[Federal Register Volume 63, Number 241 (Wednesday, December 16, 1998)]
[Notices]
[Pages 69332-69353]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-33206]
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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. Pub. L. 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 20, 1998, through December 4, 1998.
The last biweekly notice was published on December 2, 1998 (63 FR
66590).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of 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 15, 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.
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Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
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 amendments request: November 19, 1998.
Description of amendments request: The proposed amendment revises
Technical Specification 3.7.6, ``Service Water (SRW) System'' to allow
operation of Calvert Cliffs with one SRW plate and frame heat exchanger
(PHE) secured for maintenance or other reasons, and removing one
containment air cooler (CAC) from service to enable the affected
subsystem to remain operable. Specifically, the proposed change adds
``One SRW heat exchanger inoperable'' as a new condition for Limiting
Condition for Operation (LCO) 3.7.6. The required actions for the new
condition are to secure one CAC within one hour and restore the heat
exchanger to operable condition within 7 days, or be in Mode 3 in 6
hours and Mode 5 in 36 hours. This limits the effect of one inoperable
PHE to only one containment cooling train made inoperable by the PHE.
Consequently, the new action statement introduced in the SRW LCO for an
inoperable PHE is similar to the one that already exists in the CAC LCO
for one inoperable containment cooling train.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
None of the systems associated with the proposed revision to the
Calvert Cliffs Technical Specifications are accident initiators. The
Saltwater (SW) and SRW systems are used to mitigate the effects of
accidents analyzed in the Updated Final Safety Analysis Report
(UFSAR). The SW and SRW Systems provide cooling to safety-related
equipment following an accident. The CACs are provided with SRW to
remove heat from the Containment in the event of an accident. They
support accident mitigation functions; therefore, the proposed
modification does not increase the probability of an accident
previously evaluated.
The proposed revision will provide greater availability of
safety-related equipment during PHE maintenance activities. It
ensures that the safety features provided by the SW and SRW, except
for the isolated CAC, are maintained, i.e., the availability of
safety-related equipment required to mitigate the radiological
consequences of an accident described in the UFSAR is enhanced by
the flexibility provided by this Technical Specification revision.
Furthermore, the proposed revision will not change, degrade, or
prevent actions described or assumed in any accident described in
the UFSAR. The proposed activity will not alter any assumptions
previously made in evaluating the radiological consequences of any
accident described in the UFSAR.
Therefore, the proposed modification does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
None of the systems associated with this modification are
identified as accident initiators in the UFSAR. The SW and SRW
Systems and the CACs are used to mitigate the effects of accidents
analyzed in the UFSAR. None of these functions required of these
systems have been changed by the proposed revision to the Technical
Specifications. This activity does not modify any system, structure,
or component such that it could become accident initiator, as
opposed to its current role as an accident mitigator.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The safety design basis for the SW and SRW Systems is the
availability of sufficient cooling capacity to ensure continued
operation of equipment during normal and accident conditions. The
redundant cooling capacity of these systems, assuming a single
failure, is consistent with assumptions used in the accident
analysis.
With one SRW subsystem inoperable, the remaining SRW subsystem
is adequate to perform the heat removal function. However, the
reliability is reduced because a single failure in the operable SRW
subsystem could result in loss of SRW function. The proposed change
will allow continued operation of some SRW-cooled components while a
PHE is being out-of-service. The second SRW subsystem will still be
available to perform the SRW function. In addition, the reliability
of many diesel generator-backed components will be improved since
the second diesel generator will remain operable while in this
action statement.
During a design basis accident, a minimum of one containment
cooling train (two of the four CACs) and one containment spray
train, is required to maintain the containment peak pressure and
temperature, below the design limits. Under the existing Technical
Specification requirement, with one containment cooling train
inoperable, the inoperable containment cooling train must be
returned to operable status within seven days. The remaining
operable containment spray and cooling units provide iodine removal
capabilities and are capable of removing at least 100% of the heat
removal needs after an accident. The seven-day completion time was
developed taking into account the redundant heat removal
capabilities afforded by combinations of the containment spray and
cooling systems, and the low probability of a design basis accident
occurring during this period. The proposed change to Technical
Specification 3.7.6 would allow three CACs to remain operable during
maintenance on a PHE, instead of the two that are maintained under
the current Technical Specification requirement.
[[Page 69334]]
For the above reasons, the margin of safety has been preserved,
and in some cases increased, by the proposed revision to the
Technical Specifications.
Therefore, this proposed modification does not significantly
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: S. Singh Bajwa, Director.
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 amendments request: November 20, 1998.
Description of amendments request: On September 9, 1996, a final
rule amending 10 CFR 50.55a was issued requiring owners to implement,
by September 9, 2001, the requirements of the 1992 Addenda of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code Section XI, Subsections IWE and IWL, as modified and supplemented
by 10 CFR 50.55a. Baltimore Gas and Electric Company (BGE) have
developed a program plan to effect the implementation of Subsection IWE
and IWL. BGE's submittal requests a license amendment in support of the
program plan. One Technical Specification (TS) change requested is an
administrative change that removes a TS originally developed from
Regulatory Guide (RG) 1.35. Compliance with RG 1.35 is not sufficient
to comply with 10 CFR 50.55a, as amended. The other TS changes request
the removal from the TSs requirements that are a duplication of 10 CFR
50.55a.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The Containment Building is a passive safety structure that
prevents the release of radioactive materials to the environment in
post-accident conditions. The proposed Technical Specification
changes delete requirements of the Technical Specifications that
have been made obsolete by the improvements of the Containment
Building inspections required by the changes in the regulations. The
improved inspections required by the American Society of Mechanical
Engineers Code serve to maintain Containment response to accident
conditions, by causing the identification and repair of defects in
the Containment Buildings.
Relocating existing requirements, eliminating requirements that
duplicate regulations, and making administrative improvements
provide Technical Specifications that are easier to use. Because
existing requirements are controlled by regulation, there is no
reduction in commitment and adequate control is still maintained.
Likewise, the elimination of requirements that duplicate regulations
enhances the usability of the Technical Specifications without
reducing commitments. Therefore, the proposed changes would not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The Containment Building is a passive safety structure designed
to contain radioactive materials released from the Reactor Coolant
System. The performance of the Containment Building is not evaluated
as the causal factor in any accident at Calvert Cliffs Nuclear Power
Plant. The proposed Technical Specification changes delete
requirements of the Technical Specifications that have been made
obsolete by the improvements of the Containment Building inspections
required by the changes in the regulations. Revising the Technical
Specifications, to comply with current regulations and to eliminate
duplication of requirements, does not create the possibility of a
new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The safety function of the Containment Building is to provide a
boundary to the release of radioactive material to the environment
during post-accident conditions. The changes to the Technical
Specifications incorporate improved inspection techniques and
criterial to ensure optimum Containment integrity and, therefore,
optimum containment response in the event of an accident resulting
in a release of radioactive material from the Reactor Coolant
System.
Optimizing containment integrity will result in maintaining the
margin of safety allowed by the Containment Buildings. Therefore,
the proposed changes 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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: S. Singh Bajwa, Director.
CBS Corporation acting through its Westinghouse Electric Company
Division (licensee), Westinghouse Test Reactor, Waltz Mill Site,
Westmoreland, Pennsylvania, Docket No. 50-22, License No. TR-2.
Date of amendment request: September 28, 1998, supplemented on
November 17, 1998.
Description of amendment request: CBS Corporation acting through
its Westinghouse Electric Company Division is the licensee for the
Westinghouse Test Reactor (WTR) at Waltz Mill, Pennsylvania. The
licensee is authorized to only possess the reactor and a
decommissioning plan has been approved. The licensee is planning to
sell most of its nuclear related facilities to other entities, but will
retain the WTR. One of the arrangements made with the purchasers of the
other facilities is that the Westinghouse name will be conveyed with
these facilities, and because of this arrangement, the licensee
requests that the license associated with the Westinghouse Test Reactor
be changed to simply CBS Corporation, to eliminate any reference to the
name Westinghouse.
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
considerations. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). A proposed amendment to a
license of a facility involves no significant hazards consideration if
operation of the facility in accordance with the proposed amendment
would not: (1) involve a significant increase in the probability or
consequences of an accident previously evaluated; or (2) create the
possibility of a new or different kind of accident from any accident
previously evaluated; or (3) involve a significant reduction in a
margin of safety.
The staff agrees with the licensee's no significant hazards
consideration determination submitted on November 17, 1998, for the
following reason.
This corporate name change does not involve any change in the
management, organization, location, facilities equipment, or procedures
related to the licensed activities under the WTR
[[Page 69335]]
license. The employees responsible for the licensed WTR facility will
still be responsible, either directly through the CBS Corporation or
through contractual arrangements for which CBS Corporation is
ultimately responsible, notwithstanding the new name of the licensee.
Based on a review of the licensee's analysis, and on the staff's
analysis detailed above, 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.
Attorney for license: Lisa A. Campagna, Assistant General Counsel,
Law Department, CBS Corporation, P.O. Box 355, Pittsburgh, Pennsylvania
15230.
NRC Project Director: Seymour H. Weiss.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos.
1 and 2, Will County, Illinois.
Date of amendment request: October 30, 1998.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) to reduce the spent fuel pool
(SFP) inadvertent draindown level to account for the effects of
potential failures of the SFP cooling and skimmer loops.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
This change to the TS does not involve an increase in the
probability of an accident previously evaluated. The initial
conditions of the limiting dewatering incidents involve initiating
circumstances/failures such as accidental gate openings, gate seal
failures, or an open transfer tube.
Specifying a revised inadvertent drain limit which meets the SRP
[Standard Review Plan, NUREG-0800] acceptance criteria is unrelated
to the probability of occurrence of the precursors or initiating
events. These initiators are not affected by the SFP cooling or
skimmer loop piping/component failure scenarios. There is no change
being made to the approved design, nor is there any operational
change being made which would increase the probability of
occurrence.
This change to the TS does not involve an increase in the
consequences of an accident previously evaluated. As documented in
NUREG-0876, Byron SER, Section 9.1.3, page 9-5, the anti-siphon
protection design of the SFP cooling and clean-up piping was
reviewed and found to be acceptable stating that ``all connections
to the spent-fuel pool are either near the normal water level or are
provided with antisiphon holes to preclude possible siphon draining
of the pool water.'' This review is applicable to Braidwood as
documented in NUREG-1002, Braidwood SER. The anti-siphon attributes
employed in the SFP skimmer loops at Braidwood, (under consideration
at Byron), are similar in design as well as their submergence levels
previously evaluated for the SFP cooling loops. The proposed change
revises the SFP inadvertent drain limit from approximately 423 feet
to 410 feet to bound the failure effects of both the SFP cooling and
skimmer loops, while considering any maloperation or failure
scenario. The revised value meets the SRP acceptance criteria of
maintaining at least 10 feet above the active fuel ensuring that
adequate radiation shielding is maintained as previously analyzed.
There is no physical or operational change being made which would
alter the sequence of events, plant response, or conclusions of the
affected analysis. There is no change in the type or amount of any
effluents released, and no change in either the Onsite or Offsite
dose consequences as a result of this change.
Therefore, based on this evaluation, this proposed amendment
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
This proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This change specifically identifies the SFP level sufficient to
ensure that the SRP acceptance criteria for inadvertent draining are
met while accounting for the failure effects of both the SFP cooling
and skimmer loops. Any inadvertent SFP draining due to potential
failures of the SFP skimmer loops is similar in nature to the
inadvertent SFP draining effects previously considered due to
failures of the SFP cooling loops. No new equipment is being
installed, and no installed equipment is being operated in a new or
different manner with this change. There is no change in plant
operation that affects previously evaluated failure modes. This
change does not represent a new failure mode or accident from what
has been previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The current TS value does not address inadvertent SFP draining
due to potential failures of the SFP skimmer loops or cooling
suction lines as was done for the SFP cooling discharge lines. This
change specifically identifies the SFP level sufficient to ensure
that the SRP acceptance criteria for inadvertent draining are met
while accounting for the failure effects of both the SFP cooling and
skimmer loops in determining the proposed TS value. The most
limiting postulated SFP dewatering incidents involve SFP drainage to
either a dry transfer canal, a dry transfer canal and cask fill
area, or a dry transfer canal and cask fill area which additionally
communicates through an open transfer tube to an empty refuel
cavity. The initial conditions of the dewatering incident analysis
and resultant water levels over the spent fuel are not affected by
this SFP skimmer/cooling loop issue because these incident
initiators are not effected by the SFP cooling or skimmer loop
failures, thus preserving the previously analyzed and approved
margin for these dewatering incidents.
For the less-limiting SFP skimmer/cooling loop failure issue,
the proposed TS change inadvertent drain limit meets the SRP minimum
requirement of at least 10 feet above the top of the active fuel
ensuring that adequate radiation shielding is maintained. This
change would allow for the conservative acceptance criteria for the
current UFSAR [Updated Final Safety Analysis Report] design analysis
to continue to be met.
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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Stuart A. Richards.
Commonwealth Edison Company, Docket No. 50-374, LaSalle County
Station, Unit 2, LaSalle County, Illinois.
Date of amendment request: November 9, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.3.2, ``Isolation Actuation
Instrumentation'' to add/revise various isolation setpoints for leak
detection instrumentation. These changes are necessary due to
modifications to the Reactor Water Cleanup (RWCU) System to restore
``hot'' suction to the RWCU pumps and due to a re-evaluation of the
high energy line break analysis. In addition, the amendment would
eliminate isolation actuation trip functions for the Residual Heat
Removal (RHR) system steam
[[Page 69336]]
condensing mode and shutdown cooling mode.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
(a) There is no effect on accident initiators so there is no
change in probability of an accident. A line break in the subject
areas, would consist of an instantaneous circumferential break
downstream of the outermost isolation valve of one of these systems.
The leak detection isolation is only a precursor of a break, and
thus does not affect the probability of a break.
(b) There is minimal effect on the consequences of analyzed
accidents due to changing the leak detection ambient temperature or
Delta T setpoint and allowable values to detect 25 gpm equivalent
leakage. The addition of more ambient temperature and T
leak detection monitoring, along with the addition of the high flow
break detection will actually decrease the consequences of the
associated accidents. The worst case accident outside the primary
containment boundary is a main steam line break which bounds the
dose consequences of all line breaks and therefore bounds any size
of leak.
The deletion of the RHR steam condensing mode isolation
actuation instrumentation trip functions from the LaSalle Technical
Specifications does not increase the probability or consequences of
an accident previously evaluated, because this mode of operation of
the RHR system has been deleted from the LaSalle design basis and
the lines that were previously high energy lines are isolated during
unit operation, including Operational Condition 1 (Run mode),
Operational Condition 2 (Startup mode), and Operational Condition 3
(Hot Shutdown).
The deletion of the RHR shutdown cooling mode leak detection T
and Delta T isolation actuation instrumentation trip functions from
the LaSalle Technical Specifications does not increase the
probability or consequences of an accident previously evaluated,
because the leak detection is only a precursor of a break, and thus
does not affect the probability of a break. Also, there are two
other methods of detecting abnormal leakage and isolating the system
in Technical Specification trip functions A.6.a, Reactor Vessel
Water Level--Low, Level 3 and A.6.c, RHR Pump Suction Flow--High. In
addition, other means to detect leakage from the RHR system, such as
sump monitoring and area radiation monitoring, are also available.
In accordance with Technical Specification Administrative
Requirement 6.2.F.1, LaSalle has a leakage reduction program to
reduce leakage from those portions of systems outside primary
containment that contain radioactive fluids. RHR, including piping
and components associated with the shutdown cooling mode, is part of
this program, which includes periodic visual inspection of the
system for leakage. The sump monitoring, radiation monitoring and
periodic inspections for system leakage makes the probability of a
leak of 5 gpm going undetected for more than a day very low.
Also, due to the low reactor pressures (less than 135 psig) at
which RHR shutdown cooling mode is able to operate, reactor coolant
makeup and outflow is very low compared to normal plant operation. A
change in flow balance due to a leak is thus more readily detectable
with reactor coolant water level changes and makeup flow rate, and
thus precludes a significant leak going undetected before break
detection instrumentation would cause automatic isolation.
Therefore, this proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The purpose of the leak detection system, as it applies to the
RWCU and RHR system areas, is to provide the capability for leak
detection and automatic isolation of the system as necessary in the
event of leakage in these areas. This change maintains this
capability with at least two different methods of detection of
abnormal leakage for protection from the flooding concerns of a
significant leak or line break when the RHR system is operating in
the shutdown cooling mode, so that redundant systems will not be
affected.
This change also maintains or adds primary containment isolation
logic for the leak detection isolation based on temperature
monitoring in RWCU areas and break detection based on RWCU pump
suction flow--high. The additional instrumentation and the
associated isolation logic is the same or similar to existing
instrumentation and logic for containment actuation instrumentation,
so no new failure modes are created in this way.
Therefore, these proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) Involve a significant reduction in the margin of safety
because:
The change to the automatic isolation setpoint for high Delta T
leak detection in the heat exchanger rooms is based on current
configuration calculated/analyzed response to a small leak compared
to a circumferential break. The increased leakage rate in the RWCU
heat exchanger rooms that is necessary to actuate isolation on
ambient temperature during winter conditions, does not adversely
affect the margin of safety. This increased leakage rate is below
the critical crack leakage rate as represented in UFSAR [Updated
Final Safety Analysis Report] Figure 5.2-11. Additionally,
differential temperature leak detection is conservative under these
same conditions, and will actuate isolation at a leakage rate less
than the established limit. The leak detection isolation logic is
unchanged and thus remains single failure proof.
The addition of automatic primary containment isolation on
ambient temperature and Delta T-High for the Reactor Water Cleanup
System (RWCU) Pump, Pump Valve, Holdup Pipe, and Filter/
Demineralizer (F/D) Valve Rooms and the addition of the RWCU Pump
Suction Flow High line break isolation add to the margin of safety
with respect to leak detection and line breaks in the RWCU system,
because the system isolation diversity is increased and the amount
of system piping monitored for leakage is increased.
The setpoints for the ambient temperature and Delta T leak
detection isolations being changed or added and the RWCU pump
suction flow--high are set sufficiently high enough so as not to
increase the possibility of spurious actuation. In the event that a
spurious actuation does occur, little safety significance is
presented since the RWCU system performs no safety function. The
setpoints and allowable values for the proposed changes also assure
sufficient margin to the analytical values and are high enough to
prevent spurious actuations based on calculations consistent with
Regulatory Guide 1.105.
The deletion of the RHR steam condensing mode isolation
actuation instrumentation does not effect the margin of safety,
because this mode is no longer utilized by LaSalle in Operational
Conditions 1, 2, or 3 (Run mode, Startup mode, or Hot Shutdown).
The elimination of the temperature based trip functions for the
RHR shutdown cooling mode area is based on the determination that
temperature is not the appropriate parameter for leak detection as
it does not provide meaningful indication and will not provide
setpoints that would be sufficiently above the normal range of
ambient conditions to avoid spurious isolations.
There are two other methods of detecting abnormal leakage and
isolating the system in Technical Specification trip function A.6,
which are A.6.a, Reactor Vessel Water Level--Low, Level 3 and A.6.c,
RHR Pump Suction Flow--High. In addition, other means to detect
leakage from the RHR system, such as sump monitoring and area
radiation monitoring, are also available. Also, in accordance with
Technical Specification Administrative Requirement 6.2.F.1, LaSalle
has a leakage reduction program to reduce leakage from those
portions of systems outside primary containment that contain
radioactive fluids. RHR, including piping and components associated
with the shutdown cooling mode, is part of this program, which
includes periodic visual inspection of the system for leakage.
The previous evaluation of diversity of isolation parameters, as
presented in Table 5.2-8 of the UFSAR remains unchanged. Adequate
diversity of isolation parameters is maintained because there are at
least two different methods available to detect and allow isolation
of the system for a line break, as necessary.
Therefore, 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
[[Page 69337]]
proposes to determine that the requested amendment involves no
significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Stuart A. Richards.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York.
Date of amendment request: October 9, 1998.
Description of amendment request: The proposed amendment would
revise Section 6.0, administrative controls, of the Technical
Specifications (TSs). Specifically, TS Sections 6.5.2.1.j, 6.7.1.c, and
6.8.1.a would be revised to correct typographical errors. In addition,
TS Section 6.5.2.2 would be revised to change the membership of the
Nuclear Facility Safety Committee (NFSC). This change would provide
Consolidated Edison (Con Ed) with the flexibility to obtain industry
experts outside of Con Ed to perform the duties of Chairman, or Vice
Chairman, and members of the NFSC.
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. There is no significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment is administrative in nature. It involves
a change in 1) the Nuclear Facilities Safety Committee (NFSC)
Chairman or Vice Chairman to allow the services of an individual
other than a senior official of the Company, and 2) allowing NFSC
membership by other than Con Edison employees. In either case,
concurrence by the Senior Vice President, Nuclear Operations is
required.
These changes do not affect possible initiating events for
accidents previously evaluated or alter the configuration or
operating of the facility. The Limiting Safety Systems Settings and
Safety Limits specified in the current Technical Specifications
remain unchanged. Therefore, the proposed changes to the subject
Technical Specification would not increase the probability or
consequences of an accident previously evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
As stated above, the proposed changes are administrative in
nature. The safety analysis of the facility remains complete and
accurate. There are no physical changes to the facility, and the
plant conditions for which the design basis accidents have been
evaluated are still valid. The operating procedures and emergency
procedures are unaffected. Consequently, no new failure modes are
introduced as a result of the proposed changes. Therefore, the
proposed changes will not initiate any new or different kind of
accident.
3. There has been no significant reduction in the margin of
safety.
The proposed changes are administrative in nature. Since there
are no changes to the operation of the facility or physical design
the Updated Final Safety Analysis Report (UFSAR) design basis,
accident assumptions, or Technical Specification Bases are not
affected. Therefore, the proposed changes will not result in a
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: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: S. Singh Bajwa, Director.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan.
Date of amendment request: November 9, 1998.
Description of amendment request: The proposed amendment would
delete the Chemical and Volume Control System (CVCS) operability
requirements currently in technical specifications (TS) 3.2 and 3.17.6,
and the associated surveillance testing requirements currently in TS
4.2 and 4.17. The requirements have been added to the Palisades
Operating Requirements Manual (ORM).
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:
Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes delete certain TS requirements which do not
meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical
requirements have been added to a document (the ORM) controlled
under 10 CFR 50.59.
10 CFR 50.59 specifically prohibits changes to the facility as
described in the safety analysis report, and to procedures described
in the safety analysis report ``if the probability of occurrence or
the consequences of an accident or malfunction of equipment
important to safety previously evaluated in the safety analysis
report may be increased''. Since the conditions which limit changes
performed under 50.59 are more restrictive than the conditions which
define changes considered to involve a significant hazards
consideration, moving of a requirement from the TS to a document
which is controlled under 50.59 cannot involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
The proposed changes delete certain TS requirements which do not
meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical
requirements have been added to a document (the ORM) controlled
under 10 CFR 50.59.
10 CFR 50.59 specifically prohibits changes to the facility as
described in the safety analysis report, and to procedures described
in the safety analysis report ``if a possibility for an accident or
malfunction of a different type than any evaluated previously in the
safety analysis report may be created''. Since the conditions which
limit changes performed under 50.59 are more restrictive than the
conditions which define changes considered to involve a significant
hazards consideration, relocation of a requirement from the TS to a
document which is controlled under 50.59 cannot create the
possibility of a new or different kind of accident from any
previously evaluated.
Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed changes delete certain TS requirements which do not
meet the criteria of 10 CFR 50.36(c)(2)(ii), but identical
requirements have been added to a document (the ORM) controlled
under 10 CFR 50.59.
10 CFR 50.59 specifically prohibits changes to the facility as
described in the safety analysis report, and to procedures described
in the safety analysis report if the margin of safety is reduced.
Since the conditions which limit changes performed under 50.59 are
more restrictive than the conditions which define changes considered
to involve a significant hazards consideration, relocation of a
requirement from the TS to a document which is controlled under
50.59 cannot 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.
[[Page 69338]]
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423-3698.
Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: Cynthia A. Carpenter.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina.
Date of amendment request: July 22 and October 22, 1998.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to reflect the licensee's
planned use of fuel supplied by Westinghouse. The Westinghouse fuel has
different design characteristics from the fuel currently in use.
Accordingly, the following changes would need to be made to the TS:
Figure 2.1.1-1, ``Reactor Core Safety Limits--Four Loops in
Operation''; various core operating parameters specified by
Surveillance Requirements 3.2.1.2, 3.2.1.3, and 3.2.2.2; Section 4.2.1,
``Fuel Assemblies''; and Section 5.6.5, ``Core Operating Limits Report
(COLR).''
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:
First Standard
Implementation of this LAR [license amendment request] would not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The revised Reactor Core Safety
Limits Figure further restricts acceptable operation. Moving an
uncertainty factor from the Improved Technical Specifications to the
Core Operating Limits Report (COLR) does not exempt this factor from
regulatory restrictions. COLR parameters are generated by NRC
approved methods with the intent of ensuring that previously
evaluated accidents remain bounding. The COLR is submitted to the
NRC upon implementation of each fuel cycle or when the document is
otherwise revised. No accident probabilities or consequences will be
impacted by this LAR.
Second Standard
Implementation of this LAR would not create the possibility of a
new or different kind of accident from any previously evaluated. The
revised Reactor Core Safety Limits Figure further restricts
acceptable operation. Moving an uncertainty factor from the Improved
Technical Specifications to the COLR does not exempt this factor
from regulatory restrictions. Since the parameter in question is not
being deleted, the possibility of a new or different kind of
accident from any previously evaluated does not exist.
Third Standard
Implementation of this LAR would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. Use of the ZIRLOTM
cladding material has been reviewed and approved in Reference 1 (as
listed in Chapter 2.1 of Topical Report DPC-NE-2009/DPC-NE-2009P,
Duke Power Company Westinghouse Fuel Transition Report).
ZIRLOTM cladding has been extensively used in
Westinghouse nuclear reactors. The changes proposed in this LAR are
necessary to ensure that the performance of the fission product
barriers (cladding) will not be impacted following the replacement
of one fuel design for another. No safety margin will be
significantly impacted.
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: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
NRC Project Director: Herbert N. Berkow.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No.
50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana.
Date of amendment request: November 20, 1998.
Description of amendment request: The licensee has proposed an
amendment to Facility Operating License No. NPF-47, Appendix A--
Technical Specifications (TS) Section 3.1.6, ``Control Rod Pattern.''
The proposed change will be implemented through the establishment of a
new specification added to Section 3.10, ``Special Operations.'' The
proposed specification will be TS Section 3.10.9, ``Control Rod
Pattern--Cycle 8.'' The new TS 3.10.9 is required due to a current
plant-specific configuration where 5 control rods have been inserted
into the reactor core for neutron flux suppression surrounding 2 fuel
assemblies which have been identified as having possible fuel cladding
defects. The new requirement is intended to be effective for the
remainder of the current fuel cycle (Cycle 8), and is in force when rod
withdrawal operations begin from a condition of 100% rod density to 20%
rated thermal power (RTP).
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 request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Accidents analyzed in the SAR have been examined for any impact
caused by this exception to the [Banked Position Withdrawal
Sequence] BPWS operation. The limiting event is the [Control Rod
Drive Accident] CRDA as described in SAR Sections 4.3.2 and 15.4.9.
The limit on energy addition to the fuel is 280 cal/gm as identified
in the SRP section 15.4.9. Bank Position Withdrawal Sequence is
established to reduce maximum incremental control rod worths and
thus minimize consequences resulting from an accident. The reactor
will be operated as before using BPWS. Having the current rod
configuration with 5 rods to minimize impact on the two fuel
cladding imperfections, in lieu of eight rods inoperable separated
by two cells, will not affect initiators of a Control Rod Drop
Accident. In addition, this existing rod configuration has been
analyzed and the resulting consequences continue to be bounded by
the licensing evaluations. The insertion of the identified control
rods will not affect the assumed reactivity insertion time of any
event. The location of the control rods has been reviewed by GE
using the NRC approved methodology. Operation within these limits
will ensure that the consequences of a transient or accident remain
within the acceptable limits of the evaluation. Specifically, rod
worths for the proposed configuration are bounded by the rod worths
allowed for these configurations per TS; thus, the proposed
configuration is more conservative than that allowed per TS. The
results confirm all assumed limits are maintained. The proposed
change ensures that the consequences of abnormal operation and
accidents are acceptable.
The additional Technical Specification will control the
configuration of the plant to that supported by the evaluation. If
this evaluated configuration is not supported, the plant will be
required to be placed in a configuration where the Control Rod Drop
Accident is not applicable, as the current specification requires.
The plant is therefore maintained within limits as currently
allowed. With these limits the consequences of an event are not
increased.
The probability of an accident is not affected by the proposed
Technical Specification changes since the operation of systems or
equipment that could initiate an accident are not affected.
Therefore, the proposed changes do not significantly increase the
probability or consequences of any previously evaluated accident.
(2) The request does not create the possibility of occurrence of
a new or different
[[Page 69339]]
kind of accident from any accident previously evaluated.
The proposed changes do not involve any alteration of plant
hardware or significant change in plant operation. Assuming the 5
suppression rods are bypassed in lieu of eight rods separated by two
cells does not affect event initiators or event consequences. No
plant modifications are required which would affect plant operation.
Operation with the control rod pattern in the proposed configuration
will ensure the results of a CRDA will remain within the assumptions
of the current safety analysis. The system will continue to ensure
that the limits of control rod worth remain within the assumptions
of the CRDA. The revised Technical Specifications will continue to
assure that plant operation is consistent with the assumptions,
initial conditions, and assumed power distribution and, therefore,
will not create a new type of accident.
The proposed Technical Specifications will maintain the plant in
a configuration supported by evaluation. The response to a CRDA will
be within current accepted limits and therefore no event of a
different kind has been created. The proposed Technical
Specification changes do not introduce any new modes of plant
operation nor involve new system interactions. Therefore, operation
with the 5 suppression rods inserted does not create the possibility
of an occurrence of a new or different kind of accident from any
accident previously evaluated.
(3) The request does not involve a significant reduction in a
margin of safety.
The proposed Technical Specification and the rod pattern control
system will continue to ensure the limits of control rod worth
remain within the assumptions which support the CRDA analysis of 280
cal/gm maximum energy heat addition to the fuel. This imposed limit
of 280 cal/gm provides a margin of safety from the experimental
value of approximately 330 cal/gm at which the fully molten state
for UO2 occurs. The existing rod configuration with 5
suppression rods inserted to minimize impact on the two fuel
cladding imperfections has been analyzed using NRC approved
methodology. Cycle specific evaluation has confirmed that the
consequences resulting from a CRDA continues to be bounded by the
licensing analysis for this event. Since there are no changes in the
acceptance criteria, the proposed changes will not create a
reduction in the margin of safety. These limits establish the
necessary restrictions on power operation and thereby ensure that
the core is operated within the assumptions and initial conditions
of the transient and accident analyses.
As demonstrated in the evaluation, operation within these limits
will ensure that the margin of safety will be maintained to the same
level described in the Technical Specifications Bases and the USAR
and the consequences of the postulated transient or accidents are
not increased. This limit of 280 cal/gm is not exceeded during any
transient or postulated accident. Therefore, the proposed Technical
Specifications to allow startup and continued operation in the low
power region with these control rods inserted do not involve a
significant reduction in 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: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Project Director: John N. Hannon.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana.
Date of amendment request: July 2, 1998.
Description of amendment request: The proposed change will modify
the ACTION Requirements for Technical Specification (TS) 3/4.3.2 for
the Emergency Feedwater Actuation Signal (EFAS). A change to the TS
Bases Section 3/4.3.2 has been included to support this change. The
objective of this change is to add a restriction on the period of time
a channel of EFAS instrumentation can remain in the tripped condition.
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. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The proposed revision to the TS changes the allowed outage time
that a channel of EFAS SGDPI [Steam Generator Differential Pressure
Instrumentation] can be in the tripped condition from a maximum of
approximately 18 months when one channel is inoperable and 92 days
when two channels are inoperable to 48 hours. If a channel were in
the tripped condition and a single failure occurred (failure of one
other channel of EFAS SGDPI), an inadvertent EFAS signal would be
generated. During a Design Basis MSLB [Main Steam Line Break] or FLB
[Feedwater Line Break] Accident, this single failure would send EFW
[Emergency Feedwater] to the faulted steam generator. The Waterford
3 safety analysis assumes that the excess Reactor Coolant System
(RCS) cooldown and return to power associated with the MSLB will be
terminated when the faulted steam generator empties. If additional
EFW were added, the RCS cooldown would be extended and the return to
power may increase.
Reducing the time that a channel of EFAS SGDPI can be placed in
the tripped condition will reduce the probability of this scenario
occurring during a Design Basis Accident. Since the allowed outage
time for a channel of EFAS SGDPI is being limited to 48 hours, this
is considered an off-normal operation and a single failure is not
required to be postulated during a Design Basis Accident in the
accident analysis. Reducing the time the channel can be placed in
the tripped condition and thus, the exposure time to this scenario,
would not be an accident initiator. The proposed change of being
more conservative relative to allow[ed] outage time in the tripped
condition will not affect the assumptions, design parameters, or
results of any accident previously evaluated.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change provides a more conservative
allowed outage time for the channel to be in the tripped condition.
There has been no physical change to plant systems, structures or
components nor will the proposed change reduce the ability of any of
the safety-related equipment required to mitigate Anticipated
Operational Occurrences or accidents. The configuration required by
the proposed specification is permitted by the existing
specification.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The proposed change provides a more conservative allowed outage
time for the channel to be in the tripped condition. By reducing the
allowed outage time, the probability is reduced that a single
failure (failure of one channel of EFAS SGDPI with one channel in
the tripped condition) would occur that would send EFW to the
faulted steam generator. Therefore, the only change to the margin of
safety would be an increase. Since the allowed outage time for a
channel of EFAS SGDPI is being limited to 48 hours, this is
considered an off-normal operation and a single failure is not
required to be postulated during a Design Basis Accident in the
accident analysis. The proposed changes do not affect the limiting
conditions for operation or their bases.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
[[Page 69340]]
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 New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street NW., Washington, DC 20005-3502.
NRC Project Director: John N. Hannon.
Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida.
Date of amendment request: December 31, 1997, as supplemented
November 25, 1998.
Description of amendment request: The proposed amendment will
revise the St. Lucie Unit 2 Technical Specifications to permit an
increase in the allowed Spent Fuel Pool (SFP) storage capacity.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Analyses to support the proposed fuel pool capacity increase
have been developed using conservative methodology. The analysis of
the potential accidents summarized below has shown that there is no
significant increase in the consequences of any accident previously
analyzed. A review of relevant plant operations has also
demonstrated that there is no significant increase in the
probability of occurrence of any accident previously analyzed. This
conclusion is also discussed below.
Previously evaluated accidents that were examined for this
proposed license amendment include: Fuel Handling Accident, Spent
Fuel Cask Drop Accident, and Loss of all Fuel Pool Cooling.
There will be no change in the mode of plant operation or in the
availability of plant systems as a result of this proposed change;
the systems interfacing with the spent fuel pool have previously
encountered borated pool water and are designed to interact with
irradiated spent fuel and remove the residual heat load generated by
isotopic decay. The proposed amendment does not require a change in
the maintenance interval or maintenance scope for the fuel pool
cooling system or for the spent fuel cask crane. The frequency of
cask handling operations and the maximum weight carried by the crane
is not increased as a result of the proposed license amendment.
Thus, there will be no increase in the probability of a loss of fuel
pool cooling or in the probability of a failure of the cask crane as
a result of the proposed amendment.
There will not be a significant increase in the frequency of
handling discharged assemblies in the fuel pool as a result of this
change; any handling of fuel in the spent fuel pool will continue to
be performed in borated water. If the license amendment is approved,
there will be a one-time repositioning of certain discharged
assemblies stored in the fuel pool to comply with the revised
positioning requirements, but the increased pool storage capacity
will permit the deferral of spent fuel handling associated with cask
loading operations. Fuel manipulation during the repositioning
activity will be performed in the same manner as for fuel placed in
the spent fuel pool during refueling outages. There will be no
changes in the manner of handling fuel discharged from the core as a
result of refueling; administrative controls will continue to be
used to specify fuel assembly placement requirements. The relative
positions of Region I and Region II storage locations will remain
the same within the fuel pool. Therefore, the probability of a fuel
handling accident has not been significantly increased.
The consequences of a fuel handling accident have been
evaluated. The radioactive release consequences of a dropped fuel
assembly are not affected by the proposed increase in fuel pool
storage capacity. They remain bounded by the results of calculations
performed to justify the existing St. Lucie Unit 2 fuel storage
racks and burnup limits. At the limiting fuel assembly burnup,
radioactive releases from a dropped assembly would be only a small
fraction of NRC guidelines. The input parameters employed in
analyzing this event are consistent with the current values of fuel
enrichment, discharge burnup and uranium content used at St. Lucie
Unit 2 and with future use of the ``value-added'' fuel pellet
design. Thus, the consequences of the fuel assembly drop accident
would not be significantly increased from those previously
evaluated.
The capability of the fuel pool cooling system to handle the
increased number of discharged assemblies has been examined. The
impact of a total loss of spent fuel pool cooling flow on available
equipment recovery time and on fuel cladding integrity has also been
evaluated. For the limiting full core discharge, sufficient time
remains available to restore cooling flow or to provide an alternate
makeup source before boiloff results in a fuel pool water level less
than that needed to maintain acceptable radiation dose levels.
Analysis has shown that in the event of a total loss of fuel pool
cooling fuel cladding integrity is maintained. Therefore, the
consequences of a loss of fuel pool cooling event, including the
effect of the proposed increase in fuel pool storage capacity, have
not been significantly increased from previously analyzed results
for this type of accident.
The analysis of record pertaining to the radiological
consequences of the hypothetical drop of a loaded spent fuel cask
just outside the Fuel Handling Building was examined to determine
the impact of the increased fuel storage capacity on this accident's
results. The results of the previously performed analysis were
determined to bound the conditions described by the proposed license
amendment, thus the consequences of the cask drop accident would not
be significantly increased as a result of this change.
It is concluded that the proposed amendment to increase the
storage capacity of the St. Lucie Unit 2 spent fuel pool will not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different type of accident from any accident previously
evaluated.
In this license amendment FPL [Florida Power & Light Co.]
proposes to credit the negative reactivity associated with a portion
of the soluble boron present in the spent fuel pool. Soluble boron
has always been present in the St. Lucie Unit 2 spent fuel pool; as
such the possibility of an inadvertent fuel pool dilution has always
existed. However, the spent fuel pool dilution analysis demonstrates
that a dilution of the Unit 2 spent fuel pool which could increase
the pool keff to greater than 0.95 is not a credible
event. Neither implementation of credit for the reactivity of fuel
pool soluble boron nor the proposed increase in the fuel pool
storage capacity will create the possibility of a new or different
type of accident at St. Lucie Unit 2.
An examination of the limiting fuel assembly misload has
determined that this would not represent a new or different type of
accident. None of the other accidents examined as a part of this
license submittal represent a new or different type of accident;
each of these situations has been previously analyzed and determined
to produce acceptable results.
The proposed license amendment will not result in any other
changes in the mode of spent fuel pool operation at St. Lucie Unit 2
or in the method of handling irradiated nuclear fuel. The spatial
relationship between the fuel storage racks and the cask crane range
of motion is not affected by the proposed change.
As a result of the evaluation and supporting analyses, FPL has
determined that the proposed fuel pool capacity increase does not
create the possibility of a new or different type of accident from
any accident previously evaluated.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
FPL has determined, based on the nature of the proposed license
amendment that the issue of margin of safety, when applied to this
fuel pool capacity increase, should address the following areas:
1. Fuel Pool reactivity considerations
2. Fuel Pool boron dilution considerations
3. Thermal-Hydraulic considerations
4. Structural loading and seismic considerations
The Technical Specification changes proposed by this license
amendment, the proposed spent fuel pool storage
[[Page 69341]]
configuration and the existing Technical Specification limits on
fuel pool soluble boron concentration provide sufficient safety
margin to ensure that the array of fuel assemblies stored in the
spent fuel pool will always remain subcritical. The revised spent
fuel storage configuration is based on a Unit 2 specific criticality
analysis performed using methodology consistent with that approved
by the NRC. Additionally, the soluble boron concentration required
by current Technical Specifications ensures that the fuel pool
keff will be always be maintained substantially less than
0.95.
The Unit 2 criticality analysis established that the
keff of the spent fuel pool storage racks will be less
than 1.0 with no soluble boron in the fuel pool water, including the
effect of all uncertainties and tolerances. Credit for the soluble
boron actually present is used to offset uncertainties, tolerances,
off-normal conditions and to provide margin such that the spent fuel
pool keff is maintained less than or equal to 0.95. FPL
has also demonstrated that a decrease in the fuel pool boron
concentration such that keff exceeds 0.95 is not a
credible event.
Current Technical Specifications require that the fuel pool
boron concentration be maintained greater than or equal to 1720 ppm.
This boron value is substantially in excess of the 520 ppm required
by the uncertainty and reactivity equivalencing analyses discussed
in this evaluation and the 1266 ppm value required to maintain
keff less than or equal to 0.95 in the presence of the
most adverse mispositioned fuel assembly.
The St. Lucie Unit 2 fuel pool boron concentration will continue
to be maintained significantly in excess of 1266 ppm; the proposed
license amendment will not result in changes in the mode of
operation of the refueling water tank (RWT) or in its use for makeup
to the fuel pool. Thus, operation of the spent fuel pool following
the proposed change, combined with the existing fuel pool boron
concentration Technical Specification limit of 1720 ppm, will
continue to ensure that keff of the fuel pool will be
substantially less than 0.95.
Even if this not-credible dilution event was to occur, no
radiation would be released; the only consequence would be a
reduction of shutdown margin in the fuel pool. The volume of
unborated water required to dilute the fuel pool to a
keff of 0.95 is so large (in excess of 358,900 gallons to
dilute the fuel pool to 520 ppm boron) that only a limited number of
water sources could be considered potential dilution sources. The
likelihood that this level of water use could remain undetected by
plant personnel is extremely remote.
In meeting the acceptance criteria for fuel pool reactivity, the
proposed amendment to increase the storage capacity of the existing
fuel pool racks does not involve a significant reduction in the
margin of safety for nuclear criticality.
Calculations of the spent fuel pool heat load with an increased
fuel pool inventory were performed using ANSI/ANS-5.1-1979
methodology. This method was demonstrated to produce conservative
results through benchmarking to actual St. Lucie Unit 2 fuel pool
conditions and by comparison of its results to those generated by a
calculation using Auxiliary Systems Branch Technical Position 9-2
methodology. Conservative methods were also used to demonstrate fuel
cladding integrity is maintained in the absence of cooling system
forced flow. The results of these calculations demonstrate that, for
the limiting case, the existing fuel pool cooling system can
maintain fuel pool conditions within acceptable limits with the
increased inventory of discharged assemblies.
Therefore, the proposed change does not result in a significant
reduction in the margin of safety with respect to thermal-hydraulic
or spent fuel cooling considerations.
The primary safety function of the spent fuel pool and the fuel
storage racks is to maintain discharged fuel assemblies in a safe
configuration for all environments and abnormal loadings, such as an
earthquake, a loss of pool cooling or a drop of a spent fuel
assembly during routine spent fuel handling. The proposed increase
in spent fuel inventory on the fuel pool and the existing storage
racks have been evaluated and show that relevant criteria for fuel
rack stresses and floor loadings have been met and that there has
been no significant reduction in the margin of safety for these
criteria.
The NRC staff has reviewed the licensee's analysis and the changes
proposed in the November 25, 1998 supplement to the original submittal
and based on this review, it appears that the three standards of
50.92(c) continue to be satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida.
Date of amendment request: October 27, 1998.
Description of amendment request: The licensee proposed to change
Technical Specification (TS) 6.3, Facility Staff Qualifications, in
order to incorporate qualifications for the Multi-Discipline
Supervisor. The current TS requires that plant staff meet the
requirements of the American National Standards Institute (ANSI) N18.1-
1971, which requires non-licensed supervisors to have a high school
diploma or equivalent and a minimum of 4 years experience in the craft
or discipline they supervise. The proposed change requires the Multi-
Discipline Supervisor to have, (1) a high school diploma or equivalent,
(2) a minimum of 4 years of related technical experience, which shall
include 3 years of power plant experience of which one year is at a
nuclear power plant, and (3) completed the Multi-Discipline Supervisor
training 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 facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed changes are administrative in nature addressing
personnel qualification issues. The Multi-Discipline Supervisor
(MDS) position will be filled with personnel who are experienced in
one or more technical disciplines (maintenance, operations,
engineering, or other related technical discipline). Fundamental
working knowledge of tasks being performed will be acquired through
the MDS initial training program. The training concentrates on
developing the skills and knowledge of an MDS to safely oversee
tasks for multi-discipline work teams. Therefore, four years
experience in any related technical discipline or disciplines
combined with the MDS training program provide adequate technical
knowledge for proper job oversight. These proposed changes will not
involve a significant increase in the probability or consequences of
an accident previously evaluated because they do not affect
assumptions contained in plant safety analyses, the physical design
and/or operation of the plant, nor do they affect Technical
Specifications that preserve safety analysis assumptions. Therefore,
the proposed changes do not affect the probability or consequences
of accidents previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The changes being proposed are administrative in nature and do
not affect assumptions contained in plant safety analyses, the
physical design and/or modes of plant operation defined in the
facility operating license, or Technical Specifications that
preserve safety analysis assumptions. These changes address
qualification requirements for the MDS position. Since the proposed
changes do not change the qualifications for those individuals
responsible for the actual licensed operation of the facility,
operation of the facility in accordance with the proposed amendments
would not create the possibility of a new or different kind of
accident from any accident
[[Page 69342]]
previously evaluated. No new failure mode is introduced due to the
administrative changes since the proposed changes do not involve the
addition or modification of equipment nor do they alter the design
or operation of affected plant systems, structures, or components.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are unchanged by the proposed
amendments. The proposed changes to add the MDS position have
management and administrative controls associated with the required
qualification requirements. The Turkey Point Technical
Specifications will ensure that any individual filling the MDS
position has the requisite education, experience, and training. As a
result, operation of the facility in accordance with the proposed
changes would not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey.
Date of amendment request: November 5, 1998.
Description of amendment request: The proposed Technical
Specification change will modify the safety limits and surveillances of
the LPRM and APRM systems and related Bases pages to ensure the APRM
channels respond within the necessary range and accuracy and to verify
channel operability. In addition, an unrelated change to the Bases of
Specification 2.3 is included to clarify some ambiguous language.
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 technical specification changes to the limits
and surveillance requirements of the LPRM and APRM systems are
provided to ensure the APRM channels respond within the necessary
range and accuracy and to verify channel operability. If one or more
monitored parameters exceeded their specified limits, the RPS
initiates a reactor scram signal to preserve the integrity of the
fuel cladding and the Reactor Coolant System and minimize the energy
that must be absorbed following a loss of coolant accident.
Therefore, the probability of occurrence or the consequences of an
accident previously evaluated in the [safety analysis report] SAR
will not increase as a result of these changes.
2. The proposed technical specification changes to the limits
and surveillance requirements of the LPRM and APRM systems are
provided to ensure the APRM channels respond within the necessary
range and accuracy and to verify channel operability. The proposed
changes are designed to ensure the APRM system responds in a manner
that ensures the safety limits, limiting safety system settings,
limiting conditions for operations, as well as design parameters for
the APRM system and individual components are continuously met.
Therefore, the proposed activity does not create the possibility for
an accident or malfunction of a different type than any previously
identified in the SAR.
3. The proposed change does not involve a significant reduction
in the margin of safety. When the APRMs exceed their specified
limits, the RPS initiates a reactor scram signal to preserve the
integrity of the fuel cladding and the Reactor Coolant System and
minimize the energy that must be absorbed following a loss of
coolant accident. The proposed changes are designed to assure the
APRM system responds in a manner that ensures the safety limits,
limiting safety system settings, limiting conditions for operations,
as well as design parameters for the APRM system and individual
components are continuously met. Therefore, the margin of safety
will not be 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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
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.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania.
Date of amendment request: November 25, 1998.
Description of amendment request: The proposed amendment will
change the surveillance specification for Once Through Steam Generator
(OTSG) inservice inspections for TMI-1 Cycle 13 refueling outage
examinations which would be applicable for the next operating cycle
only, Operating Cycle 13.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed changes do not represent a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed flaw disposition strategy, based on measurable eddy
current parameters of axial and circumferential extent for Inside
Diameter (ID) Initiated Inter-Granular Attack (IGA), will continue
to provide high confidence that unacceptable flaws that do not have
the required structural integrity to withstand a postulated MSLB
[main steam line break] are removed from service. The axial and
circumferential length limits for eddy current ID degradation
indications meet the Draft Regulatory Guide 1. 121 * * * acceptance
criteria for margin to failure for MSLB-applied differential
pressure and axial tube loads. The capability for detection of flaws
is unaffected; and the identification of tubes that should be
repaired or removed from service is maintained. The operation of the
OTSGs or related structures, systems, or components is otherwise
unaffected. Therefore, neither the probability nor consequences of
[an] SGTR [steam generator tube rupture] is significantly increased
either during normal operation or due to the limiting loads of [an]
MSLB accident.
Neither the change in voltage normalization for the eddy current
examinations, nor the administrative change in clarification of the
reporting requirements, as described above, could significantly
affect the probability of occurrence or consequences of any accident
previously evaluated. These changes are administrative only.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
because there are no hardware changes involved nor changes to any
operating practices. These changes involve only the OTSG tube
inservice inspection surveillance requirements, which could only
affect the potential for OTSG primary-to-secondary leakage. The
proposed changes continue to impose flaw length limits for ID IGA to
assure tube structural and leakage integrity, as confirmed by 12R
(and post 12R) tube pull sample examinations and pressure testing.
In addition, neither the change in voltage normalization for the
eddy current examinations nor the administrative change in the
description of the reporting requirements, as described above, could
possibly create the possibility of an accident
[[Page 69343]]
of a new or different type from any previously evaluated. These
changes are included only to modify the plant's eddy current
normalization to the industry standard, and clarify the reporting
period for submittal of the OTSG inspection results to the NRC
[Nuclear Regulatory Commission]. Therefore, these changes do not
create the potential for any other kind of accident different from
those that have been evaluated.
C. These proposed changes do not involve a significant reduction
in a margin of safety because the margins of safety defined in Draft
Regulatory Guide 1. 121 * * * are retained. The probability of
detecting degradation is unchanged since the bobbin coil eddy
current methods will continue to be the primary means of initial
detection and the probability of leakage from any indications left
in service remains acceptably small. The strategy for dispositioning
ID initiated IGA will continue to provide a high level of confidence
that tubes exceeding the allowable limits for tube integrity are
repaired or removed from service.
In addition, neither the change in voltage normalization for the
eddy current examinations nor the administrative change in the
description of the reporting requirements, as described above, could
significantly affect a margin of safety. These changes are
administrative in nature and are included only to align TMI-1's
voltage normalization to the industry standard, and clarify the
reporting period, respectively.
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.
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut.
Date of amendment request: November 10, 1998.
Description of amendment request: The proposed changes would modify
Technical Specifications 3.3.1.1, ``Reactor Protective
Instrumentation,'' and 3.3.2.1, ``Engineered Safety Feature Actuation
System Instrumentation'' to restrict the time a reactor protection or
engineered safety feature actuation channel can be in the bypass
position to 48 hours, from an indefinite period of time. Most of these
proposed changes were originally submitted in a letter dated May 14,
1998. The licensee withdrew its original request and submitted a new
request in its November 10, 1998, letter.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
In accordance with 10CFR50.92, NNECO [Northeast Nuclear Energy
Company] has reviewed the proposed changes and has concluded that
they do not involve a significant hazards consideration (SHC). The
basis for this conclusion is that the three criteria of
10CFR50.92(c) are not compromised. The proposed changes do not
involve an SHC because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to restrict the time [* * *] reactor
protection or engineered safety feature actuation channels can be in
the bypass position to 48 hours, from an indefinite period of time,
has no effect on the design of the Reactor Protection System (RPS)
or the Engineered Safety Feature Actuation System (ESFAS) and does
not affect how these systems operate. In addition, this will
minimize the susceptibility of these systems to the remote
possibility of fault propagation between channels. However, this
proposed change will require an inoperable pressurizer high pressure
reactor protection channel to be placed in the tripped condition
within 48 hours. With a pressurizer pressure channel in the tripped
condition, the high failure of a second pressurizer pressure channel
would initiate a reactor trip and open both pressurizer power
operated relief valves (PORVs). Opening the pressurizer PORVs would
result in an undesired loss of primary coolant. Thus, this change
will increase the probability of occurrence of a previously
evaluated accident. However, this would not place the plant in an
unanalyzed condition since FSAR [Final Safety Analysis Report]
Section 14.6.1 analyzes the inadvertent opening of both PORVs, the
release of reactor coolant can be terminated by closure of the PORV
block valves from the control room, and the Emergency Operating
Procedures provide guidance on how to address this situation.
Therefore, this change does not significantly increase the
probability or consequences of an accident previously evaluated.
The proposed change to increase the time a second RPS or ESFAS
channel can be removed from service (from 2 hours to 48 hours),
provided one of the inoperable channels is placed in the tripped
condition, has no effect on the design of the RPS or ESFAS and does
not affect how these systems operate. These systems will still
function as designed to mitigate design basis accidents. However,
this change will also impact the probability of occurrence of a
previously evaluated accident since it will allow a second
pressurizer high pressure reactor protection channel to be placed in
the tripped condition for 48 hours instead of the current 2 hour
time limit. The impact of this change is bounded by the proposed
change to require an inoperable pressurizer high pressure reactor
protection channel to be placed in the tripped condition after 48
hours as previously discussed. Therefore, this change does not
significantly increase the probability or consequences of an
accident previously evaluated.
The proposed change to apply a more restrictive action statement
to the loss of turbine load reactor trip function has no effect on
the design of this trip function and does not affect how this trip
function operates. Also, this trip function is not assumed to
operate to mitigate any design basis accident. Therefore, this
change does not significantly increase the probability or
consequences of an accident previously evaluated.
The proposed change to require a channel calibration every 18
months for the loss of turbine load reactor trip function and for
the wide range logarithmic neutron flux monitors has no effect on
the design of either the loss of turbine load reactor trip function
or the wide range logarithmic neutron flux monitors. Also, neither
of these are assumed to operate to mitigate any design basis
accident. Therefore, this change does not significantly increase the
probability or consequences of an accident previously evaluated.
The proposed change to exclude the neutron detectors from the
channel calibration requirement has no effect on the design of the
neutron detectors and has no significant effect on how these
detectors operate. The detectors are passive devices with minimal
drift. In addition, slow changes in the sensitivity of the linear
power range flux detectors is compensated for by performing the
daily calorimetric calibration and the monthly calibration using the
incore detectors. These detectors will still function as designed to
mitigate design basis accidents. Therefore, this change does not
significantly increase the probability or consequences of an
accident previously evaluated.
The proposed change to add the license amendment numbers to
Technical Specification Page 3/4 3-9 will not result in a technical
change to the Millstone Unit No. 2 Technical Specifications. The RPS
will continue to function as before. Therefore, this change does not
significantly increase the probability or consequences of an
accident previously evaluated.
The proposed change to correct the surveillance requirement
referenced in an action statement has no effect on the design of the
ESFAS and does not affect how this system operates. The ESFAS will
still function as designed to mitigate design basis accidents.
Therefore, this change does not significantly increase the
probability or consequences of an accident previously evaluated.
The proposed change to add a reference to the reactor coolant
pump low speed reactor trip function to a note that states this trip
[[Page 69344]]
may be bypassed <5% power, and that the bypass must be automatically
removed [greater than or equal to] 5% will not affect this reactor
trip function. This bypass capability currently exists in the design
of the Millstone Unit No. 2 RPS, and is the same bypass feature
referenced for the reactor coolant flow low reactor trip function.
Both of these reactor trip functions provide protection for a
reduction in RCS [reactor coolant system] flow. The addition of this
note will not result in any technical change to the Millstone Unit
No. 2 RPS. The RPS will continue to function as before. Therefore,
this change does not significantly increase the probability or
consequences of an accident previously evaluated.
The proposed change to correct the power level high trip
setpoint on Technical Specification Page 2-4 will not result in any
change to the actual plant setpoint for this RPS trip function. As a
result of this proposed change, the setpoint listed on Page 2-4 will
agree with the setpoint previously approved by the NRC, and
currently used by the RPS. The change has no effect on the design of
the RPS and does not affect how this system operates. The RPS will
still function as designed to mitigate design basis accidents.
Therefore, this change does not significantly increase the
probability or consequences of an accident previously evaluated.
The information added to the Bases of the affected Technical
Specifications to provide a discussion of how the RPS and ESFAS are
affected by the proposed changes, the effect the action statements
have on the operation of the RPS and ESFAS, and to discuss the
impact of surveillance testing on RPS operability will have no
effect on equipment operation. The RPS and ESFAS will continue to
function as designed to mitigate design basis accidents. Therefore,
this change does not significantly increase the probability or
consequences of an accident previously evaluated.
Thus, this License Amendment Request does not impact the
probability of an accident previously evaluated nor does it involve
a significant increase in the consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes do not
introduce any new failure modes. They will not alter assumptions
made in the safety analysis and licensing basis. The RPS and the
ESFAS will still function as designed to mitigate design basis
accidents.
Therefore, these changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since
they have no impact on any safety analysis assumption. The proposed
changes do not decrease the scope of equipment currently required to
be operable or subject to surveillance testing, nor do the proposed
changes affect any instrument setpoints or equipment safety
functions.
The effectiveness of Technical Specifications will be maintained
since the changes will not alter the operation of any RPS or ESFAS
function. In addition, most of the changes are consistent with the
Calvert Cliffs RPS and ESFAS Technical Specifications model provided
in Enclosure 3 of the NRC correspondence dated April 16, 1981 [R. A.
Clark letter to W. G. Council, Evaluation of the Reactor Protection
System Inoperable Channel Condition at Millstone Nuclear Power
Station, Unit No. 2, dated April 16, 1981] and with the new,
improved Standard Technical Specifications (STS) for Combustion
Engineering plants (NUREG-1432).
Therefore, there is no significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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.
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota.
Date of amendment request: November 25, 1997, as supplemented
September 25 and November 11, 1998. The September 25, 1998, supplement
incorrectly references the original request as October 31, 1997, rather
than November 25, 1997.
Description of amendment request: The proposed amendment would
revise the Technical Specifications for the condensate storage tank
(CST) low level suction transfer setpoint for the high pressure coolant
injection (HPCI) and reactor core isolation cooling (RCIC) systems to
allow removing one CST from service for maintenance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed setpoint change and temporary level switch cross
connection will not affect the way the suction transfer equipment
functions, introduce new failure modes, or significantly increase
the probability of failure of this equipment.
A slight increase in the probability of failure of the CST
suction low level automatic transfer function may result, however,
during plant operation with one CST in service and the CST low level
transfer switches temporarily cross connected. This temporary
modification preserves the redundancy of the automatic level
transfer logic and allows HPCI and RCIC to remain aligned to the
condensate storage system.
When the switches are cross connected, sections of piping and
instrument tubing will be shared by both level switches. The
probability that freezing or plugging of a common section of piping
or tubing will disable both switches will be slightly higher than
during two CST operation with the level switch piping in its normal
configuration.
The level switches would be cross connected at infrequent
intervals to permit prudent and timely CST preventive maintenance
and at the same time continue to provide HPCI and RCIC with a source
of reactor makeup quality water. In the unlikely event of a spurious
actuation of either system, only high quality water would be
injected into the reactor vessel.
Overall, the possibility of freezing or plugging of piping and
tubing associated with the automatic transfer level switches has
been shown to be very small, with or without the temporary level
switch cross connection in place. During periods of operation with
one CST, we believe the small additional opportunity for level
instrument failure due to freezing or plugging is more than
compensated for by the benefits of maintaining a high quality source
of water to the HPCI and RCIC pumps.
The proposed level switch cross connection will not affect the
way the suction transfer equipment functions. The cross connection
tubing will be evaluated for seismic loads equivalent to the
existing instrument piping. Rupture of the tubing will not prevent
the function of the level switches from being accomplished and no
other equipment important to safety is impacted by these changes.
Technical Specification and other specified margins of safety
are effectively increased by the proposed changes. The HPCI/RCIC low
CST level suction transfer level is being adjusted upward in the
conservative direction.
The changes do not present the opportunity for a new release
path for radioactive material.
These changes have no impact on the protection of the health and
safety of the public.
(2) The proposed amendment will not create the possibility of a
new or different
[[Page 69345]]
kind of accident from any accident previously analyzed.
No system, structure, or component (SSC) described in the USAR
[Updated Safety Analysis Report] as important to safety is affected
by these changes except for the low level CST HPCI/RCIC suction
transfer function. Postulated malfunctions related to the proposed
changes to the low level switches are bounded by the failure of the
HPCI system, which has been previously evaluated in the USAR. The
RCIC system is not relied upon to mitigate any USAR design basis
accident.
No new types of credible events could be identified which could
be created by the proposed setpoint change and level switch cross
connection. No new failure modes are associated with the proposed
changes [sic].
(3) The proposed amendment will not involve a significant
reduction in the margin of safety.
No margin of safety is reduced. Technical Specification and
other specified margins of safety are effectively increased by the
proposed activities. The HPCI/RCIC low CST level suction transfer
setpoint is being adjusted upward in the conservative direction.
Cross connecting the level switches associated with this transfer
will preserve the redundancy built into the logic during extended
outages of one CST. A small additional reduction in the reliability
of the automatic transfer logic due to possible freezing or plugging
of common instrument piping results when the level switches are
temporarily cross connected during infrequent periods of operation
with one CST in service. This small reduction in reliability of the
automatic transfer function is fully compensated for by the ability
to perform necessary and prudent preventive maintenance on the CSTs
while at the same time supplying the HPCI and RCIC systems with
water from the preferred high quality source.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota.
Date of amendment requests: November 25, 1998.
Description of amendment requests: The proposed amendments would
modify the technical specifications (TS) (TS 3.2 and Table 3.5-2B) to
allow limited inoperability of boric acid storage tank (BAST) level
channels and transfer logic channels to provide for required testing
and maintenance of the associated components.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not affect any system that is a
contributor to initiating events for previously evaluated design
basis accidents. Therefore, the proposed changes do not involve a
significant increase in the probability of an accident previously
evaluated. The proposed Actions 34, 35 and 36 will allow limited
continued plant operation with portions of BAST to RWST [refueling
water storage tank] transfer instrumentation inoperable. However,
because the proposed actions place time limits on inoperability
comparable to those already approved for use in the Prairie Island
Technical Specifications the proposed changes do not involve a
significant increase in the consequences of an accident previously
evaluated. The remaining proposed changes to Table TS.3.5-2B and to
Specification 3.2B are administrative in nature. The changes to
Table 3.5-2B incorporate design information on the BAST to RWST
transfer instrumentation which clarifies the operability
requirements for the instrumentation. The changes to Specification
3.2.B add a reference to Table TS.3.5-2B. Therefore, because of the
administrative nature of the changes, they do not involve a
significant increase in the consequences of an accident previously
evaluated.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The proposed changes do not alter the design or function of any
plant component and do not install any new or different equipment.
The proposed changes do not alter the operation of any plant
component in a manner which could lead to a new or different kind of
accident. Therefore the possibility of a new or different kind of
accident from those previously analyzed has not been created.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The proposed Actions 34, 35 and 36 will allow limited continued
plant operation with portions of the BAST to RWST transfer
instrumentation inoperable. However, because the proposed actions
place time limits on inoperability comparable to those already
approved for use in the Prairie Island Technical Specifications the
proposed changes do not involve a significant reduction in the
margin of safety. The remaining proposed changes to Table TS.3.5-2B
and to Specification 3.2.B are administrative in nature. The changes
to Table 3.5-2B incorporate design information on the BAST to RWST
transfer instrumentation which clarifies the operability
requirements for the instrumentation. The changes to Specification
3.2.B add a reference to Table TS.3.5-2B. Therefore, because of the
administrative nature of the changes, they do not involve a
significant reduction in the margin of safety.
NRC staff has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo
County, California.
Date of amendment request: September 3, 1998.
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 to change TS 3.4.9.1, ``Reactor Coolant
System--Pressure/Temperature Limits,'' Figure 3.4-2, ``Reactor Coolant
System Heatup Limitations--Applicable Up to 12 EFPY,'' and Figure 3.4-
3, ``Reactor Coolant System Cooldown Limitations--Applicable Up to 12
EFPY,'' to extend the applicability up to 16 effective full power years
(EFPY). The affected TS Bases would also be appropriately revised.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to Figures 3.4-2 and 3.4-3 of Technical
Specification (TS) 3.4.9.1 and the associated Bases adjust the
reactor
[[Page 69346]]
coolant system (RCS) heatup and cooldown pressure/temperature (P/T)
limits to permit operation through 16 effective full power years
(EFPY). The 16 EFPY P/T limits are more restrictive than the current
limits; this accounts for an expected incremental increase in
reactor vessel embrittlement, and assures the reactors will continue
to be operated within acceptable stresses and at temperatures for
which the reactor vessel metal exhibits ductile properties. The P/T
limits developed for 16 EFPY were determined in accordance with 10
CFR 50, Appendix G, and maintain the same margins of safety as the
current limits. The proposed changes will not impact the probability
of overpressurization or brittle fracture of the vessel, and
therefore will not impact the consequences of an accident.
The present low temperature overpressure protection (LTOP)
pressure and enable temperature setpoints were reviewed and found to
be acceptable and conservative for use through 16 EFPY, based on use
of ASME Code Case N-514, which provides acceptable margins to the
prevention of vessel overpressurization and brittle fracture.
Therefore, there is no change to the consequences of accidents
previously analyzed. Since no changes are proposed in the actual
LTOP setpoints, nor any physical alteration of the LTOP system, nor
a change to the method by which the LTOP system performs its
function, there would be no change to the probability of an accident
previously evaluated. The proposed change to the Bases incorporates
use of ASME Code Case N-514, which will benefit DCPP by not
resulting in a reduced RCS P/T window and reduced power-operated
relief valve (PORV) pressure setpoint for LTOP. This maintains the
current level of operator flexibility during heatup and cooldown,
and prevents an increase in the probability of an accident
associated with an inadvertent PORV actuation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to TS 3.4.9.1, ``Reactor Coolant System--
Pressure/Temperature Limits,'' do not involve any physical
alteration to any plant system or change the method by which any
safety-related system performs its function. The changes to TS
3.4.9.1 account for the effects of an incremental increase in
reactor vessel embrittlement and are requested in order to restrict
future reactor operation to within acceptable stress levels and
temperature regimes in accordance with 10 CFR 50, Appendix G,
requirements. These changes are needed to maintain the current P/T
limit margins of safety as defined by 10 CFR 50, Appendix G, and
ASME XI, Appendix G, for operation through 16 EFPY. The possibility
of a new kind of accident such as catastrophic failure of the
reactor vessel is prevented by maintaining acceptable margins of
safety.
The present LTOP pressure setpoint was reviewed and found to be
acceptable and conservative for the extension of the P/T curves to
16 EFPY.
Additionally, the proposed changes will not affect the ability
of the LTOP system to provide pressure relief at low temperatures,
thereby maintaining the LTOP design basis.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes to TS 3.4.9.1 adjust the RCS heatup and
cooldown P/T limits to permit operation through 16 EFPY. The P/T
limits have been determined in accordance with 10 CFR 50, Appendix
G, and include the safety margins with regard to brittle fracture
required by the ASME Section XI, Appendix G, which maintain the same
margins of safety as the current limits.
The LTOP setpoints were reevaluated using the requirements of
ASME Code Case N-514. This code case was developed to provide the
necessary margins of safety for the prevention of reactor vessel
overpressurization and brittle fracture. The LTOP evaluation results
conclude the current LTOP setpoints are conservative for operation
through 16 EFPY. In addition, avoiding an unnecessary reduction in
the LTOP, the PORV pressure setpoint prevents an increase in the
likelihood of an inadvertent PORV actuation.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room Location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas &
Electric Company, P.O. Box 7442, San Francisco, California 94120.
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: October 29, 1998.
Description of amendment request: The proposed change will relocate
Technical Specification 3/4.7.9 requirements for Snubbers and the
associated Bases to the Technical Requirements Manual.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relocates requirements and surveillances for
Technical Specification 3/4.7.9 that do not meet the criteria for
inclusion in Technical Specifications as identified in 10 CFR
50.36(c)(2)(ii). The affected components are not assumed to be
initiators of analyzed events and are not assumed to mitigate
accident or transient events. The requirements and surveillances for
these affected systems and components will be relocated from the
Technical Specifications to the Technical Requirements Manual, which
is incorporated in the STP UFSAR and will be maintained pursuant to
10 CFR 50.59. In addition, the Snubber operability is 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. The associated changes to the Index are administrative.
Therefore, the change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change relocates requirements and surveillances
applicable to snubbers which 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. The associated changes to the Index are administrative.
Therefore, the change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change relocates requirements and surveillances for
snubbers, that do not meet the 10 CFR 50.36(c)(2)(ii) criteria for
inclusion in Technical Specifications. The change will not reduce a
margin of safety since it has no impact on any safety analysis
assumptions. In addition, the relocated requirements and
surveillances for the affected structure, system, component, or
variable remain the same as the existing Technical Specifications.
Since any future changes to these requirements or the surveillance
procedures will be evaluated per the requirements of 10 CFR 50.59,
there will be no reduction in a margin of safety. The associated
changes to the Index are administrative and have no potential effect
on the margin of safety.
[[Page 69347]]
The proposed change is also consistent with the Westinghouse
Plants Standard Technical Specification, NUREG-1431 approved by the
NRC Staff, revising the Technical Specifications to reflect the
approved content 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, NW, Washington, DC 20036-5869.
NRC Project Director: John N. Hannon.
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: October 29, 1998.
Description of amendment request: The proposed change will relocate
Specification 3/4.3.4, ``Turbine Overspeed Protection,'' to the
Technical Requirements Manual.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relocates the requirements of specification
3/4.3.4, ``Turbine Overspeed Protection,'' that do not meet the
criteria for inclusion in Technical Specifications as identified in
10 CFR 50.36(c)(2)(ii). The specification is not related to any
assumed initiators of analyzed events and are not assumed to
mitigate accident or transient events. The requirement to perform
the testing is not altered by the proposed change. The requirements
of the limiting condition for operation and surveillance testing
will be relocated from the Technical Specifications to the Technical
Requirements Manual, which is incorporated in the STP UFSAR and will
be maintained pursuant to 10 CFR 50.59. In addition, the
surveillance testing details 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. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change relocates the requirements of specification
3/4.3.4, ``Turbine Overspeed Protection,'' that do 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. Does this change involve a significant reduction in a margin
of safety?
The proposed change relocates the requirements of specification
3/4.3.4, ``Turbine Overspeed Protection,'' that do not meet the 10
CFR 50.36 criteria for inclusion in Technical Specifications. The
change will not reduce a margin of safety since it has no impact on
any safety analysis assumptions. In addition, the relocated
requirements applicable to the turbine overspeed protection remain
the same as the existing Technical Specifications requirements.
Since any future changes to these requirements or the surveillance
procedures will be evaluated per the requirements of 10 CFR 50.59,
there will be no reduction in a margin of safety.
The proposed change is also consistent with the Westinghouse
Plants Standard Technical Specification, NUREG-1431 approved by the
NRC Staff. Revising the Technical Specifications to reflect the
approved content, 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, NW, Washington, DC 20036-5869.
NRC Project Director: John N. Hannon.
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: October 29, 1998.
Description of amendment request: The proposed change will relocate
descriptive details of Surveillance Requirement 4.8.1.1.2.g, regarding
maintenance of the diesel generator fuel oil storage tanks (DGFOSTs),
to the Technical Requirements Manual.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relocates descriptive details of
surveillance requirement 4.8.1.1.2.g that do not meet the criteria
for inclusion in Technical Specifications as identified in 10 CFR
50.36(c)(3). The affected descriptive testing details are not
related to any assumed initiators of analyzed events and are not
assumed to mitigate accident or transient events. The requirement to
perform the testing is not altered by the proposed change. The
descriptive details of the surveillance testing will be relocated
from the Technical Specifications to the Technical Requirements
Manual, which is incorporated in the STP UFSAR and will be
maintained pursuant to 10 CFR 50.59. In addition, the surveillance
testing details 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. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change relocates descriptive details of
surveillance testing applicable to the DGFOSTs, which do not meet
the criteria for inclusion in Technical Specifications as identified
in 10 CFR 50.36(c)(3). 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. Does this change involve a significant reduction in a margin
of safety?
The proposed change relocates descriptive details of the
surveillance testing applicable to the DGFOSTs, that do not meet the
10 CFR 50.36 criteria for inclusion in Technical Specifications. The
change will not reduce a margin of safety since it has no impact on
any safety analysis assumptions. In addition,
[[Page 69348]]
the relocated surveillance testing details for the DGFOSTs remain
the same as the existing Technical Specifications. Since any future
changes to these requirements or the surveillance procedures will be
evaluated per the requirements of 10 CFR 50.59, there will be 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 Specifications to
reflect the approved NUREG-1431 content 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, NW, Washington, DC 20036-5869.
NRC Project Director: John N. Hannon.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri.
Date of application request: October 31, 1997, as supplemented by
letter dated September 29, 1998. This notice supersedes the staff's
proposed no significant hazards consideration determination evaluation
for the requested changes that was published on January 14, 1998 (63 FR
2283).
Description of amendment request: The proposed amendment
application would change Tables 3.3-3, 3.3-4, and 4.3-2 of the
technical specifications (TS) to revise the engineered safety feature
actuation system (ESFAS) Functional Unit 6.f, Loss of Offsite Power-
Start Turbine-Driven Pump. Table 3.3-2 would be revised to create
separate functional units for the analog and digital portions of the
ESFAS function associated with starting the turbine-driven auxiliary
feedwater pump (TDAFP) upon a loss of offsite power. Table 3.3-4 would
be revised to create separate functional units for the analog and
digital portions of the ESFAS function associated with starting the
TDAFP upon a loss of offsite power. Table 4.3-2 would be revised to
create separate functional units for the analog and digital portions of
the ESFAS function associated with starting the TDAFP upon a loss of
offsite 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.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The recognition that different
operability and surveillance requirements apply to analog vs.
digital circuitry does not impact any previously analyzed accidents.
The proposed change will not affect any of the analysis assumptions
for any of the accidents previously evaluated. The proposed change
does not alter the current method or procedures for meeting the
surveillance requirements in Table 4.3-2. The proposed change will
not affect the probability of any event initiators nor will the
proposed change affect the ability of any safety-related equipment
to perform its intended function. There will be no degradation in
the performance of nor an increase in the number of challenges
imposed on safety-related equipment assumed to function during an
accident situation. Therefore, the proposed 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.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. The separation of analog and digital portions of
Functional Unit 6.f will not impact the normal method of plant
operation.
The operability requirements, ACTION Statement, and surveillance
requirements for the analog portion, new Functional Unit 6.f.1), are
identical to those of Functional Unit 8.a. The requirements for the
digital portion, new Functional Unit 6.f.2), are consistent with the
current Technical Specifications, other than the new ACTION
Statement 39 provisions that eliminate the transient imposed on the
plant from a 3.0.3 shutdown and the performance of a refueling
interval TADOT [Trip Actuating Device Operational Test]. There is no
safety benefit associated with shutting the plant down under LCO
3.0.3, if both logic trains were inoperable, when considering the
fact that the pump is allowed to be inoperable for 72 hours. This
unnecessary shutdown would be detrimental to plant safety. The
``new'' TADOT requirement is a reflection of current plant testing
practice. These changes do not change any ESFAS design standards and
are appropriate for digital functions such as this. No new accident
scenarios, transient precursors, failure mechanisms, or limiting
single failures are introduced 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.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not affect the acceptance criteria for
any analyzed event. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection functions. There will be no impact
on any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: William H. Bateman.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri.
Date of application request: July 30, 1998.
Description of amendment request: The proposed amendment
application would change Table 4.3-2 of the technical specifications
(TS) by adding a table notation to clarify that verification of the
time delays associated with engineered safety feature actuation system
(ESFAS) Functional Units 8.a and 8.b, ``Loss of Power,'' is only
performed as part of the channel calibration.
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.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The protection systems will continue
to function in a manner consistent with the plant design basis. The
proposed change will not affect any of the analysis assumptions for
any of the accidents previously evaluated. Neither the Trip
Setpoints and Allowable Values in Technical Specification Table 3.3-
4 nor the response times listed in FSAR [Final Safety
[[Page 69349]]
Analysis Report] Table 16.3-2 are affected. The proposed change will
not affect the probability of any event initiators nor will the
proposed change affect the ability of any safety-related equipment
to perform its intended function. There will be no degradation in
the performance of nor an increase in the number of challenges
imposed on safety-related equipment assumed to function during an
accident situation. There will be no change to normal plant
operating parameters or accident mitigation capabilities. Therefore,
the proposed 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.
There are no hardware changes associated with this license
amendment nor are there any changes in the method by which any
safety-related plant system performs its safety function. The normal
manner of plant operation is unchanged. Verification of the time
delays need not be performed on a monthly basis when response time
testing is performed on an alternating 18 month basis per the
provisions of Technical Specifications 4.3.1.2 and 4.3.2.2 and the
verification of LOCA [loss-of-coolant accident] and shutdown
sequencer timing and analog channel time constant calibrations are
performed on a refueling frequency. No new accident scenarios,
transient precursors, failure mechanisms, or limiting single
failures are introduced as a result of this change. There will be no
adverse effect or challenges imposed on any safety-related system 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.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). There will be no effect on the manner in which safety
limits or limiting safety system settings are determined nor will
there be any effect on those plant systems necessary to assure the
accomplishment of protection functions. There will be no impact on
the overpower limit, DNBR [Departure from Nucleate Boiling Ratio]
limits, FQ, Nuclear Enthalpy Rise Hot Channel Factor,
LOCA PCT [Peak Clad Temperature], peak local power density, or any
other 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: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: William H. Bateman.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339.
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia.
Date of amendment request: November 18, 1998.
Description of amendment request: The proposed amendments would
make changes to the North Anna Power Station (NAPS), Unit 1 and 2,
Technical Specifications (TS) Surveillance Requirement (SR) 4.7.13.1,
``Groundwater Surveillance Requirements'' and related Table 3.7-6,
``Allowable Groundwater Levels--Service Water Reservoir.'' The change
in the SR requests that the measuring device numbers assigned to
piezometers be eliminated from the TS SR in order to avoid redundancy,
and eliminate confusion as well as the need to initiate TS changes
whenever new piezometers are added, older devices are replaced or
abandoned in-place. The proposed change in groundwater threshold levels
will raise the allowable groundwater levels to those consistent with
the allowable levels in the ``Stability of Service Water Reservoir
(SWR) Slope Under Increased Phreatic Surface'' calculations.
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, which is
presented below:
Specifically, operation of the North Anna Power Station in
accordance with the proposed TS Change Request will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, since: (a)
removing non-safety related SWR piezometer device numbers from the
TS and raising TS allowable groundwater surface threshold elevation
levels in the southeast section of the SWR will have no effect on
the way the safety-related Service Water System was designed to
operate, (b) Periodic Test Procedures will continue to identify all
open-tube piezometers and require that they be monitored in order to
obtain as much information as possible regarding changing
groundwater levels, (c) sufficient redundancy will continue to exist
since at least two (2) open-tube (standpipe-type) piezometers, not
subject to mechanical failure, have been installed in each of the
three (3) SWR zones to meet the TS Surveillance Requirement that
``at least one measurement per zone be available'' and (d) recent
calculations have confirmed that raising the allowable water level
in the southeast section of the SWR will not affect the stability of
the SWR dike as indicated in the original design basis calculation.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, since: (a) the frequency of
piezometer monitoring and the intent of monitoring groundwater
surface threshold elevations in order to maintain stability of the
SWR slope have not changed, (b) no physical modification to the
plant or new mode of plant operation is involved, (c) changes are
consistent with the assumptions made in the Safety Analyses and
original design basis calculation and (d) failure of the SWR dike
and ensuing loss of service water was the most serious accident
postulated and considered credible. Operation of the SWR is not
being changed. Therefore, a new or different kind of accident is
[not] created by the change in groundwater level. In addition, since
both the SWR and Lake Anna reservoir provide redundant sources of
service water, failure of the SWR is not considered as a credible
accident.
3. Involve a significant reduction in a margin safety, since:
(a) increasing the allowable phreatic surface in the SE section of
the SWR dike will not lower the factor of safety with respect to the
stability of the SWR as defined by the original design basis
calculation, (b) the margin to failure of the SWR dike has been
proven by calculation to have not been reduced as defined by the
original design basis calculation and (c) subject changes will not
impact the performance of structures, systems or components relied
upon for accident mitigation or any safety analysis assumptions,
therefore the margin of safety is not changed by the proposed
[change] in groundwater level at the SWR.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for Licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Project Director: Herbert N. Berkow.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the
[[Page 69350]]
Commission's rules and regulations. The Commission has made appropriate
findings as required by the Act and the Commission's rules and
regulations in 10 CFR Chapter I, which are set forth in the license
amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina.
Date of application for amendment: February 27, 1997, as
supplemented August 24, 1998.
Brief description of amendment: This amendment changes Technical
Specification (TS) 3/4.4.5, ``Steam Generators,'' by adding sleeve
installation as an alternative to tube plugging for repairing degraded
steam generators.
Date of issuance: November 23, 1998.
Effective date: November 23, 1998.
Amendment No.: 85.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17225).
The August 24, 1998, supplemental letter provided clarifying
information only, and 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 November 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket No. 50-254, Quad Cities Nuclear
Power Station, Unit 1, Rock Island County, Illinois.
Date of application for amendment: August 14, 1998, as supplemented
by letters dated October 13 and November 23, 1998.
Brief description of amendment: The amendment changes the Quad
Cities Technical Specifications (TS) to reflect the use of Siemens
Power Corporation ATRIUM-9B fuel. Specifically the amendment
incorporates the following into the TS: (a) new methodologies that will
enhance operational flexibility and reduce the likelihood of future
plant derates, (b) administrative changes that eliminate the cycle
specific implementation of ATRIUM-9B fuel and adopt Improved Standard
Technical Specification language where appropriate, and (c) changes to
the Minimum Critical Power Ratio.
Date of issuance: December 3, 1998.
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 182.
Facility Operating License No. DPR-29: The amendment revised the
TSs. Public comments requested as to proposed no significant hazards
consideration: Yes (63 FR 59588 dated November 4, 1998). This notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by December 4, 1998, but indicated that if the Commission makes
a final no significant hazards consideration determination any such
hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final no significant hazards consideration
determination are contained in a Safety Evaluation dated December 3,
1998.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida.
Date of application for amendment: October 31, 1997, as
supplemented December 13, 1997, February 27 and April 24, 1998.
Brief description of amendment: The amendment proposed to revise
the Final Safety Analysis Report (FSAR) to reflect changes to the
credited methodology for boron precipitation prevention, as approved by
the NRC.
Date of issuance: November 30, 1998.
Effective date: November 30, 1998.
Amendment No.: 171.
Facility Operating License No. DPR-72: Amendment revised the
Operating License to reflect the change to the FSAR.
Date of initial notice in Federal Register: November 12, 1997 (62
FR 60731). The supplemental letters contained clarifying information
that did not change the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 30, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida.
Date of application for amendment: October 29, 1998.
Brief description of amendment: The amendment revised the
terminology used in the St. Lucie Plant Technical Specifications (TS)
relative to the implementation and automatic removal of certain
protection system trip bypasses to ensure that the meaning of explicit
terms used in the TS are consistent with the intent of the stated
requirements.
Date of Issuance: November 24, 1998.
Effective Date: November 24, 1998.
Amendment No.: 98.
Facility Operating License No. NPF-16: Amendment revised the TS.
Date of initial notice in Federal Register: November 5, 1998 (63 FR
59809).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 24, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
[[Page 69351]]
Generating Station, Ocean County, New Jersey.
Date of application for amendment: July 21, 1998.
Brief description of amendment: The amendment (1) revises Technical
Specification (TS) 6.2.2.2(a) to provide flexibility to accommodate
unexpected absence of on-duty shift crew members, (2) eliminates
reference to the Manager, Plant Operations in Specification 6.2.2.2(j)
as the position has been eliminated, (3) reduces the maximum time in
which to forward audit reports to the responsible manager from 60 days
to 30 days, (4) replaces the term ``Vice President'' with the term
``Corporate Officer'' in several places in Section 6, and (5) corrects
several typographical errors.
Date of Issuance: November 30, 1998.
Effective date: November 30, 1998, to be implemented within 30 days
Amendment No: 203.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 26, 1998 (63 FR
45525).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated November 30, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
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: October 8, 1998.
Brief description of amendments: The amendments would revise the
Technical Specification Section 3.4.1.3, ``Reactor Coolant System--
Shutdown,'' and its associated bases to provide separate requirements
for the Reactor Coolant system in MODE 4, MODE 5 with the reactor
coolant loops filled, and MODE 5 with the reactor coolant loops not
filled.
Date of issuance: November 27, 1998.
Effective date: November 27, 1998, with full implementation within
30 days.
Amendment Nos.: 224 and 208.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1998 (63 FR
57322).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 27, 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. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York.
Date of application for amendment: June 19, 1998, as supplemented
November 6, 1998.
Brief description of amendment: This amendment changes Technical
Specification 3.2.2 and the associated Bases to update pressure-
temperature operating curves and tables for continued plant operation
up to 28 effective full-power years.
Date of issuance: November 25, 1998.
Effective date: As of the date of issuance to be implemented before
core operation exceeds 18 effective full-power years.
Amendment No.: 164.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40557)
The November 6, 1998, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
November 25, 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. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York.
Date of application for amendment: November 25, 1998, as
supplemented November 27, 1998.
Brief description of amendment: This change adds a note to certain
specific containment isolation valves listed in Table 4.4-1. The note
permits the licensee to operate Indian Point Unit 3 for the remainder
of the current cycle (Cycle 10) without pneumatic leakage rate testing
of these isolation valves. These valves have been leakage rate tested
in the past using water pressurized with nitrogen gas. Without this
emergency amendment, there would have had to delay its resumption of
plant operation at power until the Technical Specifications required
test was performed.
Date of issuance: November 27, 1998.
Effective date: As of the date of issuance to be implemented
immediately.
Amendment No.: 184.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications. The Commission's related evaluation of the
amendment, finding of emergency circumstances, and final determination
of no significant hazards consideration, are contained in a Safety
Evaluation dated November 27, 1998.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.
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: August 3, 1998, as supplemented
October 20, 1998.
Brief description of amendment: The amendment provides for
application of the existing minimum critical power ratio safety limit
to Cycle 14 operation.
Date of issuance: November 25, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 246.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48264).
The October 20, 1998, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 25, 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.
Southern California Edison Company, et al., Docket No. 50-362, San
Onofre Nuclear Generating Station, Unit No. 3, San Diego County,
California.
Date of application for amendment: September 22, 1998.
[[Page 69352]]
Brief description of amendment: The proposed amendment would modify
the Technical Specifications (TS) to change the parameter used to
establish and remove the bypasses for high reactor power trips. The
parameter would be changed from the current ``THERMAL POWER'' to
logarithmic power.
Date of issuance: November 23, 1998.
Effective date: November 23, 1998.
Amendment Nos.: 136.
Facility Operating License No. NPF-15: The amendments revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56259).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama.
Date of application for amendments: June 12 and August 14, 1998
(TS-390).
Brief description of amendments: Changes the technical
specifications (TS) to accommodate surveillance intervals to be
compatible with a 24-month fuel cycle.
Date of issuance: November 30, 1998.
Effective date: November 30, 1998.
Amendment Nos.: 235, 255, 215.
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the TS.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48269).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 30, 1998.
No significant hazards consideration comments received: None.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
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 21, 1996 (TS 96-03).
Brief description of amendments: The amendments revise the SQN
Technical Specification (TS) 3.7.1.3 to extend the limiting condition
for operation of the condensate storage tanks to Mode 4 when steam
generator is relied upon for heat removal.
Date of issuance: November 19, 1998.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: 238 and 228.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TSs.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52967).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 19, 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: April 30, 1998 (TS 98-01).
Brief description of amendments: The amendments revise the SQN
Technical Specification Surveillance Requirement 4.4.3.2.1.b by
changing the mode requirement to allow power-operated relief valve
stroke testing in Modes 3, 4, and 5 with a steam bubble in the
pressurizer rather than only in Mode 4.
Date of issuance: November 19, 1998.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: 239 and 229.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38204).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 19, 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 No. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee.
Date of application for amendment: May 6, as supplemented June 5,
1998.
Brief description of amendment: The requested changes would allow
an increase in the limit, up to 5.0 percent, for the U-235 enrichment
of new (unirradiated) fuel stored in the new fuel storage racks and
limit the fuel storage locations to assure that k-effective values are
met.
Date of issuance: December 1, 1998.
Effective date: December 1, 1998.
Amendment No.: 15.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43214).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 1, 1998.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 50-
440 Perry Nuclear Power. Plant, Unit 1, Lake County, Ohio.
Date of application for amendment: September 3, 1998.
Brief description of amendment: This amendment revised Technical
Specification 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,''
by increasing the Division 3 Diesel Generator fuel oil level
requirements to account for (1) a rounding error in the calculation,
and (2) the unusable volume due to vortex formation at the eductor
suction nozzle located in the fuel oil storage tank.
Date of issuance: November 23, 1998.
Effective date: November 23, 1998.
Amendment No.: 94.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53960).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 50-
440 Perry Nuclear Power Plant, Unit 1, Lake County, Ohio.
Date of application for amendment: August 28, 1997.
Brief description of amendment: This amendment revised Pressure-
[[Page 69353]]
Temperature (P/T) Limits contained in Technical Specification 3.4.11 as
a result of the Reactor Vessel Material Surveillance Program
Requirements contained in Appendix H of 10 CFR Part 50.
Date of issuance: December 2, 1998.
Effective date: December 2, 1998.
Amendment No.: 95.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 19, 1997 (62
FR 61846).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin.
Date of application for amendment: April 15, 1998 as supplemented
by letters dated August 13, 1998, September 28, 1998, and November 24,
1998.
Brief description of amendment: The amendment incorporates changes
to TS 2.1, ``Safety Limits'' and TS 3.10, ``Control Rod and Power
Distribution Limits.'' These changes revise the power distribution
peaking factor limits and limits operating parameters related to the
Minimum Departure from Nucleate Boiling Ratio (MDNBR) in support of
cycle 23 fuel and reload changes. A change associated with the fuel and
reload changes, is the removal, from the current licensing basis, of
the fuel pool turbine missile hazards analysis
Date of issuance: December 2, 1998.
Effective date: December 2, 1998.
Amendment No.: 142.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1998 (63FR25120
).
The supplemental submittals did not affect the initial
determination of no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
Dated at Rockville, Maryland, this 9th 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. 98-33206 Filed 12-15-98; 8:45 am]
BILLING CODE 7590-01-P