure 3.7.18-1 would be replaced with a revised figure based
on the criticality analyses for the Unit 1 racks containing boral.
2. The criticality information for Unit 2 would be placed
unchanged into Section 4.3.1.2, and Section 4.3.1.1. would be
revised to address Unit 1.
3. Design Features Section 4.3.1.1.c would be revised to
indicate 600 ppm as the required amount of soluble born to maintain
Keff less than or equal to 0.95.
4. Design Features Section 4.3.1.1.d would be revised to include
the reference Keff that is equivalent to the combination
of burnup and initial enrichment defined by Figure 3.7.18-1.
5. Design Features Section 4.3.1.1.e would be revised to
indicate that fuel assemblies with up to 5 weight percent U-235 may
be stored in 3-out-of-4 checkerboard storage configurations; delete
Figure 4.3.1-1; eliminate the reference to 2-out-of-4 storage for
the Unit 1 pool and include the reference K acceptable for all cell
storage in the Unit 1 fuel storage racks.
6. Design Features Section 4.3.1.1.f would be revised to include
the pitch of the Unit 1 fuel storage racks.
[[Page 68318]]
7. Design Features Section 4.3.3 would be revised to indicate
the Unit 1 fuel storage pool capacity of 1476 fuel assemblies.
8. The titles on Figures 4.3.1-4, 4.3.1-6, and 4.3.1-7 would be
revised to reflect the elimination of 2-out-of-4 storage
configuration requirements for the Unit 1 fuel storage pool.
Changes to the TS Bases are also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The analyses methodologies are the same as previously
approved for use by the NRC. The results of the analyses resulted in
fuel pool boron concentrations, and fuel assembly storage
limitations that are similar to those already submitted to the NRC.
The increased number of fuel assemblies will remain less than the
number previously accepted by the NRC for storage in VEGP [Vogtle
Electric Generating Plant] Unit 2, which has a similarly designed
and constructed facility, with the exception of the number of fuel
storage locations.
Therefore, based on the conclusions of the above analysis, the
proposed changes will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The effects of accidents that could affect the fuel were
analyzed for the fuel storage racks, however the types of accidents
have not changed. The fuel to be stored in the Unit 1 pool is
expected to meet the all cell storage requirements. The racks will
be placed in the Unit 1 pool without lifting any loads over spent
fuel. After installation of the new racks, the Unit 1 pool will have
1476 storage locations which is well within the 2098 locations that
the pool and structure is capable of storing, based on its
similarity to the Unit 2 pool.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident.
3. The changes to the technical specifications are necessary to
incorporate the parameters resulting from the criticality analyses.
The criticality analyses were performed using methods and criteria
previously accepted by the NRC. The requirements are similar to the
previously submitted requirements. The margins of safety provided by
the previous technical specifications are not significantly affected
because the new racks are based on the same acceptance values. The
larger number of fuel assemblies to be stored in the Unit 1 pool
remains well within the capability of the pool.
Therefore, the proposed changes in this license amendment will
not result in a significant reduction in the plant's 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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Project Director: Herbert N. Berkow.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia
Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia
[Docket Nos. 50-424 and 50-425]
Date of amendment request: November 20, 1997.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) to provide for the following
with regard to the Reactor Trip System (RTS) and Engineered Safety
Feature Actuation System (ESFAS) instrumentation trip setpoints:
1. The inequalities as they are applied to the Trip Setpoint
column of Tables 3.3.1-1 and 3.3.2-1 would be deleted, and the
column heading would be changed from ``Trip Setpoint'' to ``Nominal
Trip Setpoint.''
2. A footnote would be added to the new ``Nominal Trip
Setpoint'' column of Tables 3.3.1-1 and 3.3.2-1 that would allow the
trip setpoints to be set more conservative than the nominal value as
necessary to respond to plant conditions.
3. The Allowable Value for Table 3.3.1-1, Function 14.b, Turbine
Trip--Turbine Stop Valve Closure, would be revised from ``[greater
than or equal to] 96.7% open'' to ``[greater than or equal to] 90%
open.''
4. Footnotes l and m of Table 3.3.1-1 would be revised to refer
to the ``Nominal Trip Setpoint'' and delete the inequalities applied
to the trip setpoints.
5. A superscript ``(a)'' would be deleted from the heading of
the ``Trip Setpoint'' column on page 6 of 8 of Table 3.3.1-1.
6. Notes 1 and 2 to Table 3.3.1-1, Overtemperature T
and Overpower T, respectively, would be revised to refer to
the ``Nominal Trip Setpoint.'' In addition, these notes will be
revised to delete the inequalities from the values for the constants
K1 through K6 (except for K5
[greater than or equal to] 0 for decreasing temperature and
K6 = 0 for T [less than or equal to] T''), and for T',
T'', and P'.
7. The inequality applied to the ESFAS Allowable Value for Steam
Line Pressure--Low (Table 3.3.2-1, Function 1.e) would be changed
from ``[less than or equal to]'' to ``[greater than or equal to].''
Associated changes to the TS Bases are also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed changes affect only the presentation of the
trip set points for the RTS and ESFAS in the VEGP [Vogtle Electric
Generating Plant] Units 1 and 2 TS. The calibration of the channels
whose setpoints are specified in the TS will continue to be
performed in a manner consistent with the setpoint methodology
described in WCAP-11269 Rev. 1. There will be no adverse effect on
the ability of those channels to perform their safety functions as
assumed in the safety analyses. Since there will be no adverse
affect on the trip setpoints or the instrumentation associated with
those trip setpoints, there will be no increase in the probability
of any accident previously evaluated. Similarly, since the ability
of the instrumentation to perform its safety function is not
adversely affected, there will [be] no increase in the consequences
of any accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change affects only the presentation of trip
setpoint requirements in the TS. Plant operation will not be
changed, and the response of safety related equipment as assumed in
the accident analyses will not be adversely affected. Therefore, the
proposed change does not involve a new or different kind of accident
than any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety [?]
No. As described above, the RTS and ESFAS instrumentation will
remain capable of performing its safety function as assumed in the
accident analyses. The treatment of trip setpoints as nominal values
is consistent with the methodology used to establish those
setpoints. As such, margin is not affected by the proposed change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Project Director: Herbert N. Berkow.
[[Page 68319]]
Vermont Yankee Nuclear Power Corporation
[Docket No. 50-271]
Vermont Yankee Nuclear Power Station, Windham County, Vermont
Date of amendment request: October 10, 1997, as supplemented
October 31, 1997.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to reflect the installation of
a generator no-load disconnect to facilitate use of the main step-up
transformer backfeed as the delayed access offsite power source. Also,
the amendment would revise existing limiting conditions for operation
and required action statements for operation with inoperable ac power
sources to be consistent with current guidance.
Specifically, the changes proposed are: (1) TS Limiting Conditions
for Operation Section--Normal Operation, 3.10.A.4 (2) TS Limiting
Conditions for Operation Section--Operation with Inoperable Components,
3.10.B.3, (3) TS Surveillance Requirements--Normal Operation, 4.10.A.4,
(4) TS Surveillance Requirements--Operation with Inoperable Components,
Section 4.10.B.3, (5) Bases Section 3.10.A, (6) Bases Section 3.10.B,
(7) Bases Section 4.10.A, and (8) Bases Section 4.10.B
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
(1) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment removes credit for the Vernon Tie,
Vermont Yankee's station blackout source of power, from the
Technical Specifications and reflects the installation of the
generator no load disconnect as part of the backfeed. Neither the
backfeed through the main transformers nor the Vernon Tie are
accident initiators; therefore, the change does not involve a
significant increase in the probability of an accident previously
evaluated. The change does not affect the capability, availability,
maintenance or operation of the Vernon Tie. Installation of the
generator no load disconnect switch is being implemented by a design
change in order to enhance plant safety by reducing time necessary
to establish the backfeed through the main transformer. A separate
10 CFR 50.59 evaluation is being prepared to document that the
modification does not create an unreviewed safety question.
The proposed amendment also clarifies the allowable out of
service times, and required actions; and updates surveillance
requirements for the immediate and delayed access offsite power
sources. These changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Modification of a technical specification out of service time and
required action cannot affect the probability or consequences of an
accident. Enhancing surveillance requirements to provide assurance
that the backfeed can be achieved when required and to provide
assurance that remaining power sources are available when an offsite
source is unavailable improves plant safety and does not increase
the probability or consequences of an accident.
Therefore, the change 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 proposed amendment removes the Vernon Tie, Vermont Yankee's
station blackout source of power, as a delayed access source from
the Technical Specifications and reflects the improvements to the
main transformer backfeed delayed access source because of
installation of the generator no load disconnect. Neither the
removal of the Vernon Tie from Technical Specifications nor the
improvements to the delayed access power source (backfeed) can
create the possibility of a new or different kind of accident from
any previously evaluated.
The proposed amendment also clarifies the allowed outage times,
and action statements; and updates surveillance requirements for the
immediate and delayed access offsite power sources. A clarification
of a technical specification out of service time and required action
cannot create a new or different kind of accident from any accident
previously evaluated. Enhancing surveillance requirements to provide
assurance that the backfeed can be achieved when required and to
provide assurance that remaining power sources are available when an
offsite source is unavailable improves plant safety and cannot
create a new or different kind of accident from any accident
previously evaluated.
Therefore, this change would not create the possibility of a
different type of accident than previously evaluated.
(3) The proposed amendment will not involve a significant
reduction in a margin of safety.
The proposed amendment removes the Vernon Tie, Vermont Yankee's
station blackout source of power, as a delayed source of offsite
power from the Technical Specifications and reflects the
improvements to the main transformer backfeed delayed access source
because of installation of the generator no load disconnect. No
existing safety margins are adversely affected. The backfeed is
modified so that it may be established in sufficient time to
``assure that specified acceptable fuel design limits and design
conditions of the reactor coolant pressure boundary are not
exceeded''. Vernon Tie will not be affected by the modification and
remain available as a station blackout source; thus this change does
not involve a significant reduction in the margin of safety.
The proposed amendment also clarifies the allowed out of service
times, and required actions; and updates surveillance requirements
for the immediate and delayed access offsite power sources. A
clarification of a technical specification out of service time and
required action does not involve a significant reduction in the
margin of safety in the Technical Specifications. Enhancing
surveillance requirements to provide assurance that the backfeed can
be achieved when required and to provide assurance that remaining
power sources are available when an offsite source is unavailable
improves plant safety and 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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Project Director: Ronald Eaton, Acting Director.
Vermont Yankee Nuclear Power Corporation
[Docket No. 50-271]
Vermont Yankee Nuclear Power Station, Windham County, Vermont
Date of amendment request: November 20, 1997.
Description of amendment request: The proposed amendment would
revise the existing requirements for the Auxiliary Electrical Power
Systems as identified in Technical Specifications (TSs) 3/4.10.A and TS
3.10.A.2.b. The specific changes are:
(1) The requirements in TS 3.10.A.2.b. are revised to omit the
allowance for Spare Charger AB to substitute for either Charger A or B.
(2) The Bases in TS 3.10.A. are revised to omit the statements that
justify Spare Charger AB to substitute for either Charger A or Charger
B.
The proposed changes provide more limiting requirements for
operation with the standby battery charger in service.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
[[Page 68320]]
Neither batteries, nor their chargers, are considered to be an
initiator of any previously analyzed accident. Therefore, this
change will not significantly increase the probability of any
previously analyzed accident.
At least one Battery System is required to be available to
mitigate the consequences of a Design Basis Accident. This change
removes an allowance which places the unit in a more vulnerable
condition through the unrestricted use of the spare battery charger.
Since this change limits such a condition, it maintains the
assumptions of the safety analysis, and therefore, will not
significantly increase the consequences of any previously analyzed
accident.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) nor is operation of the currently installed equipment
changed. The change will, however, limit a currently allowed
configuration with the spare charger and is more conservative. Thus,
this change will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
The proposed change continues to provide the previous margin of
safety regarding the capability to withstand a single failure. At
least one Battery System will continue to be available to provide
the required safety function. The change will limit a currently
allowed configuration with the spare charger and is thus more
conservative. Therefore, this change will not significantly reduce 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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Project Director: Ronald Eaton, Acting Director.
Vermont Electric and Power Company
[Docket Nos. 50-280 and 50-281]
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 5, 1997.
Description of amendment request: The proposed change to Technical
Specifications 5.3 and 5.4 would reflect an increase in the maximum
permitted fuel enrichment to 4.3 weight percent U235 from
the current 4.1 weight percent U235. Fuel burnup limits and
reactor operating power level would remain unchanged.
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:
Virginia Electric and Power Company has reviewed the Technical
Specifications changes for Surry Units 1 and 2 against the criteria
of 10 CFR 50.92. It has been concluded that use of fuel with the
slightly higher initial enrichment does not involve a significant
hazards consideration as defined in 10 CFR 50.92. An increase in the
maximum initial fuel enrichment from 4.1 to 4.3 weight percent
U235 will not:
1. Involve a significant increase in the probability of
occurrence or the consequences of an accident previously evaluated.
The only accidents for which the probability of occurrence is
potentially affected by the fuel enrichment involve criticality
events during handling and storage. Analyses have demonstrated that
the K-effective will be low enough to ensure subcriticality during
both normal operation and under postulated accident conditions
during the handling and storage of both new and spent fuel.
Therefore, the probability of occurrence of criticality during fuel
handling or storage is not increased. Safety analyses of record are
based on inputs which bound the proposed increase in fuel
enrichment. Since no changes to the fuel burnup limits are
requested, the radiological consequences of previously evaluated
accident scenarios will not be increased. Therefore, neither the
probability of occurrence nor the consequences of any accident
previously evaluated is significantly increased.
2. Create the possibility for a new or different type of
accident from any accident previously evaluated. Fuel with the
higher initial enrichment will meet all applicable design criteria
and will operate within existing Technical Specifications limits.
Adherence to these standards and criteria precludes new challenges
to components and systems that could introduce a new type of
accident. All design and performance criteria will continue to be
met. In addition, the use of a slightly higher initial fuel
enrichment does not involve any alteration to plant equipment or
procedures which would introduce any new or unique operational modes
or accident precursors. Therefore, the possibility for a new or
different kind of accident from any accident previously evaluated is
not created.
3. Involve a significant reduction in the margin of safety.
Surry Units 1 and 2 will continue to operate in compliance with the
Technical Specifications, ensuring that the plants continue to
provide acceptable levels of protection for the health and safety of
the public. The Technical Specifications are based upon
assumption[s] made in the safety and accident analyses, including
those relating to the fuel enrichment and the design of the fuel
storage areas. Analyses have demonstrated that subcriticality will
be ensured during fuel storage and handling accident scenarios for
both new and spent fuel. Additionally, safety analyses of record for
core operation will remain applicable for Surry Unit 1 and 2 cores
which use fuel with the slightly higher U235 enrichment.
Therefore, the regulated margin of safety as defined in the Bases to
the Surry Technical Specifications is not reduced.
Based on the preceding information, it has been determined that
the use of fuel with an initial enrichment of up to 4.3 weight
percent U235 satisfies the no significant hazards
consideration criteria of 10 CFR 50.92.
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: Swern Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: James E. Lyons.
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
[[Page 68321]]
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.
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 30, 1997, as supplemented
on September 25, 1997.
Brief description of amendments: The amendments grant partial
credit for boron in the spent fuel pools to maintain the
subcriticality.
Date of issuance: December 4, 1997.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 94, 94, 86 and 86.
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 22, 1997 (62 FR
54868).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 4, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Duquesne Light Company, et al.
[Docket Nos. 50-334 and 50-412]
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: March 14, 1997, as
supplemented. July 29, 1997, and August 13, 1997. The July 29, 1997,
and August 13, 1997, letters provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination or expand the amendment request beyond the scope of the
May 7, 1997, Federal Register notice.
Brief description of amendments: These amendments relocate certain
administrative control Technical Specifications (TSs) from the Beaver
Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2), TSs to the
licensee's operational quality assurance program description, which is
presented in Section 17.2 of the BVPS-2 Updated Final Safety Analysis
Report (UFSAR). Section 17.2 of the BVPS-2 UFSAR contains the quality
assurance program description for both BVPS-1 and BVPS-2. The following
TSs are being relocated to the quality assurance program description.
BVPS-2 TS 6.2.3 (Independent Safety Evaluation Group)
BVPS-1 and BVPS-2 TS 6.5.1 (Onsite Safety Committee)
BVPS-1 and BVPS-2 TS 6.5.2 (Offsite Review Committee)
BVPS-1 and BVPS-2 TS 6.8.2 (Procedures, Review and)
BVPS-1 and BVPS-2 TS 6.8.3 (Temporary Procedure Changes, Review and
Approval)
BVPS-1 and BVPS-2 TS 6.10.1 (Records Retention, At least 5 Years)
BVPS-1 and BVPS-2 TS 6.10.2 (Records Retention, Duration of Operating
License)
Date of issuance: December 10, 1997.
Effective date: Both units, as of date of issuance, to be
implemented within 60 days.
Amendment Nos.: 209 and 87.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications, and Appendix C of the License.
Date of initial notice in Federal Register: May 7, 1997 (62 FR
24986).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Entergy Operations, Inc.
[Docket No. 50-382]
Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 17, 1996, as supplemented by
letters dated June 3, and July 7, 1997. Also, application dated April
11, 1997.
Brief description of amendment: The amendment changes the Appendix
A Technical Specification (TS) 3.7.1.3 by increasing the minimum
required contained water volume in Condensate Storage Pool from 82
percent to 91 percent indicated level. In addition, this amendment
expands the applicability of TS 3.7.1.3 to include Mode 4 operational
requirements. The amendment also deletes Action (b) in TS 3.7.1.3 and
its associated surveillance requirement in Waterford 3 TSs.
Date of issuance: December 18, 1997.
Effective date: December 18, 1997, to be implemented within 60
days.
Amendment No.: 137.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14461), July 30, 1997 (62 FR 40849) and April 22, 1997 (62 FR 19624).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 18, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Florida Power Corporation, et al.
[Docket No. 50-302]
Crystal River Unit No. 3 Nuclear Generating Plant, Citrus County,
Florida
Date of application for amendment: August 26, 1997.
Brief description of amendment: The amendment involves a revision
to the design basis of the Emergency Diesel Generator (EDG) Air
Handling System at Crystal River 3 resulting from the EDG upgrade
modification which increased the 200-hour and 2000-hour service ratings
for each EDG.
Date of issuance: December 12, 1997.
Effective date: December 12, 1997.
Amendment No.: 160.
Facility Operating License No. DPR-31: Amendment revises the Final
Safety Analysis Report.
Date of initial notice in Federal Register: September 24, 1997 (62
FR 50004).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 12, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal River, Florida 34428
[[Page 68322]]
Indiana Michigan Power Company
[Docket Nos. 50-315 and 50-316]
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: September 19, 1997
(AEP:NRC:1278).
Brief description of amendments: The amendments modify Technical
Specification 4.5.2.d.1 to delete the interlock that would close the
Residual Heat Removal (RHR) suction valves if the Reactor Coolant
System (RCS) pressure were to increase to 600 psig while retaining the
interlock that would prevent the suction valves from opening while the
RCS pressure is above the RHR system design pressure. This change
maintains the open interlock function and allows continued deactivation
of the isolation valves to assure RHR availability and provide low
temperature overpressure protection.
Date of issuance: December 10, 1997.
Effective date: December 10, 1997, with full implementation within
45 days.
Amendment Nos.: 219 and 203.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54861).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
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 7, 1996, as
supplemented by letter dated May 9, 1997.
Brief description of amendments: These amendments modify
Susquehanna Steam Electric Station, Units 1 and 2, Technical
Specifications Table 3.3.2-2 by revising the trip setpoints and
allowable values for secondary containment isolation radiation
monitors.
Date of issuance: December 8, 1997.
Effective date: December 8, 1997.
Amendment Nos.: 170 and 143.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66716).
The May 9, 1997, letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 8, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
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: April 4, 1997, as supplemented
April 14, June 6, and September 2, 1997.
Brief description of amendments: These amendments clarify the scope
of the surveillance requirements for response time testing of
instrumentation in the reactor protection system, isolation actuation
system, and emergency core cooling system in the Technical
Specifications for each unit (Sections 4.3.1.3, 4.3.2.3, and 4.3.3.3).
Date of issuance: December 8, 1997.
Effective date: December 8, 1997.
Amendment Nos.: 171 and 144.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 17, 1997 (62 FR
17885).
The April 14, June 6, and September 2, 1997, letters provided
clarifying information that did not change the original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 8, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Power Authority of the State of New York
[Docket No. 50-333]
James A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: December 14, 1995, as
supplemented September 26, 1997.
Brief description of amendment: The amendment proposes to change
the James A. FitzPatrick Technical Specifications to incorporate the
inservice testing requirements of Section XI of the American Society of
Mechanical Engineers Boiler and Pressure Vessel Code.
Date of issuance: December 2, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 241.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1635).
The September 26, 1997, 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 December 2, 1997.
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.
Rochester Gas and Electric Corporation
[Docket No. 50-244]
R. E. Ginna Nuclear Power Plant, Wayne County, New York
Date of application for amendment: September 29, 1997, as
supplemented October 8, 1997.
Brief description of amendment: The amendment revises the Ginna
Station Technical Specifications (TS) to allow referencing of revision
of the Ginna Station pressure and temperature limits report for the
reactor coolant system pressure and temperature limits and low
temperature overpressure protection limits. The amendment also corrects
a typographical error in the TSs.
Date of issuance: December 9, 1997.
Effective date: December 9, 1997.
Amendment No.: 70.
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
[[Page 68323]]
Date of initial notice in Federal Register: November 5, 1997 (62 FR
59921).
The September 29 and October 8, 1997, superseded in their entirety
the applications dated December 13, 1996, April 24, 1997, and June 3,
1997.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 9, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Southern California Edison Company, et al.
[Docket Nos. 50-361 and 50-362]
San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego
County, California
Date of application for amendments: December 22, 1995, as
supplemented by letter dated November 25, 1997.
Brief description of amendments: These amendments revise License
Conditions 2.E and 2.G for the San Onofre Nuclear Generating Station
(SONGS), Units 2 and 3. The amendments delete the physical protection
program reporting requirement from License Condition 2.G, and clarify
in License Condition 2.E that not all documents composing the physical
protection program plans necessarily contain safeguards information.
Date of issuance: December 16, 1997.
Effective date: December 16, 1997.
Amendment Nos.: Unit 2--138; Unit 3--130.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: November 5, 1997 (62 FR
59921). The November 25, 1997, letter provided additional clarifying
information and did not change the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated December 16, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Virginia Electric and Power Company, et al.
[Docket Nos. 50-338 and 50-339]
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: May 14, 1997, as supplemented
October 15, 1997. The October 15, 1997, submittal provided clarifying
information only, and did not change the proposed no significant
hazards consideration determination.
Brief description of amendments: The proposed action consists of
changes to the Technical Specifications (TS) revising Surveillance
Requirement 4.7.1.7.2.a for both units to clarify the testing and
inspection methodology of the turbine governor control valves. The
proposed changes also provide clarification in the TS Bases Section 3/4
7.1.7 for the Turbine Valve Freedom Testing of the turbine governor
control valves.
Date of issuance: December 4, 1997.
Effective date: December 4, 1997.
Amendment Nos.: 207 and 188.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: July 30, 1997 (62 FR
40860).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 4, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Dated at Rockville, Maryland, this 24th day of December 1997.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 97-33968 Filed 12-30-97; 8:45 am]
BILLING CODE 7590-01-P
