[Federal Register Volume 63, Number 250 (Wednesday, December 30, 1998)]
[Notices]
[Pages 71962-71984]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-34440]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 7, 1998, through December 17, 1998.
The last biweekly notice was published on December 16, 1998 (63 FR
69332).
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
[[Page 71963]]
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 Administration 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 29, 1999, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, 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).
[[Page 71964]]
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 amendment request: October 27, 1998.
Description of amendment request: The Carolina Power & Light
Company, licensee for the Brunswick Steam Electric Plant (BSEP), Unit
Nos. 1 and 2, proposed amendments to the Operating Licenses for the
BSEP units. The amendments are administrative in nature and would
delete various completed license conditions, make editorial changes,
and provide clarifying information.
The licensee has concluded that the proposed license amendments do
not involve a Significant Hazards Consideration. In support of this
determination, an evaluation of each of the three standards set forth
in 10 CFR 50.92 is provided below.
Basis for a 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 license amendments do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The proposed changes revise the BSEP, Unit Nos. 1 and 2, Facility
Operating Licenses to delete various license conditions that have been
completed, make editorial changes, and provide clarifying information.
The changes are administrative and only provide updated and clarifying
information. No physical or operational changes to the facility will
result from the proposed changes. Therefore, the proposed license
amendments do not involve an increase in the probability or
consequences of an accident previously evaluated.
2. The proposed license amendments will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed changes revise the BSEP, Unit Nos. 1 and 2, Facility
Operating Licenses to delete various license conditions that have been
completed, make editorial changes, and provide clarifying information.
The changes are administrative and only provide updated and clarifying
information. The proposed license amendments do not alter any plant
operation and will not result in a physical change to the facility.
Therefore, the proposed license amendments do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The proposed changes revise the BSEP, Unit Nos. 1 and 2, Facility
Operating Licenses to delete various license conditions that have been
completed, make editorial changes, and provide clarifying information.
The changes are administrative and only provide updated and clarifying
information. No physical or operational changes to the facility will
result from the proposed changes. Therefore, the proposed license
amendments do not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: Frederick J. Hebdon.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendment request: November 30, 1998.
Description of amendment request: This amendment request proposes
to relocate, to a licensee controlled document, the requirement for
removal of the Reactor Protection System (RPS) shorting links. Removal
of the shorting links enables a non-coincident scram on high neutron
flux as detected by the Source Range Monitors (SRMs).
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:
Does the change involve a significant increase in the probability
or consequences of an accident previously evaluated?
The RPS shorting links are not precursors to any previously
evaluated accident. The Source Range Monitors (SRMs), and the ability
of the SRMs to provide a RPS trip, are also not precursors to any
previously evaluated accident. Therefore, relocating the RPS shorting
link requirement to administrative controls [the Updated Final Safety
Analysis Report (UFSAR)] will not increase the probability of an
accident previously evaluated.
The RPS shorting links are not assumed to be removed in any
accident analysis, and the SRMs are not assumed to provide a RPS trip
in any accident analysis. The refueling interlocks and SHUTDOWN MARGIN
calculations will continue to provide assurance of reactivity control.
Therefore, relocating the RPS shorting link requirements to
administrative controls [the UFSAR] will not increase the consequences
of an accident previously evaluated.
The RPS shorting link requirements will be relocated to
administrative controls that are administered pursuant to the
requirements of 10 CFR 50.59, thereby reducing the level of regulatory
control. The level of regulatory control has no impact on the
probability or consequences of an accident previously evaluated.
Consequently, this proposed amendment does not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different kind
of accident from any accident previously evaluated?
Relocating the RPS shorting link requirements to administrative
controls [the UFSAR] does not create any new failure mechanisms. No new
equipment will be installed or utilized, and no new operating
conditions will be initiated as a result of this change. Therefore, the
proposed change does not create the possibility of a new or different
kind of accident from any previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The refuel interlocks and SHUTDOWN MARGIN calculations will
continue to ensure that the reactor stays
[[Page 71965]]
subcritical in the Refuel Mode. The margin to safety as represented by
the SHUTDOWN MARGIN designed into the core and verified in the SHUTDOWN
MARGIN calculations will be unaffected by relocation of the RPS
shorting link requirements to administrative controls [the UFSAR]. The
margin to safety as represented by the fuel bundle drop assumptions
protected by the refuel interlocks will be unaffected. In addition, no
accident analysis assumes that the RPS shorting links are removed. In
addition, the RPS shorting link requirements will be relocated to
administrative controls [the UFSAR] for which future change will be
evaluated pursuant to the requirements of 10 CFR 50.59. Therefore,
there will be no change in the types or significant increase in the
amounts of any effluents released offsite, and, thus, these changes do
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments requested involve no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Stuart A. Richards.
Florida Power Corporation, et al. (FPC), Docket No. 50-302, Crystal
River Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County,
Florida
Date of amendment request: October 30, 1998 (LAR-236).
Description of amendment request: The proposed amendment would
change the Crystal River Unit 3 (CR-3) Improved Technical
Specifications (ITS) Section 5.6.2.19, Section 3.4.11, Bases 3.4.11 and
Bases 3.4.3. The changes reflect the use of fluence methodology
described in Topical Report BAW-2241P, ``Fluence and Uncertainty
Methodologies,'' and the use of American Society of Mechanical
Engineers (ASME) Code Case N-514, ``Low Temperature Overpressure
Protection,'' for developing Low Temperature Overpressure Protection
(LTOP) limits. Reference to Topical Report BAW-1543A, ``Integrated
Reactor Vessel Surveillance Program,'' was also added to ITS Section
5.6.2.19. ITS Section 3.4.11 (Low Temperature Overpressure Protection
System), was revised to reflect the new LTOP limits based on revised
fluence projections through 32 Effective Full Power Years (EFPY). The
Pressure/Temperature (P/T) Limits Report is being revised to reflect
the new P/T limits for heatup, cooldown, hydrostatic and leak test, and
to incorporate the CR-3 LTOP curve.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
LAR [License Amendment Request] #236 proposes several changes to
the ITS operational limits. These changes are being proposed to
maintain the necessary margins of safety through 32 EFPY using analyses
based on methodologies that have been previously approved for use at
CR-3, ASME Code Case N-514 and LTOP SER [Safety Evaluation Report], and
are currently being reviewed by the NRC staff:
--NRC to FPC letter, 3N1293-30, dated December 20, 1993, ``Crystal
River Unit 3--Issuance of Amendment RE: Improved Technical
Specifications (TAC No. M74563)''
--NRC to FPC letter, 3N1297-16, dated December 22, 1997, ``Crystal
River Unit 3--Staff Evaluation and Issuance of Amendment RE: Low-
Temperature Overpressure Protection (TAC No. M99277)''
--NRC to FPC letter, 3N079705, dated July 3, 1997, ``Crystal River 3--
Exemption from Requirements of 10 CFR 50.60, Acceptance Criteria for
Fracture Prevention for Lightwater Nuclear Power Reactors for Normal
Operation (TAC No. M98380)''
--BAW-2241P, ``Fluence and Uncertainty Methodologies''
The limiting transient for LTOP remains a failed-open makeup valve.
Existing LTOP controls (maximum of one makeup pump capable of injecting
into the RCS [reactor coolant system], high pressure injection (HPI)
deactivated, the CFTs [core flood tanks] isolated, pressure relief
capability and maintaining a gas volume in the RCS) remain unchanged
from the current ITS 3.4.11 as approved by Reference 3, except the
setpoints proposed herein. The setpoints are being updated to reflect
the new 32 EFPY fluence analysis and P/T limits. Therefore, this change
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes will not create the possibility of a new or
different kind of accident from any previously evaluated since they do
not introduce new systems, failure modes or plant perturbations.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not involve a significant reduction in
the margin of safety since the proposed P/T limitations have been
developed consistent with the requirements of 10 CFR 50.60. The
operational limits have been developed to maintain the necessary
margins of safety as defined by ASME through 32 EFPY using
methodologies previously reviewed and approved by the NRC. The
objective of these limits is to prevent non-ductile failure during any
normal operating condition, including anticipated operational
occurrences and system hydrostatic tests.
The LTOP safety factors are based on reanalyzed conditions for 32
EFPY of operation utilizing methodology contained in ASME Code Case N-
514 which has been approved for use at CR-3. The Code Case provides an
acceptable margin of safety against flaw initiation and reactor vessel
failure. The application of Code Case N-514 for CR-3 ensures an
acceptable level of safety. Therefore, this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC-A5A, P. O. Box 14024, St. Petersburg, Florida
33733-4042.
NRC Project Director: Frederick J. Hebdon.
[[Page 71966]]
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida
Date of amendment request: October 30, 1998.
Description of amendment request: The proposed amendment requests
approval of a change to the Crystal River Unit 3 (CR-3) Final Safety
Analysis Report (FSAR) regarding the methodology for performing the
Spent Fuel Pool (SFP) B criticality analysis. Recent Boraflex samples
from the SFP B demonstrate a weight loss in excess of the available
margin within the current licensing basis calculation. The criticality
analysis calculations proposed in this amendment request demonstrate
that the burnup/enrichment curves in the current Improved Technical
Specifications (ITS) have sufficient margin to accommodate up to a 20%
loss in Boraflex neutron absorption, and still maintain SFP B at less
than or equal to 0.95 k-effective when fully loaded and flooded with
unborated water. Florida Power Corporation has concluded that the
change in the criticality analysis methodology represents an unreviewed
safety question, and thus requires prior NRC approval.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No. The two possible accidents are: (1) criticality during normal
storage and (2) criticality due to a misloaded fuel assembly during
handling fuel. Each are discussed below:
(1) Criticality during normal storage.
For criticality during normal storage to occur, there must be a
loss of negative reactivity since an addition of positive reactivity is
not possible without fuel movement. A loss in negative reactivity could
result only from reduction in Boraflex inventory below that needed to
meet the design basis. The proposed criticality analysis for Spent Fuel
Pool B demonstrates that Spent Fuel Pool B is capable of maintaining
the design basis requirement of k-effective less than or equal to 0.95
when flooded with unborated water and with a loss of up to 20% of the
Boraflex absorber material. Therefore, allowing up to 20% Boraflex loss
with the new analysis does not significantly increase the probability
of an accident previously evaluated.
(2) Criticality during fuel handling.
Criticality during fuel handling could occur due to loss of
negative reactivity, or the addition of positive reactivity. Loss of
negative reactivity could result from loss of Boraflex as discussed
above.
Addition of positive reactivity would result from the misloading of
fuel in a fashion not in accordance with ITS LCO 3.7.15, such as the
misloading of a fresh 5.05% enriched fuel assembly into Region 2 or
side-by-side with another fresh fuel assembly in Region 1. The minimum
required boron concentration of ITS LCO 3.7.14 and CR-3 FSAR 9.3.2.1.2
are intended to compensate for just such an accident. Consistent with
the double-contingency principle, a boron dilution is not required to
be considered concurrent with a misloaded new fuel assembly (bases of
ITS LCO 3.7.14). The use of a new calculational method will not
increase the probability of fuel assembly misloading. A boron dilution
event without an accompanying misloaded fuel assembly is not impacted
by the new criticality analysis, since the design basis allows for
unborated water for normal storage conditions.
Therefore, since the proposed criticality analysis does not
increase the probability of a misloaded fuel assembly, the probability
of an occurrence of an accident previously evaluated is not
significantly increased.
Boraflex is credited with preventing inadvertent criticality. It is
not credited with mitigating the effects, or dose consequences, to the
public or to plant personnel from an inadvertent criticality. The
criticality analysis does not affect or mitigate the dose consequences
to the public or plant personnel from an inadvertent criticality.
There are no other SAR accidents that could be affected. Therefore,
the use of the proposed criticality analysis, does not significantly
increase the consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
No. The only purpose, or function, of Boraflex is reactivity
control. Therefore, the use of the proposed criticality analysis can
only result in reactivity related accidents, such as an inadvertent
criticality. Though a spent fuel pool criticality accident is not
discussed in detail, a calculation to ensure such an accident could not
occur is referenced by both FSAR 9.3 and 9.6. Therefore, this is an
accident already discussed by the SAR and dependence on a new
criticality analysis does not create the possibility of an accident of
a new or different kind than any previously evaluated.
3. Involve a significant reduction in a margin of safety.
No. The proposed analysis demonstrates that the safety function and
design basis are met even for a Boraflex loss of up to 20%. Though the
proposed criticality analysis methodology is more realistic, and has
been licensed at other sites, it is less conservative than the
existing, NRC approved analysis that is currently part of the CR-3
licensing basis. Additionally, it permits operation with a greater loss
of Boraflex than the existing analysis.
The current licensing basis, BAW-2209, ``Crystal River Unit 3 Spent
Fuel Storage Pool Criticality Analysis'', provides the analytical basis
of both ITS LCO 3.7.14 and LCO 3.7.15. This analysis uses very
conservative assumptions and methodologies, and results in very little
margin remaining for identified Boraflex loss. The margin of safety,
although less than previously evaluated, is not significantly reduced
with reliance on the current criticality analysis. The margin of safety
is restored with use of the proposed criticality analysis. Therefore,
the margin of safety is not significantly reduced with use of the
proposed criticality analysis.
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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Project Director: Frederick J. Hebdon.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida
Date of amendment request: November 23, 1998.
Description of amendment request: The proposed amendment would
change the CR-3 Improved Technical Specifications (ITS) to raise the
Engineered Safeguards Actuation System (ESAS) setpoint for reactor
coolant system (RCS) low pressure from
[[Page 71967]]
1500 psig to 1625 psig. This change is intended to provide for earlier
actuation of high pressure injection (HPI) following certain small
break loss of coolant accidents and result in a lower peak center line
temperature (PCT) during these transients. The applicability
requirement for ESAS operability would be changed from greater than
1700 psig to greater than 1800 psig to maintain the previous margin
above the ESAS setpoint. Similarly, the reactor protection system (RPS)
setpoint for RCS low pressure and the RPS setpoint for Shutdown Bypass
(RCS High Pressure) would each be raised by 100 psig to maintain the
previous pressure margins. In addition, Surveillance Requirement
3.5.2.5 would be revised such that valves in the HPI flowpath that are
throttled to balance flow between the four HPI lines would be verified
in the correct position. The need for these changes resulted from
planned modifications to the HPI system to improve performance and
reliability of this system. Changes to ITS Bases necessitated by the
system modifications and setpoint changes are included in the
submittal.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The setpoint changes for reactor trip and High Pressure Injection
(HPI) actuation will result in a very small (approximately one-percent)
increase in the probability for reactor trips. Review of industry data
shows that this increase is not significant. The revised accident
analysis has determined that transients which reduce Reactor Coolant
System (RCS) pressure below the new setpoints, warrant the associated
action. Engineered Safeguards Actuation System (ESAS) and Reactor
Protection System (RPS) actuations are used to mitigate accidents and
are not the initiator of analyzed accidents. Therefore, the probability
of previously evaluated accidents is not affected.
RPS and ESAS functions are assumed to actuate to mitigate
transients. The revised setpoints will ensure earlier actuation of the
RPS and ESAS on a low RCS pressure condition. Raising the ESAS Low RCS
Pressure Setpoint will ensure earlier automatic HPI actuation for a
portion of the spectrum of pressure decreasing events. For rapid
depressurization events, such as main steam line break and large break
Loss of Coolant Accident (LOCA), this will have little impact. For
slower events, or those that do not reach the current setpoint during
the initial subcooled blowdown phase, HPI will be automatically
initiated substantially earlier in the event. This will increase the
integrated HPI flow to the RCS during the time the core is likely to be
uncovered, thereby reducing the consequential PCT. This additional flow
results in a significant peak clad temperature (PCT) decrease for small
break LOCA scenarios less than 0.07 square feet. Based on the above,
the consequences of previously evaluated accidents will not be
increased.
The HPI system characteristics will not be affected such that the
probability of any accident is increased. The system flow restriction
for protection from low temperature overpressure (LTOP) events will be
maintained. The HPI system is used for accident mitigation and is not
the initiator of evaluated accidents other than LTOP. The proposed
surveillance changes will ensure that all valves throttled in the HPI
flowpath are verified and secured in the correct position. The throttle
valves and stop check valves will be positioned to ensure HPI flow is
within analyzed limits. Therefore, the consequences of accidents that
rely on HPI flow will not be increased.
Based on the above evaluation, the probability or consequences of
evaluated accidents are not significantly increased by these changes.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The change to RPS and ESAS setpoints will not change the functions
of plant equipment, no new system interactions will be created, and no
new failure modes will be introduced. The setpoint changes will permit
earlier actuation for the associated actions. However, no new plant
conditions will be introduced by the setpoint changes.
The HPI modifications include the installation of throttle valves
that will change the flow characteristics of the system. The new
throttle valves are manual valves that will be secured in position. The
revised surveillance requirements will ensure these valves are
positioned such that HPI flow is within analyzed limits. Therefore, no
conditions are created that could cause a new type of accident.
Based on the above evaluation, these changes cannot create the
possibility of an accident of a different type than previously
evaluated in the [Safety Analysis Report] SAR.
3. Does not involve a significant reduction in the margin of
safety.
The safety function of the affected portions of the RPS and ESAS
systems is to actuate their respective functions if RCS pressure drops
below the setpoint. The raised RPS and ESAS setpoints will provide
earlier actuation for these protective features. These changes will
increase the margin of safety provided by the associated Technical
Specifications.
The safety function of the HPI system is to provide cooling to
limit fuel peak clad temperature. The revised surveillance requirements
will ensure valves are positioned such that HPI flow is within analyzed
limits. Therefore, the margin of safety provided by the HPI
surveillance requirements is maintained.
Based on the above evaluation, there is no reduction in the margin
of safety associated with the equipment and systems affected by this
change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Project Director: Frederick J. Hebdon.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: December 3, 1998.
Description of amendment request: The proposed change revises the
TMI-1 Core Protection Safety Limits and Core Protection Safety Bases,
as specified in Technical Specification Figures 2.1-1 and 2.1-3, to
provide more restrictive limits which reflect the decrease in reactor
coolant system flow resulting from the analysis of increased once-
through steam generator (OTSG) tube plugging limits (total allowable
number of tubes plugged). The licensee is currently restricted to a
total of 2,000 tubes plugged in both OTSGs which corresponds to 6.4
percent of the total number of tubes. The licensee's more restrictive
Core Protection Safety Limits reflect the reduction in reactor coolant
[[Page 71968]]
flow that would exist if an average of 20 percent of the OTSG tubes
were plugged.
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
of occurrence or the consequences of an accident previously evaluated.
An increase in the average steam generator tube plugging (SGTP) level
to 20% results in a small reduction of reactor coolant system (RCS)
flow rates and primary to secondary heat transfer. These changes result
in small changes to the primary and secondary side operating
parameters, and do not result in any additional challenges to plant
equipment. The proposed Technical Specification Changes resulting from
the increase in allowable tube plugging limits are more restrictive but
remain bounded by the existing reactor protection system (RPS) trip
setpoints. The assessment of the NSSS [nuclear steam supply system]
primary components, including the reactor pressure vessel, reactor
core, reactor coolant pump, steam generator, pressurizer, control rod
drive mechanisms, and RCS piping concluded that the integrity of these
components will be unaffected by the increase in average SGTP level.
A re-analysis of the bounding Updated Final Safety Analysis Report
(UFSAR) Chapter 14 accidents, specifically the startup accident, loss
of coolant flow, loss of feedwater, and large and small break LOCA
demonstrated compliance with the acceptance criteria. The RCS pressure
boundary is not challenged, and the DNBR [departure from nucleate
boiling ratio] and peak clad temperature values remain within the
specified limits of the licensing basis. An analysis of the loss of
electric power accident demonstrated the ability of the plant to
transition smoothly to natural circulation with an average of 20% SGTP
or with asymmetric plugging. It was also determined that the current
mass and energy release data used for the containment integrity and
equipment qualification remain bounding. Since the design requirements
and safety limits continue to be met, system functions are not
adversely impacted, and the integrity of the RCS pressure boundary is
not challenged, the radiological consequences remain unchanged.
Therefore, this activity does not involve a significant increase in the
probability of occurrence or the consequences of an accident previously
evaluated.
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 previously evaluated. The proposed Technical
Specification changes are more restrictive core protection safety
limits but remain bounded by the existing RPS trip setpoints. This
proposed change assures safe operation commensurate with the effects of
steam generator tube plugging. This increase in the average level of
SGTP to 20% will not introduce any new accident initiator mechanisms.
No new failure modes or limiting single failures have been identified.
Since the safety and design requirements continue to be met and the
integrity of the RCS pressure boundary is not challenged, no new
accident scenarios have been created. This change does not add any new
equipment, modify any interfaces with existing equipment, or change the
equipment function or the method of operating the equipment. Reactor
core, RCS, and steam generator parameters remain within appropriate
design limits during normal operation. Therefore, this activity does
not create the possibility of a new or different kind of accident from
any previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The existing RPS trip setpoints bound the proposed Technical
Specification changes resulting from 20% SGTP. This change assures safe
operation commensurate with the effects of steam generator tube
plugging. The TMI-1 DNB design basis, RCS pressure limits, peak clad
temperature limits and dose criteria are maintained for all UFSAR
transients. Therefore, this activity does not reduce 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: Law/Government Publications
Section, State Library of Pennsylvania (REGIONAL DEPOSITORY), Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of amendment request: November 16, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) related to the implementation of
systems for the detection and suppression of coupled neutronic/thermal-
hydraulic instabilities in the reactor. Average Power Range Monitor
(APRM) flow control trip reference cards will initiate a reactor scram
to limit the oscillation magnitude at reactor trip so as to limit the
associated Critical Power Ratio change and, in conjunction with Minimum
Critical Power Ratio (MCPR) operating limits, assure compliance with
the MCPR safety limit. In addition, the changes would increase the APRM
flow biased neutron flux scram and control rod block settings to allow
plant operation in the Extended Load Line Limit Analysis region. Thus,
the proposed changes are in regard to setpoints and calculations for
fuel cladding integrity and the associated TS Bases. In the Bases for
TS 2.1.1, the proposed change would reference new equations in TS
2.1.2a. In TS 2.1.2a, the proposed change would be to the equation for
determining the flow biased APRM scram and rod block trip setpoints. In
the Bases for TS 2.1.2a, the proposed change would reflect the new
setpoints. In the Bases for TS 2.2.2, the proposed change would be to
the description of the setpoint methodology which is based upon General
Electric Report NEDC-31336, ``GE Instrumentation Setpoint
Methodology.'' In Note (m) of TS Table 3.6.2/4.6.2, the proposed change
would be to the calibration range for the APRM channel setpoint. In the
Bases for TS 3.6.2/4.6.2, the proposed change would be to the equations
and methodology for determining APRM scram and rod block setpoints. In
TS 6.9.1.f, which identifies documents approved by NRC for analytical
methods used to determine core operating limits, the proposed change
would add ``NEDO-32465-A, Reactor Stability Detect and Suppress
Solutions Licensing Basis Methodology for Reload Applications, August
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
[[Page 71969]]
consideration, which is presented below:
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The APRM neutron monitoring system is not an initiator or a
precursor to an accident. The neutron monitoring system monitors the
power level of the reactor core and provides automatic core protection
signals in the event of a power transient. A Restricted Region will be
maintained such that the probability of a stability event is not
increased. Therefore, the proposed TS changes cannot affect the
probability of a previously evaluated accident.
The proposed TS changes will revise the APRM flow-biased neutron
flux scram TS setting to provide automatic protection to assure that
anticipated coupled neutronic/thermal-hydraulic instabilities will not
compromise established fuel safety limits. The proposed changes will
result in a more restrictive APRM flow-biased scram trip setting in the
low flow regions of the power/flow operating map (i.e., operational
conditions where reactor instabilities are most probable). In other
words, the new settings will provide a scram sooner (at a lower power
level) than the existing settings. The associated control rod block
setting will also be revised. A margin between the control rod block
and flux scram has been determined by calculation.
The proposed changes will also revise the APRM flow-biased neutron
flux scram and control rod block TS settings to provide an increase
above the current values in operating conditions not susceptible to
reactor instabilities. Specifically, the proposed changes will
implement a 2% increase in the analytical limit of the APRM flow-biased
flux scram and a 7% increase in the analytical limit of the APRM flow-
biased control rod block. Evaluation demonstrates that these proposed
analytical limit increases have negligible impact on the transient
events results for NMP1 [Nine Mile Point Unit 1] as documented in
Chapter XV of the NMP1 UFSAR, [Updated Final Safety Analysis Report],
including the limiting transient events which are reanalyzed each
reload. Of the twenty-five (25) transient events analyzed in Section XV
of the NMP1 UFSAR, only the Inadvertent Startup of Cold Recirculation
Loop event and the Recirculation Flow Controller Malfunction--Increase
Flow event have potentially impacted results. The Chapter XV Control
Rod Drop Accident as well as the Turbine Trip with No Bypass at Partial
Power event were also evaluated.
For the Inadvertent Startup of Cold Recirculation Loop event, the
proposed 2% increase in the high neutron flux scram would result in an
increase in the fuel average surface heat flux response. However, there
is significant margin between the surface heat flux value for this
event and the current limiting MCPR [Minimum Critical Power Ratio]
event (the Feedwater Controller Failure Maximum Demand event). As such,
any small change to the fuel surface heat flux response due to the high
neutron flux scram analytical limit increase would not result in the
fuel thermal margin requirements for the Inadvertent Startup of Cold
Recirculation Loop event to exceed the MCPR limits set by the limiting
reload analysis event.
The reactor neutron flux for the Recirculation Flow Controller
Malfunction--Increase Flow event also showed an increasing trend from
its initial value. However, the peak response for this parameter (104%
of rated) is significantly below the high neutron flux scram analytical
limit. Accordingly, the proposed increase to the high neutron flux
scram analytical limit does not affect the response to this transient
event.
The Control Rod Drop Accident is included in Chapter XV of the NMP1
UFSAR. As noted in NEDE-24011-P-A, ``GESTAR II: General Electric
Standard Application for Reactor Fuel,'' the initial power burst from
this event is terminated by the Doppler reactivity feedback while the
scram provides the final event termination several seconds later. The
120% APRM scram limit was conservatively chosen. The time delay
introduced by the small change in analytical limit will be
inconsequential due to the extremely rapid power rise for this event
(i.e., the time of scram for a 120% analytical limit vs. a 122%
analytical limit is essentially the same).
The proposed Bases changes to TS 3.6.2/4.6.2 and TS 2.2.2 simply
provide details of the setpoint methodology currently used as well as
specific allowable values.
Therefore, the proposed TS changes to implement a more restrictive
flow-biased scram setting to protect against reactor instabilities and
the proposed change to increase the high neutron flux scram and rod
block analytical limits do not result in a significant increase in the
consequences of an accident previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes will revise the APRM flow-biased neutron flux
scram TS settings to assure anticipated coupled neutronic/thermal-
hydraulic instabilities will not compromise established fuel safety
limits in the low flow regions of the power/flow operating map as well
as revise the associated control rod block settings. These changes also
propose a 2% increase in the analytical limit of the APRM flow-biased
neutron flux scram and a 7% increase in the analytical limit of the
APRM flow-biased control rod block. These changes do not introduce any
new accident precursors and do not involve any alterations to plant
configurations which could initiate a new or different kind of
accident. The proposed changes do not affect the intended function of
the APRM system nor do they affect the operation of the system in a way
which would create a new or different kind of accident.
Therefore, the proposed changes will not create the possibility of
a new or different kind of accident from any previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
More conservative APRM flow-biased neutron flux scram and control
rod block settings will be implemented in the low flow regions of the
power/flow operating map. The scram setting change will assure that
anticipated coupled neutronic/thermal-hydraulic instabilities will not
compromise established fuel safety limits. The proposed changes will
also implement a 2% increase in the APRM flow-biased neutron flux scram
and a 7% increase in the APRM flow-biased control rod block in those
operating regions not susceptible to reactor instabilities. Evaluation
demonstrates that these proposed increases have negligible impact on
the transient events or accident results for NMP1. The impacted
transient events are either not the limiting MCPR event, the peak
response to the event is significantly below the high neutron flux
scram analytical limit or in the case of the Control Rod Drop Accident,
the time delay introduced by the change will be inconsequential due to
the extremely rapid power rise. No other events are adversely affected.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 71970]]
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 (NMP2), Oswego County, New York
Date of amendment request: November 19, 1998.
Description of amendment request: The proposed amendment would
change the surveillance frequencies in Technical Specifications (TSs)
4.8.4.4a, ``Surveillance Requirements--Reactor Protection System
Electric Power Monitoring (RPS Logic),'' and 4.8.4.5a, ``Surveillance
Requirements--Reactor Protection System Electric Power Monitoring
(Scram Solenoids),'' to require channel functional testing of the RPS
Motor Generator Set (M/G) and RPS Uninterruptible Power Supplies (UPS)
Electrical Protection Assemblies (EPAs) at least once every 6 months.
These TSs currently require that channel functional testing be
performed each time the plant is in cold shutdown for a period of more
than 24 hours, unless performed within the previous 6 months.
Basis for proposed no significant hazards consideration
determination: During the last refueling outage, the licensee modified
the Nine Mile Point Unit No. 2 (NMP2) design for the RPS M/G and RPS
UPS EPAs to provide relay actuated protection systems. The relays of
the new design may be individually isolated from an essential power
circuit for testing and may be actuated without tripping the associated
breaker. The relay actuated system will allow the EPA system monitoring
an essential power supply to be functionally tested with the plant on-
line. The EPA relay actuation setpoints are not affected by the
modification or the proposed TS changes. The licensee states that the
design, installation, and testing of the new units meet the criteria of
the same standards that were applied to the previous units.
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 changes affect surveillance testing frequency only.
The new relay actuated protection system design functions in the same
fail safe manner as the old units. Also, the new design in conjunction
with the testing capability has increased EPA reliability, while
introducing little risk to testing the EPAs with the plant in
operation. Therefore, the proposed changes to the NMP2 TS do 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 changes affect surveillance testing frequency of relay
actuated protection circuits only. The proposed changes do not
introduce any new or different accident initiators from any that were
previously evaluated. EPA relay actuation setpoints are not affected.
The actual fail safe system conditions required for EPA actuation will
remain the same. Therefore, the operation of NMP2, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The 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 function of the EPA systems is to isolate the loads from supply
power. That function was not altered by the proposed change.
Reliability of the EPA systems is improved. Therefore, the operation of
NMP2, in accordance with the proposed amendment, will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: S. Singh Bajwa.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station, Unit 1 (NMP1), Oswego County, New York
Date of amendment request: November 30, 1998.
Description of amendment request: The proposed amendment would
correct Technical Specification (TS) 3.1.2, ``Liquid Poison System,''
and the associated TS Bases. Specifically, in the Bases for TS 3.1.2,
the boron-10 concentration of 120 ppm (which is incorrectly calculated
using atomic percent instead of weight percent) would be changed to
109.8 ppm. In TS 3.1.2, the minimum volume of the sodium pentaborate
solution contained in the Liquid Poison System storage tank would be
increased from 1185 gallons to 1325 gallons.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The Liquid Poison System is designed to provide the capability to
bring the reactor from a full design rating to a shutdown condition
assuming none of the control rods can be inserted. The system is
manually initiated in response to a failure of the Control Rod Drive
System to shutdown the reactor. The proposed changes revise the
required liquid poison solution volume and concentration. The proposed
changes to the Technical Specifications and the Bases require no
changes to the physical facility which could adversely affect any
accident precursors. Therefore, the proposed changes cannot
significantly increase the probability of an accident.
The proposed changes will assure that the Liquid Poison System
continues to provide the capability to shutdown the reactor during an
ATWS [Anticipated Transient Without Scram] event. In addition, the
system will continue to be capable of bringing the reactor to cold
shutdown, 3 percent delta k subcritical (0.97 keff), from a
full design rating of
[[Page 71971]]
1850 megawatts thermal assuming none of the control rods can be
inserted, and considering the combined effects of coolant voids,
temperature change, fuel doppler, and xenon and samarium. Therefore,
the change to the Technical Specifications does not significantly
increase the consequences of a previously evaluated accident.
2. The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Injection of the sodium pentaborate solution into the reactor
vessel has been considered in the plant design. The proposed changes
revise the required liquid poison solution volume and concentration.
The proposed changes make no physical modification to the plant which
could create the possibility of a new or different kind of accident.
The proposed changes will maintain the capability of the Liquid Poison
System to shutdown the reactor from its full design rating assuming
none of the control rods are inserted, and considering the combined
effects of coolant voids, temperature change, fuel doppler, and xenon
and samarium. Consequently, these changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed changes revise the required liquid poison solution
volume and concentration. The proposed changes make no physical
modification to the plant which could reduce the margin of safety.
These changes will assure compliance with the requirements of
10CFR50.62, ``Requirements for Reduction of Risk from Anticipated
Transients without Scram (ATWS) Events for Light-Water-Cooled Nuclear
Power Plants.'' In addition, these changes will maintain the capability
of the Liquid Poison System to bring the reactor from a full design
rating of 1850 megawatts thermal to greater than 3 percent delta k
subcritical (0.97 keff) assuming none of the control rods
can be inserted, and considering the combined effects of coolant voids,
temperature change, fuel doppler, xenon and samarium.
The required volume of boron-10 solution in the Liquid Poison
System storage tank includes an additional 25 percent margin beyond the
amount needed to shutdown the reactor to allow for any unexpected non-
uniform mixing. Also, the total storage tank volume of sodium
pentaborate solution incorporates 197 gallons of solution which is
unavailable for injection into the reactor vessel and a 25 gallon
margin for conservatism. Additionally, using one 30 gpm Liquid Poison
System pump, the injection time is greater than 17 minutes thereby
assuring adequate mixing. The proposed changes to the liquid poison
concentration and volume ensure the NMP1 [Nine Mile Point Unit 1]
Liquid Poison System is able to meet its safety function requirements.
Therefore, this change will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: S. Singh Bajwa.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: December 4, 1998.
Description of amendment request: The proposed amendment would
eliminate the need to cycle the plant and its components through a
shutdown-startup cycle by allowing the next snubber surveillance
interval to be deferred until the end of refueling outage 6 or
September 10, 1999, whichever date is earlier.
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:
NNECO has reviewed the proposed revision in accordance with 10 CFR
50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this conclusion
is that the three criteria of 10CFR50.92(c) are not satisfied. The
proposed revision does not involve [an] SHC because the revision would
not:
1. Involve a significant increase in the probability or consequence
of an accident previously evaluated.
The proposed change is for a one time extension to the surveillance
interval of snubber inspections required by Technical Specification
4.7.10.e. The change involves revising the calendar time for snubber
interval inspections to 36 months to coincide with the time frame of
the current cycle 6 operation.
Snubber testing experience at Millstone Unit No. 3 has shown that
historical failure rates of snubbers are low. During the third
refueling outage, after an operating cycle of approximately 22 months,
the functional testing program identified multiple Type A failures
attributed primarily to original plant construction, and resulted in a
full inspection of all Type A snubbers. The snubber inspection interval
was extended to approximately 30 months by a one-time extension to the
Technical Specifications for the fourth refueling outage and only one
Type A snubber failure was identified. Subsequent outages with
operating durations of 18 and 17 months also identified only a single
Type B failure in each outage. The results of piping stress analysis
which have been performed to assess the impact of snubbers which have
failed to meet functional test acceptance criteria have shown that
neither piping system functionality or structural integrity have ever
been compromised.
During the recent cycle 6 operation Millstone 3 has experienced an
extended midcycle shutdown, where temperature, vibration effects and
normal wear on snubbers have been minimized as compared to a normal
operating cycle. The last snubber surveillance interval inspections
were completed during this midcycle shutdown. Although the calendar
surveillance interval is impacted by this change the primary conditions
that present challenges to snubbers have not been prevalent during the
extended shutdown. Given the low failure rates of snubbers over the
last 3 surveillance intervals, and the fact the operating time of the
remainder of cycle 6 will be approximately 1 year, snubber failures are
expected to be similar to previous intervals.
Accordingly the possibility of a snubber failure leading to a
Decrease in Reactor Coolant Inventory or a Decrease in Heat Removal by
the Secondary System is not increased and there is no affect on the
probability of previously evaluated accidents.
This change does not include any physical changes to the plant and
does not affect acceptance criteria or the
[[Page 71972]]
required actions for functional failures of snubbers. Accordingly there
is no increase in the consequences of previously evaluated accidents
resulting in a Decrease in Reactor Coolant Inventory or a Decrease in
Heat Removal by the Secondary System.
Thus it is concluded that the proposed revision does not involve a
significant increase in the probability or consequence of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
This proposed revision to the surveillance interval does not change
the operation of any plant system or component during normal or
accident conditions. The proposed change extends the surveillance
interval of snubber inspections required by Technical Specification
4.7.10.e. The change involves revising the calendar time for snubber
interval inspections to coincide with the time frame of current cycle 6
operation. This change does not include any physical changes to the
plant and does not affect acceptance criteria or the required actions
for functional failures of snubbers.
Thus, this proposed revision does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change extends the surveillance interval of snubber
inspections required by Technical Specification 4.7.10.e. The change
involves revising the calendar time for snubber interval inspections to
coincide with the time frame of current cycle 6 operation. This change
does not include any physical changes to the plant and does not affect
acceptance criteria or the required actions for functional failures of
snubbers. The service life of the snubbers or parts as required by
Technical Specification 4.7.10.i will not be impacted by this change
since the required replacements have already occurred and no additional
service life dates will expire prior to September 10, 1999.
Thus, it is concluded that the proposed revision does not involve a
significant reduction in a margin of safety.
In conclusion, based on the information provided, it is determined
that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, Attn: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Project Director: William M. Dean.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: November 24, 1998.
Description of amendment request: The proposed amendment would
revise the Ginna Station Improved Technical Specifications description
of the fuel cladding material (TS 4.2.1) and to update the list of
references provided in Specification 5.6.5 for the Core Operating
Limits Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Evaluation of Administrative Changes
The administrative changes [related to the update of references
provided in Specification 5.6.5 for the Core Operating Limits report]
do not involve a significant hazards consideration as discussed below:
1. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed changes
revise Administrative Controls Section 5.6.5.b to update the references
to NRC approved documents which support the analysis for the Heat Flux
Hot Channel Factor in the Core Operating Limits Report and to provide
clarification to the currently applicable methodology. It revises the
Design Features Section 4.2.1 to provide clarification of the types of
zirconium alloy filler rod material that have received previous NRC
approval and to clarify that the application shall be NRC approved.
Section 4.2.1 is revised to clarify that the analyses performed to
verify compliance with the fuel safety design bases shall be cycle
specific. As such, these changes are administrative in nature and do
not impact initiators or analyzed events or assumed mitigation of
accident or transient events. Therefore, these changes do not involve a
significant increase in the probability or consequences of an accident
previously analyzed.
2. Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed
administrative changes do not affect the manner by which the plant is
operated and no new equipment will be installed. The proposed
administrative changes will not impose any new or different
requirements. All original design and performance criteria continue to
be met, and no new failure modes have been created for any system,
component, or piece of equipment. Thus, these changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of safety.
The proposed changes will not reduce a margin of plant safety because
the methodology has been shown to meet all applicable design criteria
and ensure that all pertinent licensing basis acceptance criteria are
met. As such, no question of safety is involved, and the changes do not
involve a significant reduction in a margin of safety.
Evaluation of Less Restrictive Changes
The less restrictive change [related to the fuel cladding material
(TS 4.2.1)] does not involve a significant hazards consideration as
discussed below:
(1) Operation of Ginna Station in accordance with the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The Westinghouse
14 x 14 VANTAGE + fuel assemblies containing fuel rods fabricated with
ZIRLO alloy meet the same fuel assembly and fuel rod design bases as
Westinghouse 14 x 14 OFA [Optimized Fuel Assembly] fuel assemblies in
the other fuel regions. In addition, the 10 CFR 50.46 criteria will be
applied to the fuel rods fabricated with ZIRLO alloy. The use of these
fuel assemblies will not result in a change to the proposed Ginna
Westinghouse 14 x 14 OFA reload design and safety analysis limits. The
ZIRLO alloy is similar in chemical composition and has similar physical
and mechanical properties as that of Zircaloy-4. Thus the cladding
integrity is maintained and the structural integrity of the fuel
[[Page 71973]]
assembly is not affected. The ZIRLO clad fuel rods improve corrosion
resistance and dimensional stability. The use of ZIRLO does not impact
the radiological consequences of accidents previously evaluated in the
Safety Analysis. The RCS [reactor coolant system] isotopic inventory is
negligibly impacted; therefore, changes in postulated releases from the
RCS or the secondary systems are negligible. Assumptions of fuel
melting in the radiological analyses are not based on the type of fuel
cladding. For those accidents where fuel melting is postulated to occur
(control rod ejection, locked [seized] RCP rotor), the amount of fuel
undergoing melting and clad damage using ZIRLO clad is bounded by the
current values used in the Safety Analysis. Therefore, the probability
or consequences of an accident previously evaluated is not
significantly increased.
(2) Operation of Ginna Station in accordance with the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated. The Westinghouse
14 x 14 VANTAGE + fuel assemblies containing fuel rods fabricated with
ZIRLO alloy will satisfy the same design bases as that used for
Westinghouse 14 x 14 OFA fuel assemblies in the other fuel regions.
Since the original design criteria is being met, the fuel rods
fabricated with ZIRLO alloy will not be an initiator for any new
accident. All design and performance criteria will continue to be met
and no new single failure mechanisms have been created. In addition,
the use of these fuel assemblies does not involve any alterations 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) Operation of Ginna Station in accordance with the proposed
change does not involve a significant reduction in a margin of safety.
The Westinghouse 14 x 14 VANTAGE + fuel assemblies containing fuel rods
fabricated with ZIRLO alloy do not change the proposed Ginna
Westinghouse 14 x 14 OFA reload design and safety analysis limits. The
use of these fuel assemblies containing fuel rods fabricated with ZIRLO
alloy will take into consideration the normal core operating conditions
allowed in the Technical Specifications. For each cycle reload core,
these fuel assemblies will be specifically evaluated using approved
reload design methods and approved fuel rod design models and methods
as specified in Technical Specifications. This will include
consideration of the core physics analysis peaking factors and core
average linear heat rate effects. In addition, the 10 CFR 50.46
criteria will be applied each cycle to the fuel rods fabricated with
ZIRLO alloy. Analyses or evaluations will be performed each cycle to
confirm that 10 CFR 50.46 will be met. Therefore, the margin of safety
as defined in the Bases to the Ginna Technical Specifications is not
significantly reduced.
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: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005.
NRC Project Director: S. Singh Bajwa.
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 23, 1998.
Description of amendment requests: The proposed change would revise
the Technical Specifications (TS) to (1) reinstate the log power
reactor trip at or above 4E-5% RATED THERMAL POWER (RTP); (2) reinstate
reactor trips for Reactor Coolant Flow--Low (RCS flow), the Local Power
Density--High (LPD), and the Departure from Nucleate Boiling Ratio--Low
(DNBR); (3) remove the word ``automatically'' from notes (a) and (d) of
Table 3.3.1-1 to clarify that the manual enable of the trip is
permissible; and, (4) clarify that the setpoints on Table 3.3.1-1 are
set relative to logarithmic power, not thermal power.
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 to TS 3.3.1 does not adversely impact
structure, system, or component design or operation in a manner which
would result in a change in the frequency of occurrence of accident
initiation. SCE has re-analyzed the relevant accidents and established
that accident consequences are not significantly increased by the
proposed changes to the bypass-permissive and enable setpoints. The
reactor trip bypass and automatic enable functions are not accident
initiators. Consequently, the proposed TS change will not significantly
increase the probability of accidents previously evaluated. Therefore,
this amendment request 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.
No new or different accidents result from changing the reactor trip
bypass-permissive and automatic enable setpoints. Introducing an
uncertainty band for the enable setpoints delays the mitigation action
of the reactor trip for the design basis analysis for the events that
credit this trip. The enable setpoint itself does not cause any
accident. Therefore, the amendment request 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.
SCE [Southern California Edison Company] has re-analyzed the
accidents and determined that the consequences of the accidents are
within their acceptance criteria under the proposed amendment so that
the margin of safety that bounds the setpoint in both directions
remains intact. The analyses are relatively insensitive to the reactor
trip automatic enable setpoints, and no significant reduction in the
margins of safety ensues from the relatively minor proposed changes to
the bypass-permissive and enable setpoints, nor from establishing
allowable values for these points.
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, Irvine, California 92713.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, P.O. Box 800, Rosemead, California 91770.
[[Page 71974]]
NRC Project Director: William H. Bateman.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: November 23, 1998.
Description of amendment request: The proposed amendment relocates
descriptive design information from Technical Specification 3/4.7.1.1
(Table 3.7-2), regarding orifice sizes for main steam line Code safety
valves, to the Bases section.
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 relocates the orifice size design information
for the main steam line Code safety valves, found in Table 3.7-2, that
does not meet the criteria for inclusion in Technical Specifications as
identified in 10 CFR 50.36(c)(2)(ii). The affected descriptive design
information is not related to any assumed initiators of analyzed events
and is not assumed to mitigate accident or transient events. The
limiting condition for operation for the main steam line Code safety
valves is not altered by the proposed change. The orifice size design
information will be relocated from Table 3.7-2 of Specification 3/
4.7.1.1 to the Bases section for that same Technical Specification and
will be maintained pursuant to 10 CFR 50.59. In addition, surveillance
testing details for this Technical Specification are addressed in
existing surveillance procedures, which are also controlled by 10 CFR
50.59, and subject to the change control provisions imposed by plant
administrative procedures, which endorse applicable regulations and
standards. Therefore, the 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.
The proposed change relocates the orifice size design information
for the main steam line Code safety valves, found in Table 3.7-2, that
does not meet the criteria for inclusion in Technical Specifications as
identified in 10 CFR 50.36(c)(2)(ii). The change does not involve a
physical alteration of the plant (no new or different type of equipment
will be installed) or make changes in the methods governing normal
plant operation. The change will not impose different requirements, and
adequate control of information will be maintained. This change will
not alter assumptions made in the safety analysis and licensing basis.
Therefore, the 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 proposed change relocates the orifice size design information
for the main steam line Code safety valves, found in Table 3.7-2, that
does not meet the criteria for inclusion in Technical Specifications as
identified in 10 CFR 50.36(c)(2)(ii). The change will not reduce a
margin of safety since it has no impact on any safety analysis
assumptions. In addition, the relocated orifice size design information
remains the same as the existing Technical Specifications. Since any
future changes to this orifice size information (that will be located
in the Bases section) will be evaluated per the requirements of 10 CFR
50.59, there is no reduction in a margin of safety.
The proposed change is also consistent with the Westinghouse Plants
(Improved) Standard Technical Specification, NUREG-1431, approved by
the NRC Staff. Revising the Technical Specification to reflect the
approved content of NUREG-1431 ensures no significant reduction in the
margin of safety. Therefore, the 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: John N. Hannon.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: November 11, 1998.
Brief description of amendments: The proposed amendments revise
core safety limit curves and Overtemperature N-16 reactor trip
setpoints based on analyses of the core configuration and expected
operation for Comanche Peak Steam Electric Station (CPSES) Unit 2,
Cycle 5. The changes apply equally to CPSES Units 1 and 2 licenses
since the Technical Specifications are combined.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
A. Revision to the Unit 2 Core Safety Limits
Analyses of reactor core safety limits are required as part of
reload calculations for each cycle. TU Electric has performed the
analyses of the Unit 2, Cycle 5 core configuration to determine the
reactor core safety limits. The methodologies and safety analysis
values result in new operating curves which, in general, permit plant
operation over a similar range of acceptable conditions. This change
means that if a transient were to occur with the plant operating at the
limits of the new curve, a different temperature and power level might
be attained than if the plant were operating within the bounds of the
old curves. However, since the new curves were developed using NRC
approved methodologies which are wholly consistent with and do not
represent a change in the Technical Specification BASES for safety
limits, all applicable postulated transients will continue to be
properly mitigated. As a result, there will be no significant increase
in the consequences, as determined by accident analyses, of any
accident previously evaluated.
B. Revision to Unit 2 Overtemperature N-16 Reactor Trip Setpoints
As a result of changes discussed, the Overtemperature reactor trip
setpoint has been recalculated. These trip setpoints help ensure that
the core safety limits are protected and that all applicable limits of
the safety analysis are met.
Based on the calculations performed, no significant changes to the
safety
[[Page 71975]]
analysis values for Overtemperature reactor trip setpoint were
required. The f(delta I) trip reset function was revised due to less
top-skewed axial power distributions predicted for this cycle. The
analyses performed show that, using the TU Electric methodologies, all
applicable limits of the safety analysis are met. This setpoint
provides a trip function which allows the mitigation of postulated
accidents and has no impact on accident initiation. Therefore, the
changes in safety analysis values do not involve an increase in the
probability of an accident and, based on satisfying all applicable
safety analysis limits, there is no significant increase in the
consequences of any accident previously evaluated.
In addition, sufficient operating margin has been maintained in the
overtemperature setpoint such that the risk of turbine runbacks or
unnecessary reactor trips due to upper plenum flow anomalies or other
operational transients will be minimized, thereby, reducing potential
challenges to the plant safety systems.
C. Administrative changes to reflect plant nomenclature
Changes to the N-16 trip setpoint equation are for clarification
only to more accurately reflect CPSES plant nomenclature. This change
is administrative in nature and does not increase in the probability or
consequences of an accident previously evaluated.
Summary
The changes in the amendment request apply NRC approved
methodologies to changes in safety analysis values, new core safety
limits and new N-16 setpoint and parameter values to assure that all
applicable safety analysis limits have been met. The potential for an
operational transient to occur has not been affected and there has been
no significant impact on the consequences of any accident previously
evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes involve the calculation of new reactor core
safety limits and overtemperature reactor trip setpoint resets. As
such, the changes play an important role in the analysis of postulated
accidents but none of the changes effect plant hardware or the
operation of plant systems in a way that could initiate an accident.
Changes to the N-16 trip setpoint equation are for clarification only
to more accurately reflect CPSES plant nomenclature. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
In reviewing and approving the methods used for safety analyses and
calculations, the NRC has approved the safety analysis limits which
establish the margin of safety to be maintained. While the actual
impact on safety is discussed in response to question 1, the impact on
margin of safety is discussed below:
A. Revision to the Unit 2 Reactor Core Safety Limits
The NRC-approved TU Electric reload analysis methods have been used
to determine new reactor core safety limits. All applicable safety
analysis limits have been met. The methods used are wholly consistent
with Technical Specification BASES 2.1 which is the bases for the
safety limits. In particular, the curves assure that for Unit 2, Cycle
5, the calculated DNBR is no less than the safety analysis limit and
the average enthalpy at the vessel exit is less than the enthalpy of
saturated liquid. The acceptance criteria remains valid and continues
to be satisfied; therefore, no change in a margin of safety occurs.
B. Revision to Unit 2 Overtemperature N-16 Reactor Trip Setpoints
Because the reactor core safety limits for CPSES Unit 2, Cycle 5
are recalculated, the Reactor Trip System instrumentation setpoint
values for the Overtemperature N-16 reactor trip setpoint which protect
the reactor core safety limits must also be recalculated. The
Overtemperature N-16 reactor trip setpoint helps prevent the core and
Reactor Coolant System from exceeding their safety limits during normal
operation and design basis anticipated operational occurrences. The
most relevant design basis analysis in Chapter 15 of the CPSES Final
Safety Analysis Report (FSAR) which is affected by the Overtemperature
reactor trip setpoint is the Uncontrolled Rod Cluster Control Assembly
Bank Withdrawal at Power (FSAR Section 15.4.2). This event has been
analyzed with the new safety analysis value for the Overtemperature
reactor trip setpoint to demonstrate compliance with event specific
acceptance criteria. Because all event acceptance criteria are
satisfied, there is no degradation in a margin of safety.
The nominal Reactor Trip System instrumentation setpoint values for
the Overtemperature N-16 reactor trip setpoint (Technical Specification
Table 2.2-1) are determined based on a statistical combination of all
of the uncertainties in the channels to arrive at a total uncertainty.
The total uncertainty plus additional margin is applied in a
conservative direction to the safety analysis trip setpoint value to
arrive at the nominal and allowable values presented in Technical
Specification Table 2.2-1. Meeting the requirements of Technical
Specification Table 2.2-1 assures that the Overtemperature reactor trip
setpoint assumed in the safety analyses remains valid. The CPSES Unit
2, Cycle 5 Overtemperature reactor trip setpoint is not significantly
different from the previous cycle, and thus provides operational
flexibility to withstand mild transients without initiating automatic
protective actions. Although the value of the f(delta I) trip reset
function setpoint is different, the Reactor Trip System instrumentation
setpoint values for the Overtemperature N-16 reactor trip setpoint are
consistent with the safety analysis assumptions which have been
analytically demonstrated to be adequate to meet the applicable event
acceptance criteria. Thus, there is no reduction in a margin of safety.
Using the NRC approved TU Electric methods, the reactor core safety
limits are determined such that all applicable limits of the safety
analyses are met. Because the applicable event acceptance criteria
continue to be met, there is no significant reduction in the margin of
safety.
C. Administrative changes to reflect plant nomenclature
Changes to the N-16 trip setpoint equation are for clarification
only to more accurately reflect CPSES plant nomenclature. This change
is administrative in nature and has no impact on 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 Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, N.W., Washington, DC 20036.
NRC Project Director: John N. Hannon.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: December 10, 1998.
[[Page 71976]]
Description of amendment request: The licensee proposed to correct
an error in the technical specifications by changing to the use of
``hydrogen, balance air'' rather than the incorrect ``hydrogen balance
nitrogen'' for calibration of the Augmented Offgass System hydrogen
monitors.
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:
Based on the criteria for defining a significant hazards
consideration in 10CFR50.92, operation of VYNPS in accordance with this
change would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:
The proposed change is purely administrative in nature--correcting
instrument calibration requirements to conform the Technical
Specification with the instrument manufacturer's recommendations. The
change has no effect on plant hardware, plant design, safety limit
setting, or plant system operation and therefore does not modify or add
any initiating parameters that would significantly increase the
probability or consequences of an accident previously evaluated. This
change to the Technical Specifications is a correction of an error
which occurred when the particular Technical Specification was issued.
The function of this surveillance requirement remains unchanged.
No new modes of operation are introduced by the proposed change
such that adverse consequences would result. Accordingly, the
consequences of previously analyzed accidents are not affected by this
proposed change.
The Augmented Off-Gas (AOG) System hydrogen monitors do not serve a
reactor safety function. In this context, the determination of no
significant hazards consideration defined in 10CFR50.92 is made based
on the ``accident previously evaluated'' being a postulated hydrogen
detonation within the off-gas system downstream of the hydrogen
recombiners. The hydrogen monitors do not mitigate the consequences of
an accident, but rather function to preclude a hydrogen explosion
within the off-gas system. The function of the Augmented Off-Gas System
hydrogen monitors to prevent a hydrogen detonation is not affected by
this change.
(2) Create the possibility of a new or different kind of accident
from any accident previously evaluated, because:
Since this change merely corrects Technical Specification wording
to reflect the actual manufacturer's recommended gas mixture to be used
for calibrating these instruments, no new or different types of
accidents are created. Since the calibration gas mixture has a very low
(approximately 2%) hydrogen concentration, its use does not introduce
the possibility of fires, explosions, or other hazards which might
adversely affect safety-related equipment. Therefore, use of the proper
calibration gas does not create the possibility of a new or different
kind of accident.
This change does not affect the operation of any systems or
components, nor does it involve any potential initiating events that
would create any new or different kind of accident. Therefore, the
proposed change does not create the possibility of a new or different
kind of accident from any previously evaluated for the Vermont Yankee
Nuclear Power Station.
(3) Involve a significant reduction in a margin of safety, because:
This proposed change involving the specification of the correct
calibration gas mixture ensures that the off-gas system hydrogen
monitors are properly calibrated and therefore preserve the margin of
safety in precluding a hydrogen explosion in the off-gas system.
Administratively changing this specification only establishes the
appropriate calibration gas for the actual, installed hydrogen
monitors. Changing the specification to reflect correct practice will
not reduce the margin of safety.
The proposed change does not affect any equipment involved in
potential initiating events or safety limits. Therefore, it is
concluded that 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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Project Director: Cecil O. Thomas.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: July 30, 1998 (TSCR 206).
Description of amendment request: The purpose of the proposed
amendments is to incorporate changes to the Technical Specifications to
more clearly define the requirements for Service Water (SW) System
operability.
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 Point Beach Nuclear Plant in accordance with
the proposed amendment[s] does not result in a significant increase in
the probability or consequences of any accident previously evaluated.
The Service Water System is primarily a support system for systems
required to be operable for accident mitigation. Portions of the SW
system supplying the containment fan coolers also function as part of
the containment pressure boundary under post accident conditions.
Failures within the SW system are not an initiating condition for any
analyzed accident.
Analyses performed demonstrate that under the Technical
Specifications allowable configurations, the SW system will continue to
perform all required functions. The SW system is capable of supplying
the required cooling water flow to systems required for accident
mitigation. That is, the SW system removes the required heat from the
containment fan coolers and residual heat removal heat exchangers
ensuring containment pressure and temperature profiles following an
accident are as evaluated in the FSAR [final safety analysis report].
This in turn ensures that environmental qualification of equipment
inside containment is maintained and thus function as required post-
accident.
SW system response post accident is within all design limits for
the system. Transient and steady state forces within the system remain
within all design and operability limits thereby maintaining the
integrity of the system inside containment and the integrity of the
containment pressure boundary. Assumptions dependent on containment
pressure profile for containment leakage assumed in the radiological
consequence analyses remain valid.
In addition, removing required heat from containment ensures that
cooling
[[Page 71977]]
of the reactor core is accomplished for long-term accident mitigation.
Therefore, operation of the SW system as proposed will not result
in a significant increase in the probability or consequences of any
accident previously evaluated.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind of
accident from any accident previously evaluated.
The proposed changes do not alter the way in which the SW system
performs its design functions nor the design limits of the system. The
proposed changes do not introduce any new or different normal operation
or accident mitigation functions for the system. Therefore, no new
accident initiators are introduced by the proposed changes. Operation
of SW system as proposed cannot result in a new or different kind of
accident from any accident previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction in a
margin of safety.
Analyses performed in support of the proposed amendments
demonstrate that the SW system continues to perform its function as
assumed and credited in the accident analyses and radiological
consequence analyses performed for the Point Beach Nuclear Plant.
Therefore, the analyses and results are not changed. All analysis
limits remain met. The SW system continues to be operated and responds
within all design limits for the system. Therefore, operation of the
Point Beach Nuclear Plant in accordance with the proposed amendments
cannot result in a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: September 23, 1998 (TSCR 209).
Description of amendment request: The purpose of the proposed
amendments is to remove the test requirements for snubbers from the
Technical Specifications (TS). These requirements are already included
in the Point Beach Nuclear Plant In-Service Inspection Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant increase in
the probability or consequences of an accident previously evaluated.
These changes do not involve a significant increase in the
probability of an accident previously evaluated because no such
accidents are affected by the proposed revisions to delete TS 15.4.3.
The proposed TS change does not introduce any new accident initiators.
Initiating conditions and assumptions are unchanged and remain as
previously analyzed for accidents in the PBNP Final Safety Analysis
Report. The proposed TS change does not involve any physical changes to
systems or components, nor does it alter the typical manner in which
the systems or components are operated. Therefore, these changes do not
increase the probability of previously evaluated accidents.
As noted above, the snubber testing requirements included in the
ASME/ANSI OM-4 Code are more comprehensive and in general more
conservative than the snubber testing requirements currently contained
in TS 15.4.13.
These changes do not involve a significant increase in the
consequences of an accident or event previously evaluated because the
source term, containment isolation or radiological releases are not
being changed by these proposed revisions. The snubber program ensures
that snubbers function as required, therefore related systems continue
to function as designed and analyzed. Existing system and component
redundancy and operation is not being changed by these proposed
changes. The assumptions used in evaluating the radiological
consequences in the PBNP Final Safety Analysis Report are not
invalidated. Therefore, these changes do not affect the consequences of
previously evaluated accidents.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
These changes do not introduce nor increase the number of failure
mechanisms of a new or different type than those previously evaluated
since there are no physical changes being made to the facility. As
noted above, the snubber testing requirements included in the ASME code
in general are more comprehensive than the snubber testing requirements
currently contained in TS 15.4.13 and provide the requisite level of
assurance of snubber operability. The design and design basis of the
facility remain unchanged. The plant safety analyses remain unchanged.
Therefore, the possibility of a new or different kind of accident from
any accident previously evaluated is not introduced.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not involve a significant reduction in a
margin of safety.
The proposed changes do not involve a significant reduction in the
margin of safety because existing component redundancy is not being
changed by these proposed changes. There are no changes to the initial
conditions contributing to accident severity or consequences, and
safety margins established through the design and facility license
including the Technical Specifications remain unchanged. Therefore,
there are no significant reductions in a margin of safety introduced by
[these] proposed amendment[s].
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: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
[[Page 71978]]
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: October 7, 1998 (TSCR 207).
Description of amendment request: The purpose of the proposed
amendments is to incorporate changes to the Technical Specifications
(TS) to ensure the 4 kV bus undervoltage input to reactor trip is
controlled in accordance with the design and licensing basis for the
facility. One additional administrative change is requested which
removes the footnote related to the definition of Rated Power in TS
15.1.j.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant [PBNP] in accordance
with the proposed amendments will not create a significant increase in
the probability or consequences of an accident previously evaluated.
The changes proposed ensure the Point Beach Nuclear Plant continues
to be operated in accordance with the design and licensing basis for
the facility.
The first change removes a footnote qualifying the definition of
Rated Power as applied to PBNP Unit 2. This restriction was eliminated
with the replacement of Unit 2 steam generators as approved by
Amendments 173 and 177, dated July 1, 1997. The analyses for those
amendments were performed based on the minimum flow requirements
specified in Technical Specification 15.3.1.G.3. The note should have
been deleted from the Technical Specifications at that time.
Elimination of this note does not result in a change in the operation
of PBNP from that analyzed and approved in Amendments 173 and 177.
Therefore, this change is administrative and cannot result in an
increase in probability or consequences of an accident previously
evaluated.
The second change modifies the Limiting Condition For Operation
[LCO] for the undervoltage reactor trip protection function. This trip
function is the primary protective function credited in the complete
loss of flow event analysis in the Final Safety Analysis Report (FSAR)
Section 14.1.8. As a primary protective function, this trip is required
to be single failure proof as stipulated in proposed IEEE 279-1968
documented in FSAR Section 7.2. This change ensures that this
protective feature is maintained in a condition where single failure
considerations are satisfied. When single failure criteria cannot be
met, appropriate action is stipulated to shutdown the unit placing it
in a condition where the protective function is no longer required.
Therefore, this change ensures PBNP is operated in accordance with its
design and licensing basis and cannot result in an increase in the
probability or consequences of an accident previously evaluated.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The changes proposed by this request remove a footnote qualifying
the definition of rated power as it applies to PBNP Unit 2 operation,
and modify the LCO related to the undervoltage reactor trip protective
function to ensure this function is maintained as required by the PBNP
design and licensing basis. These changes are in agreement with
approved analyses. These changes do not introduce any new accident
initiators or alter the response of the PBNP Units to previously
analyzed accidents. Therefore, operation of PBNP in accordance with the
proposed changes cannot result in a new or different kind of accident
from any accident previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not create a significant reduction in a
margin of safety.
Operation of the PBNP in accordance with the proposed amendments is
within the bounds of approved design and licensing basis of the
facility. The design and licensing basis establish appropriate margins
of safety. Since operation of the PBNP remains within the approved
design and licensing basis of the facility, a reduction in a margin of
safety cannot result.
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: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: November 18, 1998
Description of amendment request: The proposed amendment would
revise the pressure/temperature (P/T) limits and the low-temperature
overpressure protection (LTOP) requirements in the facility technical
specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change was reviewed in accordance with the provisions
of 10 CFR 50.92 to show no significant hazards exist. The proposed
change will not:
(1) Involve a significant increase in the probability or
consequence of an accident previously evaluated.
Failure of a reactor vessel is not an accident that has been
previously evaluated; design provisions ensure that this is not a
credible event. Since the potential consequences of a reactor vessel
failure are so severe, industry and governmental agencies have worked
together to ensure that failure will not occur. Compliance with 10 CFR
50 Appendix G and H ensures that failure of a reactor vessel will not
occur. The proposed changes do not impact the capability of the reactor
coolant pressure boundary piping (i.e., no change in operating
pressure, materials, seismic loading, etc.) and therefore do not
increase the potential for the occurrence of a LOCA [loss-off-coolant
accident].
The LTOP setpoint, revised enabling temperature, and revised P/T
limits reflected in proposed Figures TS 3.1-1 and TS 3.1-2 ensure that
the Appendix G pressure/temperature limits are not exceeded, and
therefore, ensure that RCS integrity is maintained. The changes do not
modify the reactor coolant system pressure boundary, nor make any
physical changes to the facility design, material, construction
standards, or setpoints. The reactor coolant system full power
operating pressure (2235 psig) is not being changed by this proposed
amendment. The LTOP valve setpoint remains at less than or equal to 500
psig. The LTOP enabling temperature based on Figure
[[Page 71979]]
TS 3.1-2 is 200 deg.F and is consistent with ASME Code Case N-514
guidance of RTNDT + 50 deg.F. The revised enabling
temperature is lower than the 355 deg.F value in the current TS.
However, the allowable combination of Appendix G pressures and
temperatures (refer to the 0 deg.F isothermal cooldown limit) is
greater for the revised limit curves. The combination of greater
allowable Appendix G pressure and temperature limits and lower enabling
temperature produces a larger operating window. A larger operating
window reduces the likelihood of inadvertently lifting the LTOP relief
valve while maneuvering the plant through the knee of the P-T curve
during startup and shutdown. The probability of an LTOP event occurring
is independent of the pressure-temperature limits for the RCS [reactor
coolant system] pressure boundary and enabling temperature. Therefore,
the probability of a[n] LTOP event is not increased.
The revised heatup and cooldown limit curves and LTOP enabling
temperature were developed using test results from unirradiated and/or
irradiated specimens that represent the KNPP [Kewaunee Nuclear Power
Plant] reactor vessel beltline circumferential weld, closure head
flange, and intermediate forging. The circumferential beltline weld and
intermediate forging are the most limiting materials in the reactor
coolant pressure boundary due to the effects of neutron irradiation
which cause the flow properties to increase and the toughness to
decrease. 10 CFR 50, Appendix G states that the metal temperature of
the closure flange regions must exceed the material unirradiated
RTNDT by at least 120 deg.F for normal operation and
90 deg.F for hydrostatic pressure tests and leak tests when the
pressure exceeds 20 percent of the preservice hydrostatic test
pressure. Drop weight and Charpy V-notch testing of IP3571 weld metal
and the intermediate forging material has been performed and used for
derivation of the revised PTS [pressurized thermal shock] assessment,
the proposed Appendix G heatup and cooldown limit curves, and the
corresponding LTOP system enabling temperature. The revised limit
curves and corresponding LTOP enabling temperature have been developed
using accepted engineering practices, methods derived from the ASME
Boiler and Pressure Vessel Code, criteria set forth in NRC Regulatory
Standard Review Plan 5.3.2, and 10 CFR 50.61. Utilization of the
revised heatup and cooldown limit curves and corresponding LTOP
enabling temperature ensures adequate fracture toughness for ferritic
materials of the pressure-retaining components of the reactor coolant
pressure boundary. These limit curves provide adequate margins of
safety during any condition of normal operation, including anticipated
operational occurrences and system hydrostatic tests, and low
temperature overpressure protection (corresponding to isothermal events
during low temperature operations (i.e., less than or equal to
200 deg.F)) thus ensuring the integrity of the reactor coolant pressure
boundary.
The changes do not adversely affect the integrity of the RCS such
that its function in the control of radiological consequences is
affected. Radiological off-site exposures from normal operation and
operational transients, and faults of moderate frequency do not exceed
the guidelines of 10 CFR 100. In addition, the changes do not affect
any fission product barrier. The changes do not degrade or prevent the
response of the LTOP relief valve or other safety-related systems to
previously evaluated accidents. In addition, the changes do not alter
any assumption previously made in the radiological consequence
evaluations nor affect the mitigation of the radiological consequences
of an accident previously evaluated. Therefore, the consequences of an
accident previously evaluated will not be increased.
Thus, operation of KNPP in accordance with the PA does not involve
a significant increase in the probability or consequences of any
accident previously evaluated.
(2) Create the possibility of a new or different kind of accident
from any previously evaluated.
Since the potential consequences of a reactor vessel failure are so
severe, industry and governmental agencies have worked together to
ensure that failure will not occur. Compliance with 10 CFR 50 Appendix
G and H ensures that failure of a reactor vessel will not occur. The
proposed heatup and cooldown limit curves have been constructed by
combining the most conservative pressure-temperature limits derived by
using material properties of the intermediate forging, closure head
flange, and beltline circumferential weld to form a single set of
composite curves. With NRC approval to use Code Case N-588, the
intermediate forging and closure head flange become the controlling
materials for development of the heatup limit curve and the cooldown
limit curves at low temperatures. At high temperatures, the
circumferential weld continues to be limiting for development of the
cooldown limit curves. Use of conservative pressure-temperature limits
derived by using material properties of the intermediate forging,
closure head flange, and beltline circumferential weld to form a single
set of composite curves, does not modify the reactor coolant system
pressure boundary, nor make any physical changes to the LTOP setpoint
or design. Proposed Figures TS 3.1-1 and TS 3.1-2 were prepared in
accordance with regulatory and code requirements and were derived using
more conservative material property basis and more limiting
requirements of neutron exposure projections thru 33 EFPY [effective
full-power years] instead of 20 EFPY.
The revised LTOP system enabling temperature and the proposed
Appendix G pressure temperature limitations were prepared using methods
derived from the ASME Boiler and Pressure Vessel Code and the criteria
set forth in NRC Regulatory Standard Review Plan 5.3.2. The changes do
not cause the initiation of any accident nor create any new credible
limiting failure for safety-related systems and components. The changes
do not result in any event previously deemed incredible being made
credible. As such, it does not create the possibility of an accident
different than previously evaluated.
The changes do not have any adverse effect on the ability of the
safety-related systems to perform their intended safety functions. The
combination of higher allowable Appendix G pressure and temperature
limits and lower enabling temperature produces a larger operating
window. The ASME Section XI, Working Group on Operating Plant Criteria
(WGOPC) has prepared a technical bases document for Code Case N-514.
The technical bases document is contained in Attachment 3 of Reference
1. This technical bases document provides justification for enabling
the LTOP system at temperatures less than 200 deg.F or at coolant
temperatures corresponding to a reactor vessel metal temperature less
than RTNDT + 50 deg.F, whichever is greater.
WGOPC, which has responsibility for Appendix G of Section XI, has
considered the burden and safety impact imposed by the LTOP criteria,
and has developed Code guidelines for determining the LTOP set-point
pressure and the required enabling temperature. These guidelines will
relieve some operational restrictions, yet provide adequate margins
against failure for the reactor vessel. Further, by relieving the
operational restrictions, these guidelines result in a reduced
[[Page 71980]]
potential for activation of pressure relieving devices, thereby
improving plant safety. Thus, a slightly larger operating window at
KNPP is viewed to reduce the likelihood of inadvertently lifting the
LTOP relief valve while maneuvering the plant through the knee of the
P-T curve during startup and shutdown. The new LTOP operating window
(i.e., less than or equal to 200 deg.F) is within the existing
operating band for the residual heat removal system; operating
procedures allow the LTOP system to be placed into service at
<400 deg.F. At KNPP, as long as the LTOP relief valve is operable, the
LTOP system is enabled anytime the residual heat removal system is in
communication with the reactor coolant system.
The proposed changes do not make physical changes to the plant or
create new failure modes. Thus, the PA does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) Involve a significant reduction in the margin of safety.
The proposed Appendix G pressure temperature limitations and LTOP
enabling temperature were prepared using methods derived from the ASME
Boiler and Pressure Vessel Code, including Code Cases N-514 and N-588,
and the criteria set forth in NRC Regulatory Standard Review Plan
5.3.2. Reference 1 to this letter provides information to support NRC
approval to use Code Case N-514 and Code Case N-588 for the KNPP PTS
evaluation, development of the heatup and cooldown limit curves, and
establishment of the LTOP system enabling temperature. These documents
and practices along with the calculational limitations specified in 10
CFR 50.61 are an acceptable method for implementing the requirements of
10 CFR 50 Appendices G and H.
Use of the methodology set forth in the ASME Boiler and Pressure
Vessel Code, NRC Regulatory Standard Review Plan 5.3.2., 10 CFR 50.61,
and 10 CFR 50 Appendices G and H ensures that proper limits and safety
factors are maintained. Thus, the PA does not involve a significant
reduction in the margin of safety.
The revised heatup and cooldown limit curves and LTOP system
enabling temperature were prepared using drop weight and Charpy V-notch
data for the beltline weld, closure head flange, and intermediated
forging material along with practices described herein and methods
derived from the ASME Boiler and Pressure Vessel Code and 10 CFR 50.61.
The safety factors and margins used in the development of the limit
curves and LTOP system enabling temperature meet the criteria set forth
by these documents. Application of low leakage core designs decreases
the rate of shift in transition temperature from ductile to nonductile
behavior. The revised limit curves and LTOP enabling temperature
provide adequate margins of safety during any condition of normal
operation, including anticipated operational occurrences and system
hydrostatic tests, and low temperature overpressure protection
(corresponding to isothermal events during low temperature operations
(i.e., less than or equal to 200 deg.F)). With the preparation of the
revised limit curves in accordance with the latest criteria and
guidance, this PA ensures that proper limits and safety factors are
maintained.
Thus, the PA 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: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Project Director: Cynthia A. Carpenter.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: May 15, 1998, as supplemented
September 25 and October 13, 1998.
Brief description of amendment: The amendment would revise
Technical Specification 5.5, ``Storage of Unirradiated and Spent Fuel''
to reflect a planned modification to increase the number of fuel
assemblies that can be stored in the spent fuel pool from 2776 to 4086.
Date of publication of individual notice in Federal Register:
November 24, 1998 (63 FR 64973).
Expiration date of individual notice: December 24, 1998.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Notice of of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3)
[[Page 71981]]
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.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of application for amendments: October 16, 1998.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.3.1 ``Reactor Protective System (RPS)
Instrumentation-Operating'' and TS 3.3.2, ``Reactor Protective System
(RPS) Instrumentation-Shutdown,'' to clarify an inconsistency between
the TS wording and the design bases as described in the TS Bases and
the Updated Final Safety Analysis Report. Specifically, the change
replaces the operating bypass input process variable, Thermal Power, in
Footnotes (a), (b), and (d) of Table 3.3.1 and in the Note to Limiting
Condition for Operation 3.3.2 with Nuclear Instrument Power.
Date of issuance: December 8, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 229 & 204.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1998 (63 FR
57320).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated December 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: April 25, 1996, as supplemented
on September 5, 1996, August 8, 1997, March 26, July 31, and August 24,
1998.
Brief description of amendment: This amendment revises Technical
Specifications (TSs) 3/4.5.F.1, ``Core and Containment Cooling
systems'' to extend the allowed outage time (AOT) for the emergency
diesels, TSs 3.9.B.1 and 3.9.B.4, ``Auxiliary Electrical System'' to
reduce the AOT from 7 days to 3 days and reduce the AOT for the
combination of an EDG and startup transformer or shutdown transformer
from 72 hours to 48 hours, and add Configuration Risk Management
Program in TS 5.5, ``Programs and Manuals'' of Section 5.0
``Administrative Controls''. Various TS pages were re-numbered in
Section 5.0. In addition, TSs 3.9, ``Auxiliary Electrical System,'' and
3.9.A, ``Auxiliary Electrical Equipment,'' have been reformatted to be
consistent with TS 3.9.B approved in a previous amendment. The
associated Bases sections have also been changed to reflect the new
TSs.
Date of issuance: December 11, 1998.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 179.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50934).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 11, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: July 15, 1997, as supplemented
March 3, April 13, June 16, October 26, and November 5, 1998.
Brief description of amendments: The amendments revised the
Technical Specifications to add new requirements for the main steamline
break instrumentation and resolved issues related to Inspection and
Enforcement Bulletin 80-04.
Date of Issuance: December 7, 1998.
Effective date: As of the date of issuance to be implemented
coincident with implementation of the improved Technical
Specifications.
Amendment Nos.: 234--Unit 1; 234--Unit 2; 233--Unit 3.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 24, 1997 (62
FR 50001).
The March 3, April 13, June 16, October 26, and November 5, 1998,
letters provided clarifying information that did not change the scope
of the July 15, 1997, application and the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 7, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit 2, Shippingport, Pennsylvania
Date of application for amendment: September 24, 1998, as
supplemented November 3, 1998.
Brief description of amendment: This amendment revised technical
specification 3.1.2.8 in two places to change the term ``contained
volume'' to usable volume.'' This change eliminates the potential for a
non-conservative interpretation of the specification values for the
Refueling Water Storage Tank and Boric Acid Storage Tank and thereby
eliminates the need for temporary administrative controls, which have
been used correctly to properly interpret the specification values as
usable volumes.
Date of issuance: December 14, 1998.
Effective date: Effective immediately, to be implemented within 30
days.
Amendment No: 95.
Facility Operating License No. NPF-73. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59591).
The November 3, 1998, letter did not change the initial proposed no
significant hazards consideration determination or expand the amendment
request beyond the scope of the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 14, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit
1, DeWitt County, Illinois
Date of application for amendment: August 17, 1998.
[[Page 71982]]
Brief description of amendment: The amendment reduces the load at
which diesel generators are tested.
Date of issuance: December 14, 1998.
Effective date: December 14, 1998.
Amendment No.: 118.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53949).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 14, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, IL 61727.
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 application for amendments: August 1, 1997.
Brief description of amendments: The amendments delete a portion of
a technical specifications surveillance test requirement that specifies
that the steam driven auxiliary feedwater pumps be tested ``when the
secondary steam supply pressure is greater than 310 psig.'' This
removes any misunderstanding that the secondary steam pressure must be
just above 310 psig for this test.
Date of issuance: December 10, 1998.
Effective date: December 10, 1998, with full implementation within
45 days.
Amendment Nos.: 225 and 209.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68308).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 10, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station Unit No. 2, Oswego County, New York
Date of application for amendment: February 5, 1998.
Brief description of amendment: This amendment changes the
Technical Specifications to update the terminology and references to 10
CFR 50.55a(f) and (g) consistent with the 1989 edition of Section XI of
the American Society of Mechanical Engineers Boiler and Pressure Vessel
Code, and consistent with the second 10-year interval of the Inservice
Inspections and Inservice Testing Program Plans.
Date of issuance: December 3, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 84
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 11, 1998 (63 FR
11920).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 3, 1998.
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.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: August 8, 1996, as supplemented
June 30, 1997 and August 26, 1998.
Brief description of amendments: The amendments eliminate the
response time testing requirements for selected sensors and specified
instrument loops for (1) the reactor protection system, (2) the
isolation system, and (3) the emergency core cooling system.
Date of issuance: December 14, 1998.
Effective date: Both units, as of date of issuance, to be
implemented within 30 days.
Amendment Nos.: 132 and 93.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 6, 1996 (61 FR
57489).
The June 30, 1997 and August 26, 1998, letters 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 14, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: July 10, 1998, as supplemented
October 16, 1998.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.6/4.6 and associated bases to relocate portions of
the reactor coolant chemistry to the Updated Final Safety Analysis
Report and to applicable plant procedures. Changes to the relocated
requirements will be controlled by the provisions of 10 CFR 50.59.
Date of issuance: December 1, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 247.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40560).
The October 16, 1998, submittal fell with the scope of, and did not
change, the initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 1, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: March 30, 1998, as supplemented
on October 27, 1998.
Brief description of amendment: The amendment revises the
definition of logic system functional tests, and revises test frequency
requirements for certain instrumentation.
Date of issuance: December 11, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 248.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19978).
The October 27, 1998, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration.
[[Page 71983]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 11, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: August 12, 1998, as
supplemented on October 12, 1998. The October 12, 1998, letter provided
clarifying information that did not change the initial proposed no
sigificant hazards consideration determination.
Brief description of amendments: The amendments revise TS 3/
4.6.1.3, ``Containment Air Locks,'' to change the action statements for
an inoperable air lock. The amendments also revise TS Bases 3/4.6.1.2,
``Containment Leakage,'' to correct an editorial error and TS Bases 3/
4.6.1.3, ``Containment Air Locks,'' to provide additional details
regarding the air locks.
Date of issuance: December 2, 1998.
Effective date: December 2, 1998.
Amendment Nos: 215 and 195.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48265).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 2, 1998
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: December 31, 1997, as supplemented by
letter dated September 11, 1998.
Brief Description of amendments: The amendments revised the
Technical Specifications (TSs) to change the intermediate range neutron
flux reactor trip setpoint and allowable value, and delete the
reference to the reactor trip setpoints in TS 3.10.3, ``Special Test
Exceptions--Physics Tests,'' and TS 3.10.4, ``Special Test Exceptions--
Reactor Coolant Loops.''
Date of issuance: December 8, 1998.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1--140; Unit 2--132.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: February 11, 1998 (63
FR 6998).
The September 11, 1998, letter provided clarifying information that
did not change December 31, 1997, application or 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, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 20, 1996 (TS 96-09).
Brief description of amendments: The amendments change the
Technical Specifications to clarify the types of work shifts that are
acceptable when considering the requirements to ensure overtime is not
heavily used on a routine basis by unit staff.
Date of issuance: December 7, 1998.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: 240 and 230.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59596).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 7, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 22, 1998, as
supplemented on August 27 and October 8, 1998 (TS 96-08). The August
27, 1998, amendment request superseded the original (August 22, 1998)
request in its entirety.
Brief description of amendments: The amendments revise the Sequoyah
Nuclear Plant Technical Specifications by extending the allowed outage
time for the SQN emergency diesel generators from 72 hours to 7 days.
Date of issuance: December 16, 1998.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: 241 and 231.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52969), superseded by a second notice on September 9, 1998 (63 FR
48270). The October 8, 1998, letter provided clarifying information
that did not change the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: July 28, 1998, as supplemented
October 16, 1998. The October 16, 1998, letter was administrative in
nature and did not change the initial no significant hazards
consideration determination.
Brief description of amendments: The amendments revise the
Technical Specifications to change the Emergency Diesel Generator
section to be consistent with station procedures associated with
steady-state conditions.
Date of issuance: December 10, 1998.
Effective date: December 10, 1998.
Amendment Nos.: 216 and 197.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48272).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 10, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
[[Page 71984]]
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Dated at Rockville, Maryland, this 23rd day of December 1998.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-34440 Filed 12-29-98; 8:45 am]
BILLING CODE 7590-01-P