[Federal Register Volume 63, Number 145 (Wednesday, July 29, 1998)]
[Notices]
[Pages 40551-40567]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-20111]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of
1954, as amended (the Act), to require the Commission to publish notice
of any amendments issued, or proposed to be issued, under a new
provision of section 189 of the Act. This provision grants the
Commission the authority to issue and make immediately effective any
amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 3, 1998, through July 17, 1998. The
last biweekly notice was published on July 15, 1998 (63 FR 38198).
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
[[Page 40552]]
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 August 14, 1998, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the
[[Page 40553]]
Atomic Safety and Licensing Board that the petition and/or request
should be granted based upon a balancing of factors specified in 10 CFR
2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 27, 1997, as supplemented by letters
dated March 9, March 20, April 20, May 27, and June 24, 1998
Description of amendment request: The proposed amendments would
revise the current Technical Specifications (TS) of each unit to
conform with NUREG-1431, Revision 1, ``Standard Technical
Specifications--Westinghouse Plants.'' The staff had previously issued
a Notice of Consideration of Issuance of Amendments published in the
Federal Register on July 14, 1997 (62 FR 37628) covering all the
proposed changes that were indeed within the scope of NUREG-1431. The
staff subsequently published two Notices of Consideration of Issuance
of Amendments and Proposed No Significant Hazards Determination (63 FR
25106, dated May 6, 1998; 63 FR 27760 dated May 20, 1998) to cover
DEC's March 9, March 20, April 20, and May 27, 1998, supplements, which
proposed changes that are beyond the scope of NUREG-1431. On June 24,
1998, DEC identified additional beyond-scope changes. The following
descriptions and proposed no significant hazard analyses cover only
those beyond-scope changes. Associated with each change are
administrative/editorial changes such that the new or revised
requirements would fit into the format of NUREG-1431.
1. Current TS 4.8.1.1.2.f specifies that the fuel for the
emergency diesel generators (EDGs) be periodically sampled for
particulate contamination strictly in accordance with the industry
standard ASTM-D2276-78. DEC proposed to relax this requirement,
adopting only the guidance of the standard, but using a larger
particulate filter for sampling (change from 0.8-to 3-micron). The
revised requirement would show up as TS 5.5.13.c of the Improved TS.
No changes to the design and functions of the EDGs are proposed.
2. DEC proposed to revise current TS Table 4.3-1, Functions 16 and
17. The revised requirements, to show up as Table 3.3.1-1, Functions 15
and 16.b, of the Improved TS, would add an actuation logic test
surveillance for the reactor trip system interlocks and the safety
injection input from the engineered safety feature actuation system. No
changes to the design and functions of these systems are involved.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), DEC has provided its
analyses of the issue of no significant hazards consideration for each
of the above proposed changes. The NRC staff has reviewed DEC's
analyses against the standards of 10 CFR 50.92(c). The NRC staff's
analysis is presented below.
1. Will the changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
For all the changes the answer is ``no.'' The proposed changes will
not affect the safety function of the subject systems. There will be no
direct effect on the design or operation of any plant structures,
systems, or components. No previously analyzed accidents were initiated
by the functions of these systems, and the systems will continue to
perform their functions in mitigating consequences of previously
analyzed accidents. Therefore, the proposed changes will have no impact
on the consequences or probabilities of any previously evaluated
accidents.
2. Will the changes create the possibility of a new or different
kind of accident from any accident previously evaluated?
For all the changes the answer is ``no.'' The proposed changes
would not lead to any design or operating procedure change. Hence, no
new equipment failure modes or accidents from those previously
evaluated will be created.
3. Will the changes involve a significant reduction in a margin of
safety?
For all the changes the answer is ``no.'' Margin of safety is
associated with confidence in the design and operation of the plant.
The proposed changes to the TS do not involve any change to plant
design, operation, or analysis. Thus, the margin of safety previously
analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for each of the proposed changes. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Project Director: Herbert N. Berkow.
Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 15, 1997, as supplemented by
letters dated March 5, April 27, and June 15, 1998.
Description of amendment request: The staff had previously
published a Notice of Consideration of Amendments and Proposed No
Significant Hazards Consideration Determination on the licensee's
September 15, 1997, application in the Federal Register on October 8,
1997 (62 FR 52580). As a result of the staff's requests for additional
information, DEC expanded its original amendment application by letter
dated June 15, 1998. Specifically, the June 15, 1998, letter proposes
requirements regarding the Low Temperature Overpressure Protection
System to be added to the units' Technical Specifications. There is,
however, no change to plant design.
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, addressing the three standards of 10 CFR 50.92(c):
First Standard
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The Low Temperature Overpressure Protection
System is not an accident initiating system; it is an accident
mitigating system. Therefore, the addition of supplemental Technical
Specification required controls pertaining to this system cannot
impact accident initiating probabilities. The Low Temperature
Overpressure Protection System will remain fully capable of
performing its design accident mitigation function for the modes in
which it is required. Therefore, no accident consequences will be
impacted.
Second Standard
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. As noted previously, the Low Temperature
Overpressure Protection System is not an accident initiating system.
The addition of the supplemental Technical Specification
[[Page 40554]]
controls pertaining to this system as specified will not impact any
plant systems that are accident initiators. No other modifications
are being proposed to the plant which would result in the creation
of new accident mechanisms.
Third Standard
Implementation of this amendment 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. The performance of the fission
product barriers will not be impacted by implementation of this
proposed amendment supplement. The Low Temperature Overpressure
Protection System will remain fully capable of performing its design
function for the modes in which it is required. Therefore, no safety
margin will be significantly impacted.
The staff reviewed the licensee's analysis, and agrees that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Project Director: Herbert N. Berkow.
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: May 8, 1998.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) for the Power Range Neutron
Flux High Trip setpoints in the event of inoperable main steam safety
valves. The licensee has determined that the new values are more
conservative than the values in the current TS. Also, the proposed
changes would delete the references to the 3-loop operation. The
proposed changes are consistent with the proposed Improved Standard TS
submitted by the licensee on May 27, 1997.
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 the
proposed amendment involve an increase in the probability or
consequences of an accident previously evaluated?
The proposed amendment involves a reduction in the maximum
allowable power range neutron flux high setpoints in case of
inoperable main steam safety valves. All applicable UFSAR [Updated
Final Safety Analysis Report] Chapter 15 transient acceptance
criteria are met with the proposed change. Therefore, operation of
the facility in accordance with the proposed amendment will not
involve an increase in the probability or consequences of an
accident previously evaluated.
2. Will operation of the facility in accordance with the
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated:
No new equipment or operating practice is involved with this
proposed amendment. No alteration to any existing hardware is
involved with this proposed amendment. Power Range high neutron flux
setpoint calibration is continued to be performed by the same
approved procedure. Therefore, operation of the facility in
accordance with the proposed amendment will not create the
possibility of any new or different kind of accident from any
accident previously evaluated.
3. Will operation of the facility in accordance with the
proposed amendment involve a reduction in a margin of safety?
The proposed change is in a more-conservative direction. All
applicable UFSAR Chapter 15 transient acceptance criteria are met
with the proposed amendment. Therefore, operation of the facility in
accordance with the proposed amendment will not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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.
Duke Energy Corporation (DEC), et al., Docket Nos. 50-369 and 50-370,
McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North
Carolina
Date of amendment request: May 27, 1997, as supplemented by letters
dated March 9, March 20, April 20, May 27, June 3, June 24, and July 7,
1998.
Description of amendment request: The proposed amendments would
revise the current Technical Specifications (TS) of each unit to
conform with NUREG-1431, Revision 1, ``Standard Technical
Specifications--Westinghouse Plants.'' The staff had previously issued
a Notice of Consideration of Issuance of Amendments published in the
Federal Register on July 15, 1997 (62 FR 37940) covering all the
proposed changes that were indeed within the scope of NUREG-1431. The
staff subsequently published additional Notices of Consideration of
Issuance of Amendments and Proposed No Significant Hazards
Determination on May 6, 1998 (63 FR 25107 and 63 FR 25108 (two
notices)) and on May 20, 1998 (63 FR 27761) to cover DEC's March 9,
March 20, April 20, and May 27, 1998, supplements, which proposed
changes that are beyond the scope of NUREG-1431.
On June 24, 1998, DEC identified additional beyond-scope changes.
The following descriptions and proposed no significant hazard analyses
cover only those beyond-scope changes. Associated with each change are
administrative/editorial changes such that the new or revised
requirements would fit into the format of NUREG-1431.
1. Current TS 4.8.1.1.2.f specifies that the fuel for the emergency
diesel generators (EDGs) be periodically sampled for particulate
contamination in accordance with ASTM-D2276-78. DEC proposed to relax
this requirement, adopting instead the guidance of ASTM-D2276, Method
A. The revised requirement would show up as TS 5.5.13.c of the Improved
TS. No changes to the design and functions of the EDGs are proposed.
2. DEC proposed to change the required action due to inoperable
channels of the containment pressure control system as currently
contained in Table 3.3-3, Item 7. The revised requirement would show up
as Action Item 16b in Table 3.3.2-1 of the Improved TS. No changes to
the design and functions of the containment pressure control system are
involved.
3. DEC proposed to revise current TS Table 4.3-1, Functions 16 and
17. The revised requirements, to show up as Table 3.3.1-1 Functions 15
and 16.b, would add an actuation logic test surveillance for the
reactor trip system interlocks and the safety injection input from the
engineered safety feature actuation system. No changes to the design
and functions of these systems are involved.
Basis for proposed no significant hazards consideration
determination:
[[Page 40555]]
As required by 10 CFR 50.91(a), DEC has provided its analyses of the
issue of no significant hazards consideration for each of the above
proposed changes. The NRC staff has reviewed DEC's analyses against the
standards of 10 CFR 50.92(c). The NRC staff's analysis is presented
below.
1. Will the changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
For all the changes the answer is ``no.'' The proposed changes will
not affect the safety function of the subject systems. There will be no
direct effect on the design or operation of any plant structures,
systems, or components. No previously analyzed accidents were initiated
by the functions of these systems, and the systems will continue to
perform their functions in mitigating consequences of previously
analyzed accidents. Therefore, the proposed changes will have no impact
on the consequences or probabilities of any previously evaluated
accidents.
2. Will the changes create the possibility of a new or different
kind of accident from any accident previously evaluated?
For all the changes the answer is ``no.'' The proposed changes
would not lead to any hardware or operating procedure change. Hence, no
new equipment failure modes or accidents from those previously
evaluated will be created.
3. Will the changes involve a significant reduction in a margin of
safety?
For all the changes the answer is ``no.'' Margin of safety is
associated with confidence in the design and operation of the plant.
The proposed changes to the TS do not involve any change to plant
design, operation, or analysis. Thus, the margin of safety previously
analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for each of the proposed changes. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Project Director: Herbert N. Berkow.
Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station,
Unit 3, Oconee County, South Carolina
Date of amendment request: July 16, 1998.
Description of amendment request: The proposed change would extend,
on a one-time basis, certain specified Technical Specification
surveillances that are required to be performed at a frequency of 18
months from the maximum allowed frequency of 22 months, 15 days, to a
maximum of 24 months. The following surveillances are involved: (a)
Standby Shutdown Facility (SSF) Reactor Coolant System (RCS) Pressure
Instrument Calibration; (b) SSF RCS Pressurizer Level Instrument
Calibration; (c) SSF RCS Makeup Pump Flow Instrument Calibration; (d)
Reactor Protective System (RPS) RCS Flow Instrument Calibration; (e)
RPS RCS Pressure Instrument Calibration; and (f) Low Pressure Injection
System Pump Discharge Valves LP-17 and LP-18 Manual Cycle.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
This proposed change has been evaluated against the standards in
10 CFR 50.92 and has been determined to involve no significant
hazards, in 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?
No. A review of the previous two instrument channel tests and
calibrations, and two manual valve cycle tests discussed in this
amendment request concluded that no adverse effects should occur as
a result of the one-time extension.
There is a high level of confidence that the instruments and
valves should be available to perform their intended function during
the requested extension period. Thus, the probability and
consequences of an accident previously evaluated will not be
significantly increased.
(2) Create the possibility of a new or different kind of
accident from the accidents previously evaluated?
No. Since the one-time extension should not cause any adverse
effects on Standby Shutdown Facility, Reactor Protective System or
the Low Pressure Injection system, a new or different kind of
accident from the accidents which were previously evaluated will not
occur. The Standby Shutdown Facility, Reactor Protective System or
the Low Pressure Injection system should be available to perform
their intended function during the requested extension period.
(3) Involve a significant reduction in a margin of safety?
No. The margin of safety will not be significantly reduced by
this amendment request because the Standby Shutdown Facility,
Reactor Protective System or the Low Pressure Injection system
should be available to perform their intended function during the
requested extension period. In addition, the review of the previous
tests and calibrations which are discussed in the amendment request
concluded that no adverse effects should occur as a result of the
one-time extension.
Duke [Energy Corporation] has concluded, based on the above
information, that there are no significant hazards involved in this
amendment request.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC.
NRC Project Director: Herbert N. Berkow.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: April 28, 1998.
Description of amendment request: The proposed amendment would
change the scope and frequency of volumetric and surface inspections
for the reactor coolant pump motor flywheels. The current prescribed
frequency and scope are contained in U.S. NRC Regulatory Guide 1.14,
Regulatory Positions C.4.b.1 and C.4.b.2. The proposed revision
reflects the frequency and scope of volumetric and surface
examinations, which has been reviewed and approved by the NRC, as
stated in the Safety Evaluation for Topical Report WCAP-14535A.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The CR-3 [Crystal River Unit 3] components addressed by this
proposed change are the Reactor Coolant Pumps (RCPs), identified by
plant tagging procedures as RCP-1A, RCP-1B, RCP-1C,
[[Page 40556]]
and RCP-1D. The RCPs are vertical, single stage, single suction,
shaft seal, centrifugal pumps. The RCPs ensure that adequate cooling
water is circulated through the reactor coolant system. Following
loss of power to the RCP motor, the flywheel, in conjunction with
the impeller and motor rotating assembly, provide sufficient
rotational inertia to assure adequate coolant flow during RCP
coastdown, thus providing adequate core cooling. The maximum loading
on the RCP motor flywheel results from overspeed following a large
loss of coolant accident (LOCA). The estimated maximum speed in the
event of a LOCA was established conservatively. The proposed change
does not affect that analysis. Reduced coastdown times due to a
single failed flywheel is bounded by the locked rotor analysis,
therefore it will not place the plant in an unanalyzed condition.
Reducing the frequency of inspection, as proposed, will not
significantly increase the probability of an accident previously
evaluated. CR-3 is not specifically analyzed for a flywheel failure
accident. The design, fabrication, and testing of the flywheels in
accordance with the guidance found in Regulatory Guide 1.14
minimizes the potential for flywheel failure. Nevertheless, the
topical report indicates that the flywheels could be operated for
forty years without inspection, and there would be no significant
increase in the probability of failure of the flywheel. However,
inspections are proposed to continue at a frequency of once every
ten years as a conservative measure. Therefore, these changes do 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.
The purpose of the RCP motor flywheel inspection is to identify
flaws that could lead to failure of the flywheel. The design,
fabrication, and testing of the flywheels in accordance with the
guidance found in Regulatory Guide 1.14 minimizes the potential for
flywheel failure. No new failure mode is introduced due to the
change in flywheel inspection frequency since the proposed changes
do not involve the addition or modification of equipment, nor alter
the design or operation of affected plant systems, structures or
components. Therefore, these changes do not create a possibility of
a new or different kind of accident from any previously evaluated.
(3) Involve a significant reduction in a margin of safety.
As shown in the topical report, RCP motor flywheels have been
inspected for twenty years without any service induced flaws being
identified. Additionally, the analyses demonstrated that the
flywheels are manufactured from excellent quality steel, have a high
fracture toughness, and have a very high flaw tolerance. The topical
report indicates that the flywheels could be operated for forty
years without inspection, and there would be no significant increase
in the probability of failure of the flywheels. However, inspections
are proposed to continue at a frequency of once every ten years as a
conservative measure. The non-destructive examination acceptance
criteria is not changing as a result of the proposed LAR. Thus, the
margin of safety is not reduced significantly by the proposed change
in inspection frequency.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Project Director: Frederick J. Hebdon.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: May 27, 1998.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TS) to remove the requirement for
safety injection tanks (SITs) to be operable in reactor operational
Mode 4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment does not involve changes to previously
evaluated accident initiators. The proposed TS changes related to
removal of the requirement for safety injection tanks to be operable
in MODE 4 do not impact the results of existing accident analyses,
and have no adverse impact on any plant system performance.
The function of each SIT is to provide early reactor core
reflood in the event of a LBLOCA [large break loss-of-coolant
accident]. Safety injection tanks are not required for mitigating
the consequences of large RCS pipe ruptures in MODE 4, and the
proposed change to TS 3.5.1 will delete the requirement for SIT
operability when in this mode. Due to the reduced initial stored
energy and decay heat generation rate consistent with operation in
the shutdown modes, the required operable HPSI [high-pressure safety
injection] pump is sufficient to perform the function of reactor
vessel reflood and coolant inventory make-up. Therefore, operation
of the facility in accordance with the proposed amendment will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. The changes do
not involve the addition of new equipment or the modification of
existing equipment, nor do they alter the design of St. Lucie plant
systems described in the Updated Final Safety Analysis Report
(UFSAR). There are no adverse effects on any system performance due
to the proposed TS changes, and the plant configuration will
continue to remain consistent with assumptions used in the existing
accident analyses. Therefore, operation of the facility in
accordance with the proposed amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed TS changes have been evaluated with respect to the
applicable safety analyses. FPL [Florida Power and Light Co.]
determined from this new evaluation that safety injection tanks are
not required to prevent core uncovery during a loss of coolant
accident initiated in MODE 4. Due to the reduced core heat removal
requirements in this lower mode and in the absence of substantial
core uncovery, fuel cladding temperatures and clad oxidation will
remain at low levels, long term cooling will be maintained, and 10
CFR 50.46 acceptance criteria will be satisfied. Therefore,
operation of the facility in accordance with the proposed amendment
would not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: June 29, 1998.
Description of amendment request: This Technical Specification
change
[[Page 40557]]
request replaces in its entirety, a previously submitted request dated
February 22, 1996, and published in the Federal Register on March 27,
1996 (61 FR 13525). This request greatly reduces the scope of the
previous request. It retains the provision to delete the requirement
that the biennial inspection of the Emergency Diesel Generators (EDGs)
be performed during shutdown, permits skipping diesel starting battery
capacity test for recently installed batteries, and increases the
minimum loading during diesel testing from 20% to 80%. In addition,
there are wording changes to enhance clarity, and a typographical error
is corrected.
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. State the basis for the determination that the proposed
activity will or will not increase the probability of occurrence or
the consequences of an accident.
The proposed activity deletes the requirement to inspect EDGs
during shut down from the Technical Specifications and permits
skipping diesel starting battery capacity tests of recently
installed batteries. The minimum loading during the testing of the
diesels has been increased from 20% to 80%. In addition, wording
changes were made to enhance clarity and a minor typographical error
was corrected. During reactor operations other power sources are
available to compensate for one diesel being out of service. The
inspections and testing will continue to be done with the same
intervals and the 80% loading is a more stringent requirement.
Therefore, these changes do not affect the design or performance of
the EDGs or their ability to perform their design function.
2. State the basis for the determination that the activity does
or does not create a possibility of an accident or malfunction of a
different type than any previously identified in the [safety
analysis report] SAR.
The EDGs are not the source of any accident described in the
SAR. These changes do not modify the design or performance of the
EDGs and do not affect plant functions or actions. Current
specifications permit one diesel generator to be inoperable for up
to 7 days and this change will not impact that time frame.
Therefore, the proposed change does not create the possibility of an
accident or malfunction of a different type than those previously
identified.
3. State the basis for the determination that the margin of
safety is not reduced.
The proposed changes are designed to improve EDG reliability and
availability during shutdown periods by providing flexibility in the
scheduling and performance of maintenance. The surveillance
intervals are unchanged and operability requirements are not
modified. The proposed activity does not alter the basis of any
technical specification that is related to the establishment or
maintenance of a nuclear safety margin. Therefore, the margin of
safety is not significantly reduced by this action.
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.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station, Unit 1 (NMP1), Oswego County, New York
Date of amendment request: June 19, 1998.
Description of amendment request: The proposed amendment would
update Technical Specification (TS) 3.2.2, ``Minimum Reactor Vessel
Temperature for Pressurization,'' and the associated TS Bases pages. TS
3.2.2 contains tables and figures that limit the minimum reactor vessel
temperature for a given pressure. The limits are based upon the number
of Effective Full Power Years (EFPY) of core operation. The current
tables and figures are valid for up to 18 EFPYs of core operation. The
proposed amendment will substitute new tables and figures that are
valid for 20, 24 and 28 EFPYs. The word ``leakage'' would be added to
clarify that this TS applies to both leakage and hydrostatic tests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The changes to the P-T [pressure and temperature] curves are
being proposed to preclude brittle fracture of RPV [reactor pressure
vessel] materials for up to 28 EFPYs. In addition to the leakage/
hydrostatic test curve for 28 EFPYs, leakage/hydrostatic test curves
have been prepared for exposures up to 20 EFPYs and up to 24 EFPYs
to shorten outage time for startups conducted prior to these
exposures. Safety margins specified in 10 CFR Part 50, Appendix G
and Appendix G to Section III of the ASME [American Society of
Mechanical Engineers] Code will continue to be met for each of these
curves. Also, the proposed changes do not affect the probability of
any accident precursors. Therefore, operation in accordance with the
proposed change will not involve a significant increase in the
probability of an accident previously evaluated.
The RPV, as part of the reactor coolant system, provides a
barrier to the release of reactor coolant and subsequent
radiological consequences. Operation in accordance with the proposed
amendment will preclude brittle fracture of the RPV consistent with
current requirements, and consequently, not affect the consequences
of any accidents. Therefore, operation of NMP1 [Nine Mile Point Unit
1] in accordance with the proposed amendment will not involve a
significant increase in the consequences of an accident previously
evaluated.
2. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does not involve any physical alterations to
plant configurations or introduce any new accident precursors which
could initiate a new or different kind of accident. The proposed
change does not affect the intended function of the RPV nor does it
affect the operation of the RPV in a way which would create a new or
different kind of accident. The changes to the P-T curves are being
proposed to preclude brittle fracture of RPV materials for up to 28
EFPYs. Safety margins specified in 10 CFR Part 50, Appendix G and
Appendix G to Section III of the ASME Code will continue to be met.
Therefore, operation of NMPI in accordance with the proposed change
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not involve a significant reduction in
a margin of safety.
The existing NMP1 P-T curves were developed using safety margins
for brittle fracture found in 10 CFR PART 50 Appendix G and Appendix
G to Section III of the ASME Code. The proposed NMP1 P-T operation
curves, which are valid for up to 28 EFPYs of operation, were also
developed using the safety margins for brittle fracture found in 10
CFR PART 50, Appendix G and Appendix G to Section III of the ASME
Code. Accordingly, operation of NMPI in accordance with the revised
P-T operating limits will continue to preclude brittle fracture of
the RPV materials during plant heatup, cooldown, and leakage/
hydrostatic test conditions with the same margin of safety that
currently exists. Therefore, operation of NMP1 in accordance with
the proposed amendment will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 40558]]
standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes
to determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: S. Singh Bajwa.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: May 22, 1997, as supplemented by letters
dated June 12, 1997, August 28, 1997 and January 29, 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 TS 3/4.7.3.1, ``Plant Systems--Vital
Component Cooling Water System,'' to add new action statements and
surveillance requirements for the component cooling water (CCW) surge
tank pressurization system. CCW surge tank pressurization system
requirements currently exist in an equipment control guideline, but are
proposed for inclusion in TS because the CCW surge tank pressurization
system is required to support licensing basis assumptions for a design
basis loss-of-coolant accident.
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 component cooling water (CCW) surge tank pressurization
system is designed to mitigate the consequences of an accident, and
cannot initiate an accident.
The proposed changes to the Technical Specifications (TS)
incorporate requirements for the CCW surge tank pressurization
system to assure that the consequences of an accident are not
increased. The CCW surge tank pressurization system was installed to
restore the component cooling water system to its original design
and licensing basis. The design of the CCW surge tank pressurization
system ensures that a minimum pressure of 17 psig is maintained in
the surge tank at the initiation of a design basis loss of coolant
accident. This minimum pressure is sufficient to ensure that boiling
will not occur in the containment fan cooler units (CFCUs), assuming
the worst case accident conditions with a concurrent loss of offsite
power (LOOP).
Therefore, the addition of these new requirements does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The CCW surge tank pressurization system is designed to mitigate
the consequences of an accident, and cannot initiate an accident.
The proposed TS changes incorporate requirements for the CCW
surge tank pressurization system. Installation of the CCW surge tank
pressurization system provides assurance that boiling in the CFCUs
will not occur, assuming the worst case accident, with a concurrent
LOOP.
Therefore, addition of these requirements does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes to the TS incorporate requirements for the
CCW surge tank pressurization system to assure that the consequences
of an accident are not increased. The design of the CCW surge tank
pressurization system ensures that a minimum pressure of 17 psig is
maintained in the surge tank at the initiation of a design basis
accident. The minimum pressure is sufficient to ensure that boiling
will not occur in the CFCUs, assuming the worst case accident
conditions with a concurrent LOOP.
Therefore, the proposed changes do not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: March 20, 1998, as revised by
letter dated June 26, 1998.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to permit incorporation of an
End-of-Cycle Recirculation Pump Trip.(EOC-RPT) System at Peach Bottom
Atomic Power Station (PBAPS), Units 2 and 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The addition of the EOC-RPT System will not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The EOC-RPT System has been designed
to appropriate standards and specifications to ensure that the
ability of the plant to mitigate the effects of accidents is
maintained. Each division is electrically, mechanically, and
physically independent to meet the single failure criterion.
The EOC-RPT System will improve the reactor core thermal
response following a turbine trip transient caused by either a
turbine control valve fast closure or a turbine stop valve closure.
The EOC-RPT will be relied upon to reduce the fuel thermal
mechanical transient excursion such that fuel thermal limits are not
violated. Under conditions when the system is inoperable, more
conservative thermal limits will be enforced.
The new system will utilize existing RPS [Reactor Protection
System] logic to initiate the Reactor Recirculation System (RRS)
pump trips on a turbine generator trip and a generator load
rejection event. The inputs to RPS used by EOC-RPT will be from
turbine stop valve (TSV) limit switches and turbine control valve
(TCV) oil pressure switches. There will be no direct interface
between the EOC-RPT System and the main turbine control system. Thus
the new system can not initiate a turbine trip or generator load
rejection event. This change does not result in significant increase
in the probability of events described in the UFSAR [Updated Final
Safety Analysis Report]. Additionally, the probability of
inadvertent single or dual recirculation pump trips due to the
addition of the EOC-RPT components will not be significantly
increased by this modification.
No new challenges to the reactor coolant pressure boundary will
result from the incorporation of the EOC-RPT System which
[[Page 40559]]
could result in a significant increase in the consequences of an
accident.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The ECO-RPT System has been designed to appropriate standards
and specifications to ensure that no new sequence of events or
failure modes will occur such that a transient event will escalate
into a new or different type of accident.
The PBAPS UFSAR evaluates several recirculation pump trip
events, including the limiting case of a pump seizure. A spurious
dual EOC-RPT pump trip is similar to other RRS pump trip events
evaluated in the UFSAR and does not represent a different type of
accident.
Additionally, this modification will not create any new failure
mode or sequences of events that could lead to a different type of
accident than previously evaluated. The new EOC-RPT System will not
involve any new challenges to a fission product barrier. The EOC-RPT
System does not make any changes to the design function of the RRS.
Therefore, the new equipment installed by this modification cannot
create the possibility of a different type than previously evaluated
in the SAR [Safety Analysis Report].
The EOC-RPT System is classified as important-to-safety. Failure
or malfunction of the new equipment will not prevent or affect the
ability of safety-related or important-to-safety systems to respond
to the design basis accidents described in the FSAR [Final Safety
Analysis Report].
There will be no software used in the EOC-RPT System. The system
logic consists of two electrically and physically separated trip
systems; one will be used to trip one EOC-RPT System breaker, and
the other will be used to trip the second EOC-RPT System breaker for
each pump.
The design of this modification assures that the new system is
not susceptible to electromagnetic (EM) emissions and will not cause
inadvertent operation of existing plant equipment due to EM
emissions.
Based on the previous discussion, the possibility of a new or
different kind of accident from any accident previously evaluated
will not be created.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
There are no significant reductions in any margin of safety
previously approved by the USNRC as a result of this change to the
TS. The EOC-RPT System will ensure that fuel thermal limits are not
exceeded during the limiting transient. In the event that the EOC-
RPT System is determined to be inoperable, specific operating limits
are provided in the COLR. In all cases, thermal limits are not
exceeded and the margin of safety is not significantly reduced.
The plant LOCA response will not change for present core
configurations (i.e., 9 x 9 fuel) with the EOC-RPT System installed.
For GE 8 x 8 fuel, which could be used at a future time, there could
be a small increase in Peak Cladding Temperature (PCT). This
increase would still be well below the 2200 deg. F acceptance limit
defined in 10 CFR 50.46.
There will be no significant reduction in the margin of safety
as previously approved by the USNRC, since the calculated increase
in peak cladding temperature for a core containing limit 8x8 fuel
design (BP/P8 x 8R) is a small increase above the previously
analyzed peak cladding temperature. Additionally, this modification
does not impact the safety function of the RRS piping, thus reactor
coolant pressure boundary safety limits are not affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Project Director: Robert A. Capra.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, PeachBottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: May 1, 1998
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to delete requirements for the
functional testing of the safety relief valves (SRVs) during each unit
startup at Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed Technical Specification changes to the requirement
for functional testing of the SRVs during each unit startup will not
significantly increase the probability or consequences of an
accident previously evaluated. Elimination of the functional test
will not prevent the SRVs from performing their intended safety
function. The proposed change to delete the SRV functional test at
power should delete a potential initiator of SRV leakage. The
remaining testing and inspections will continue to adequately
demonstrate the operability of the SRVs for both the safety and
depressurization modes.
As a result of deleting the requirement for functional testing
of the SRVs during each unit startup and replacing these
requirements with the proposed tests contained TS SR [surveillance
requirement] 3.4.3.2 and 3.5.1.12, the only change in the frequency
of testing of the SRV components is that the main valve disc of the
SRVs will be tested every two cycles (approximately four years) as
compared to the current one cycle (approximately two years)
frequency. As described above, the lift test of the main valve disc
is currently performed at an offsite facility. A review of offsite
testing data for the years 1987 through 1998 was performed for the
PBAPS, Units 2 and 3 SRVs. Since the design of the SRVs is to ensure
operation of the overpressurization protection and the ADS
[Automatic Depressurization System] function is to reduce reactor
pressure during a small break LOCA [loss-of-coolant accident], the
review consisted of looking for any failures of the main valve disc
to stroke open during setpoint actuation. This review consisted of
reviewing ``as-found'' test data since any failures following a
rebuild would be found during the final certification testing. Based
on a review of as-found data, it was concluded that there were no
reported cases of the main disc failing to open during setpoint
pressure testing. Therefore, deleting the requirement for functional
testing of the SRVs during each unit startup is not expected to
negatively impact these test results.
Therefore, eliminating the functional test is not expected to
negatively impact these test results or involve a significant
increase in the probability of an accident previously evaluated.
As discussed in the PBAPS, Units 2 and 3 Updated Final Safety
Analyses Report (UFSAR), analyzed events resulting in a nuclear
system pressure increase, such as MSIV [main steam line isolation
valve] closure, generator load rejection, turbine trip, failure of
the turbine bypass valves to open, and loss of main condenser
vacuum, take credit for the SRVs opening to mitigate the
consequences of these events. The proposed changes will not increase
the consequences of these events, since a series of remaining tests
will ensure all SRV components will function. The SRVs will
therefore be capable of performing their design functions.
SRV second stage valve leakage can be increased as a result of
corrosion/debris introduced on the seating area surface. Second
stage leakage, if allowed to continually increase, will eventually
start to depressurize the volume above the SRV main valve piston to
the extent that sufficient differential pressure will lift the main
valve disc. Reactor vessel coolant inventory decrease due to an
inadvertent opening of a Safety Relief Valve is an abnormal
operating transient event. This event can be a precursor to fuel
failure due to gradual loss of coolant, and the mitigation is
similar to the small break LOCA. Under the proposed change, it is
expected that the probability of SRV leakage will decrease, thus the
probability of occurrences of an inadvertent SRV actuation is
reduced, therefore reducing the probability or
[[Page 40560]]
consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The SRVs will not be operated or tested in a manner contrary to
their design. As a result, no new mode of operation is introduced.
Therefore, the revised testing will not create a new failure mode of
the SRVs which could create the possibility of a new or different
kind of accident from any previously evaluated. Since other tests,
taken together, confirm the entire SRV assembly functions
adequately, this proposed change is justified. The proposed change
to delete the SRV functional test at power will not impact the
ability of the SRV to open and provide their intended safety
function.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
By removing the Technical Specification requirements to perform
the in-situ functional testing during startup, the probability of
inadvertently opening of a SRV should be reduced through the
elimination of a potential initiator of SRV second stage disc
leakage and subsequent erosion. This Technical Specification change
will aid in decreasing SRV leakage and improve SRV reliability at
power operations. Eliminating the SRV in-situ functional test during
startup will increase the margin of safety during operations,
transients, or accidents. Remaining surveillance testing and
inspections assure each component necessary for successful opening
of the SRV function properly as designed.
Removal of the functional test will not negatively impact the
Technical Specifications lift setpoints of the SRVs necessary for
the function of the safety mode. The functional test does not
completely test the safety mode of the SRV which is based on the
Technical Specifications lift setpoints.
Offsite testing at operating steam pressure ensures the
operability of the SRV pilot, second stage, and main valve function.
The valves are refurbished and post maintenance testing is performed
at a steam pressure of 1040 psig. Upon successful test completion,
the valve receives written certification from the lab and is
returned to PBAPS for reinstallation. To receive certification, the
valve must have zero main seat leakage and meet the acceptance
criteria for setpoint pressure. These tests satisfy the requirements
of the PBAPS IST [Inservice Testing] Program and TS. The tests
contained in the proposed TS SR 3.4.3.2 and 3.5.1.12 will verify the
operation of the solenoid and second stage disc movement of all 11
SRVs in the depressurization mode.
The remaining segments of the SRV tests verify the ability of
the SRV logic. In summary, this change will not involve a
significant reduction in the margin of safety, because of the
reduction in SRV degradation, and the remaining tests confirm the
valves will function properly when required.
The NRC staff has reviewed the licensee's analysis and, based on
this review, this 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 Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Project Director: Robert A. Capra.
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 amendment request: June 16, 1998.
Description of amendment request: The proposed change would
relocate the Safety Review Committee (SRC) review, audit, and related
record keeping requirements from the Technical Specifications (TSs) to
Chapter 17 of the Final Safety Analysis Report (FSAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response: This amendment application does not involve a
significant increase in the probability or consequences of an
accident previously analyzed. The relocation of the SRC review,
audit, and related record keeping requirements from the TS to the
FSAR does not alter the performance or frequency of these
activities. Future changes to the QA [Quality Assurance] program,
located in Chapter 17 of the FSAR, which constitute a reduction in
commitments, are governed by 10 CFR 50.54(a). Therefore, sufficient
controls for these requirements exist and these changes do not
involve a significant increase in the probability or consequences of
an accident previously analyzed.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: This amendment application does not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed changes involve the relocation of
SRC requirements from the TS to the FSAR. Relocation of these
requirements does not affect plant equipment or the way the plant
operates. The reviews, audits, and record keeping will continue to
be performed in the identical manner as they are currently being
performed. Therefore, the proposed revisions cannot create a new or
different kind of accident.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: This amendment application does not involve a
significant reduction in a margin of safety. The requested Technical
Specification revisions relocate SRC review, audit and related
record keeping requirements from the TS to the FSAR. These
requirements are not being altered by this relocation. The reviews,
audits, and record keeping will continue to be performed in the
identical manner as they are currently being performed. Any changes
to these requirements which constitute a reduction in commitments
will be processed in accordance with 10 CFR 50.54(a). 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 request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
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 amendment request: July 10, 1998.
Description of amendment request: The proposed changes would
relocate portions of reactor coolant chemistry requirements from the
technical specifications (TSs) to licensee-controlled procedures.
Changes to the relocated requirements will then be controlled by the
provisions of 10 CFR 50.59.
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 TS amendment will not significantly increase the
probability or consequences of any previously evaluated accidents.
The proposed changes simplify the TS, meet regulatory
requirements for relocated TS, and implement the recommendations of
the Commission's Final Policy Statement on TS improvements. Future
changes to these requirements will be controlled by 10 CFR
[[Page 40561]]
50.59. The proposed changes are administrative in nature and do not
involve any modification to any plant equipment or affect plant
operation. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of any
previously evaluated accident.
2. The proposed TS amendment will not create the possibility of
a new or different kind of accident.
The proposed changes are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety related system performs
its function. Therefore, this proposed TS amendment will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed TS amendment will not involve a significant
reduction in a margin of safety.
The proposed changes are administrative in nature, will not
alter the basic regulatory requirements, and do not affect any
safety analyses. Therefore, no margin of safety is reduced as a
result of these changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 5, 1997, and supplemented April 21,
1998.
Description of amendment request:
Part 1--DG Online Testing:
The proposed amendment involves the testing of the standby diesel
generators (DGs) and revises the Watts Bar Unit 1 (WBN) Technical
Specifications (TSs) to allow additional testing of the DGs on-line
during MODES 1 and 2. The proposed changes affect Surveillance
Requirement (SR) 3.8.1.14. The testing performed for this surveillance
fulfills the requirements of Regulatory Guide 1.9, ``Selection, Design,
Qualification, and Testing of Emergency Diesel Generator Units Used as
Class 1E Onsite Electric Power Systems at Nuclear Power Plants.'' This
testing is performed once every 18 months to ensure that the DGs can
start and run continuously for an interval of not less than 24 hours.
Specifically, the proposed amendment revises SR 3.8.1.14 and its
associated Bases to delete the note which prohibits the performance of
the on-line 24 hour test during MODES 1 or 2.
Part 2--DG Battery Testing:
As currently written, the TSs permit testing of the DG batteries
and chargers only during MODES 5 and 6 when operability of all four DGs
is not required. The proposed amendment would revise the Watts Bar Unit
1 TSs to allow testing of the DG batteries and battery chargers during
MODES 1, 2, 3, and 4 as well. Implementation of these changes will
require entry into Action B.4 of TS 3.8.1 for the affected diesel.
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:
Part 1--DG Online Testing
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment to allow the 24-hour DG endurance run to
be conducted during any mode of operation does not significantly
increase the probability or consequences of an accident previously
evaluated in Chapter 15 of the FSAR [Final Safety Analysis Report]
since the capability to safely shutdown the plant following a LOOP
[loss of offsite power], LOCA [loss-of-coolant accident] or LOCA/
LOOP coincident with a single failure is maintained throughout the
surveillance test. The 24-hour endurance test does not disable any
of the automatic actuations and interlocks of the DG control
functions, nor prevent the satisfactory completion of the LOOP or
LOCA/LOOP loading sequence if a LOOP or LOCA signal is received at
any time during the test. Required Class-1E onsite power operability
during normal operation, shutdown cooling, loss of offsite power,
and accident conditions will be the same.
In addition, the performance of proposed Surveillance
Requirement 3.8.1.14 during MODES 1 or 2 will not significantly
increase the consequences of perturbations to any of the electrical
distribution systems that could result in a challenge to steady
state operation or to plant safety systems. Performance of proposed
Surveillance Requirement 3.8.1.14 during MODES 1 or 2, or failure of
the surveillance, will not cause, or result in, an anticipated
operational occurrence with attendant challenges to plant safety
systems that has not been previously analyzed for the existing
monthly surveillances.
Therefore, TVA concludes that the above change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The requested changes do not result in a new or different kind
of accident from that previously analyzed in WBN's Final Safety
Analysis Report. The changes propose to eliminate restrictions of
the plant operating modes in which standby DG system testing may be
performed but does not change the type of testing performed and are
not due to modification of the system design. NRC's assessment of
the testing of the DGs in the configuration proposed is documented
in Section 8.3.1.12 of Supplements 13 and 14 of the Safety
Evaluation Report and in letters dated June 20, 1991, and March 28,
1994.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
As previously stated, performance of proposed Surveillance
Requirement 3.8.1.14 during Modes 1 or 2 will not cause, or result
in, an anticipated operational occurrence with attendant challenges
to plant safety systems that has not been previously analyzed for
the existing monthly surveillances. Therefore, implementation of the
proposed amendment will not reduce the margin of safety for this
system.
Part 2--DG Battery Testing
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes to the TSs apply only to the DG battery
system and do not in any way affect the vital battery system or
safety system loads supplied by the vital battery system. The
changes do not result in a condition where the design or function of
the DGs or DG battery systems would be modified. The DG battery
subsystems supply only the control and field flashing power to
support a single DG and do not supply any other unrelated system
loads or functions. Therefore, manipulation of the DG battery system
is not a credible means of perturbing the vital power distribution
system and challenging safety systems. In addition, the
surveillances for the DG batteries are required to be performed only
once every 18 months.
A DG declared inoperable due to the testing must be returned to
operable status within 72 hours in accordance with Action B.4 of TS
3.8.1. To ensure this could be achieved, the results of previous
performances of the SRs were reviewed. From this review, it was
established that in accordance with LCO [limiting condition for
operation] 3.8.6, Table 3.8.6-1, Note c, the batteries can be
restored within 72 hours to a condition where the charging current
is less than 1 ampere. Achieving this charging current for the DG
batteries is acceptable for meeting specific gravity limits
following a battery recharge for a maximum of 31 days. In addition,
the DG sets are occasionally removed from the standby condition to
perform preventative and/or corrective maintenance. The intent is to
perform this
[[Page 40562]]
testing in conjunction with other required maintenance activities
such that adverse effects on diesel unavailability are minimized.
Compliance with the 10 CFR 50.65 Maintenance Rule program
requirements for diesel unavailability ensures that any diesel
inoperability incurred by this change is minimized.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The requested changes do not result in a new or different kind
of accident from that previously analyzed in WBN's Final Safety
Analysis Report. The changes propose to eliminate restrictions of
the plant operating modes in which DG battery system testing may be
performed but does not change the type of testing performed and are
not due to modification of the system design. The requested changes
will result in a DG being declared inoperable in accordance with
Action B.4 of TS 3.8.1 for the duration of the testing, but does not
impact the existing time limitations for the LCO. This change does
not alter system performance and does not introduce any new accident
initiators or scenarios.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.
The proposed amendment concerns only the conduct of testing but
does not in any way affect the performance parameters of the safety
system or in any way affect the ability of the system to perform its
safety function of providing control and field flashing power for
the DGs. Consequently, operation of the facility in accordance with
the requested changes would not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 26, 1998
Description of amendment request: The proposed amendment would
revise the Watts Bar Nuclear Plant (WBN) Technical Specifications (TS)
and associated Bases to delete the power range neutron flux high
negative rate reactor trip function based on the analysis provided in
Westinghouse Electric Corporation topical report WCAP-11394-P-A,
``Methodology for the Analysis of the Dropped Rod Event.''
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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The negative flux rate trip deletion does not increase the
probability or consequences of core damage accidents resulting from
dropped RCCA [rod cluster control assembly] events previously
analyzed. The safety functions of other safety related systems and
components, which are related to accident mitigation, have not been
altered. All other primary protection (reactor trip and ESF)
functions are not impacted by the elimination of the negative flux
rate trip function. The consequences of accidents previously
evaluated in the FSAR [final safety analysis report] are unaffected
by this proposed change because no change to any equipment response
or accident mitigation scenario has resulted. There are no
additional challenges to fission product barrier integrity. No new
radiological analyses are required. Therefore the proposed change
will have no effect on the probability or consequences of accidents
previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The negative flux rate trip deletion does not create the
possibility of a new or different kind of accident than any accident
already evaluated in the FSAR. No new accident scenarios, failure
mechanisms, or limiting single failures are introduced as a result
of this proposed change. The proposed modification does not
challenge the performance or integrity of any safety-related
systems.
It has been demonstrated that the function of the negative flux
rate trip can be eliminated by the approved methodology described in
WCAP 11394-P-A. A Watts bar specific analysis has confirmed that for
the dropped RCCA and dropped RCCA bank event, no direct reactor trip
or automatic power reduction is required to meet the DNB [departure
from nucleate boiling] licensing basis for this Condition II event.
The negative flux rate trip function is not credited as a backup for
any other Chapter 15 event. Thus, this change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The margin of safety associated with the acceptance criteria for
any postulated WBN accident is unchanged. It has been demonstrated
that the function of the negative flux rate trip can be eliminated
by the approved methodology described in WCAP 11394-P-A. Watts Bar
specific analysis has confirmed that the dropped RCCA and dropped
RCCA bank acceptance criteria (DNB) continue to be met. Conformance
to the regulatory criteria for plant operation with the negative
flux rate trip deletion is demonstrated, and regulatory limits (DNB)
are not exceeded. The modification will have no effect on the
availability, operability, or performance of the safety-related
systems and components. Therefore, the proposed license amendment
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power
Plant, Unit 1, Lake County, Ohio
Date of amendment request: June 30, 1998.
Description of amendment request: The proposed license amendment
revises Technical Specification 3.1.7, ``Standby Liquid Control
System.'' The purpose of the proposed change is to increase the boron
concentration in the Standby Liquid Control System for the Perry
Nuclear Power Plant Cycle 8 fuel design, and to provide margin for
future cycles as required.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change will not significantly increase the
probability or consequences of an accident previously evaluated. The
change will only vary the ratio of borax to boric acid that resides
within the Standby Liquid
[[Page 40563]]
Control System (SLCS) as the neutron absorber.
Changing the definition of the solution from a mixture of Sodium
Pentaborate having a molar ratio of 0.200, to a mixture of borax and
boric acid having a nominal molar ratio of 0.229, does not degrade
the stability of the solution, change the mixing accuracy
requirements, or reduce the temperature margins that might add to
the risk of solution crystallization. For each cycle, the reload
safety analysis confirms that the SLCS boron concentration will
satisfy the Technical Specification requirements for the Perry
Nuclear Power Plant (PNPP).
The 5 deg.F margin of safety for solution solubility will
continue to be maintained and supported by the Containment Building
ambient temperatures and additionally supplemented by auto initiated
heating on the SLCS tank and piping. The chosen borax and boric acid
molar ratio will continue to maintain a limiting chemical addition
mass, to the plus 5 deg.F solubility limit, greater than or equal to
the current 0.200 mixture. Any inaccuracies associated with tank
temperature, tank volume, chemical analysis, and initial and
subsequent chemical additions to the tank will also remain the same.
The primary reactivity control system for postulated accident
conditions is the control rod system. The SLCS is a redundant
reactivity control system to the control rod system and is used in
special plant capability demonstration events cited in Appendix A of
the Updated Safety Analysis Report (USAR), Chapter 15, which are
extremely low probability non-design basis postulated incidents.
There are no postulated accidents evaluated in USAR Chapter 15 that
take credit for two or more reactivity control systems preventing or
mitigating each accident. There is no increase to the radiological
consequences of postulated incidents with the proposed change.
With the implementation of this proposed change, the SLCS will
continue to operate and perform to all of its current requirements
for providing shutdown margin under operating and ATWS conditions
per 10 CFR 50, Appendix A, General Design Criterion (GDC) 26 and 10
CFR 50.62. The proposed change will not alter the operation of any
plant equipment assumed to function in response to an analyzed event
or otherwise increase its failure probability. 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.
The proposed change does not impact the operation of the SLCS or
the function of any of the active components. No new system
interactions are created by this change and any parameters or
conditions that could contribute to the initiation of accidents
different than those already evaluated in the USAR are not impacted.
The change will only vary the ratio of borax to boric acid that
resides within the SLCS as the neutron absorber. The proposed values
for solution molar ratio and boron concentration ensure that
solution temperature margins are maintained greater than or equal to
the current required margin to prevent solution crystallization. As
a result, no new failure modes are being introduced.
Changing the definition of the solution from Sodium Pentaborate
having a nominal molar ratio of 0.200, to a mixture of borax and
boric acid having a nominal molar ratio of 0.229 does not degrade
the stability of the solution, change the mixing accuracy
requirements, or reduce the temperature margins that might add to
the risk of solution crystallization.
Sufficient margin will be maintained to allow for expected
deviations in the molar ratio and boron weight as the result of
variations in product composition, test measurement inaccuracies,
and for chemical addition inaccuracies. The boron concentration
required within the SLC system to meet the required shutdown margin,
will continue to be determined for each fuel cycle as part of the
reload safety analysis per Technical Specifications. The borax and
boric acid concentration will remain controlled via the Technical
Specification Surveillance Requirements and the associated
administrative procedures, USAR text, and existing licensing
commitments.
The SLC system will meet its design basis requirements for the
weight of boron injected and for maintaining the required
temperature margin for system operation. As the result of the
proposed change to increase the minimum boron concentration, a new
minimum required SLC pump flow rate was determined for compliance
with the NRC ATWS Rule 10 CFR 50.62.
The proposed change meets current regulations, maintains the
fundamental safety principles of plant design, and the associated
margins of safety. With the implementation of the proposed change,
the SLCS will continue to operate and perform to all of its current
requirements for providing shutdown margin under operating and ATWS
conditions per 10 CFR 50, Appendix A, GDC 26 and 10 CFR 50.62. As a
result, no new failure modes are being introduced. There are no
changes in the methods governing normal plant operations, nor are
the methods used to respond to plant transients altered. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin to safety.
The required margin of safety for the SLCS solution ensures an
adequate margin of solubility such that no precipitation will occur
in the SLC storage tank. The current margin is provided by
maintaining a minimum solution temperature that is no less than the
saturation temperature corresponding to the concentration of the
solution in the storage tank plus 5 deg.F.
This 5 deg.F provides the adequate margin for inaccuracies
associated with tank temperature, tank volume, chemical analysis,
and initial and subsequent chemical additions to the tank. The
proposed change does not impact the inaccuracies associated with
tank temperature, tank volume, chemical analysis, and initial and
subsequent chemical additions to the tank. The new analytical design
values for the molar ratio and boron concentration will continue to
maintain the solution temperature margins in excess of the current
minimum specified to prevent solution crystallization.
Ambient temperatures within the building that houses the SLC
storage tank, the Containment Building, will maintain the solution
temperature. Additionally, the solution temperature is maintained by
the presence of auto initiated tank heaters and pipe heat tracing.
The 5 deg.F margin will be maintained with the new SLCS mole ratio
and higher boron concentration with the existing instrument
setpoints and administrative controls.
The proposed change maintains the same reactor shutdown margin
for the next fuel cycle and does not reduce the margin of safety for
any system parameter as defined in the bases for the Technical
Specifications. The proposed change will not physically alter the
SLCS's physical configuration or components or introduce new system
interactions that could produce any parameters or conditions that
could contribute to a reduction of safety for any other system or
scenario. The change will only vary the ratio of borax to boric acid
that resides within the SLCS as the neutron absorber.
Therefore, with the implementation of this proposed change, the
SLCS will continue to operate and perform to all of its current
requirements for providing shutdown margin under operating and ATWS
conditions per 10 CFR 50, Appendix A, GDC 26 and 10 CFR 50.62 and
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ronald R. Bellamy (Acting).
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: June 30, 1998.
Description of amendment request: The licensee proposes to delete
the calibration requirements for emergency core cooling actuation
instrumentation--core spray (CS) subsystem and low pressure coolant
injection (LPCI) system auxiliary power
[[Page 40564]]
monitor since the relays operate from a switched input and functional
testing is sufficient to demonstrate the relay pickup/dropout
capability.
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 change does not involve a change to the plant
design or operation. The Auxiliary Power Monitor logic relays
installed are tested to fully demonstrate operability without
performance of a calibration on the pickup voltage value. The design
intent of the relays is to start LPCI and CS pumps as soon as
possible without causing loss of the normal or emergency power
supplies and within the time frames specified in the LOCA analysis
of record. The proposed change does not affect any of the parameters
or conditions that contribute to initiation of any accidents
previously evaluated. Thus, the proposed change cannot increase the
probability of an accident previously evaluated.
The proposed change does not involve a change in the operation
of the relays controlling [Residual Heal Removal] RHR and CS Pump
start with normal power available nor the initial RHR pump start on
a LOCA with normal power not available or the time delay start of
the remaining RHR or CS pumps with normal power not available.
Failure of the relays to pickup would still result in the start
sequence for normal power not available. The logic for both start
sequences is verified independent of an instrument calibration and
is consistent with the LOCA analysis and the EDG load analysis,
therefore, the proposed change does not significantly increase the
consequences of any accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated:
This proposed change will not involve any physical changes to
plant systems, structures or components (SSC), or the manner in
which these SSCs are operated or maintained. The calibration
requirement has previously been considered to be met by performance
of the Simulated Automatic Actuation Test. Deletion of the
calibration requirement will not affect the RHR or CS Pumps starting
on a LOCA signal, with or without an [Loss of Normal Power] LNP. The
operability of the Auxiliary Power Monitor relays will still be
tested under the Functional test and Trip System Logic and Simulated
Automatic Actuation tests at the frequencies specified. Therefore,
this change will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed amendment will not involve a significant
reduction in the margin of safety:
This proposed change to delete the calibration requirement for
the CS and LPCI Auxiliary Power Monitor functions will not change
operation of the RHR or CS Pump start sequences on a LOCA signal,
with or without normal power available. The instantaneous logic
sequence relays and time delay relays will function to initiate RHR
and CS Pump start as designed. RHR and CS Pump start times will
remain within the LOCA Safety Evaluation of record. Operability of
the relays and associated circuitry are still demonstrated by the
Functional test and associated Trip System Logic and Simulated
Automatic Actuation tests. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Project Director: Cecil O. Thomas.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Carolina Power & Light Company, et al., Docket No. 50-261, H.B.Robinson
Steam Electric Plant, Unit 2, Darlington County, South Carolina
Date of amendment request: June 26, 1998.
Brief Description of amendment: The proposed amendment would revise
Technical Specification (TS) 3.7.8, ``Ultimate Heat Sink (UHS),'' to
permit an 8-hour delay in UHS temperature restoration period prior to
entering the plant shutdown required actions. Also, for the duration of
the restoration, service water system (SWS) temperature will be
monitored hourly, and should the temperature exceed 99 degrees F, the
plant will enter TS 3.7.8 required action A.1, and be in MODE 3 within
6 hours.
Date of publication of individual notice in the Federal Register:
July 8, 1998 (63 FR 36967).
Expiration date of individual notice: July 22, 1998, for comments;
August 7, 1998, for hearings.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
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: March 3, 1998, as supplemented by
letters dated April 24 and May 7, 1998.
Description of amendment request: The proposed amendments would
revise Figure 5.1-1 of the Technical Specifications (TS) to show the
new location of the meteorological tower. The meteorological tower will
be relocated to a new location to facilitate use of the current
location as a construction site. The proposed TS change does not change
the related TS Section 5.1.1.
Date of publication of individual notice in Federal Register: June
29, 1998 (63 FR 35293).
Expiration date of individual notice: July 29, 1998.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant
[[Page 40565]]
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.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: January 28, 1998 (NRC-98-0002).
Brief description of amendment: The amendment revises technical
specification surveillance requirements 4.8.2.1.a.2, 4.8.2.1.b, and
4.8.2.1.c.4 to accommodate new limits associated with the design of the
replacement Division II 130/260-volt dc battery.
Date of issuance: July 9, 1998.
Effective date: July 9, 1998, with full implementation prior to
restart from the sixth refueling outage.
Amendment No.: 121.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9597).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: December 10, 1997 (NRC-97-0105),
as supplemented January 28 and April 9, 1998.
Brief description of amendment: The amendment revises Technical
Specification (TS) 2.2.1, ``Reactor Protection System Instrumentation
Setpoints,'' TS 3.3.1, ``Reactor Protection System Instrumentation,''
TS 3.3.6, ``Control Rod Block Instrumentation,'' TS 3.4.1.1,
``Recirculation Loops,'' and the associated Bases to accommodate an
upgrade of the power range neutron monitoring system. The amendment
also revises the first page of Table 3.3.6-2 to correct a typographical
error in the title.
NRC has also granted the request of Detroit Edison Company to
withdraw a portion of its December 10, 1997, application. The proposed
change would have revised TS Surveillance Requirement 4.3.1.3 and its
associated Bases to indicate response time testing is performed only on
applicable channels. However, following discussions with the NRC staff,
the licensee withdrew the proposed change in a letter dated April 9,
1998 (NRC-98-0037). For further details with respect to this action,
see the application for amendment dated December 10, 1997, as
supplemented above, and the licensee's letter dated April 9, 1998,
which withdrew this portion of the application for license amendment,
and the staff's Safety Evaluation enclosed with the amendment. The
above documents are available for public inspection at the Commission's
Public Document Room, the Gelman Building, 2120 L Street, NW.,
Washington, DC, and at the local public document room listed below.
Date of issuance: July 13, 1998.
Effective date: July 13, 1998, with full implementation prior to
restart from the sixth refueling outage. Implementation of this
amendment shall include preparation of Design Calculation DC-5721,
Volume I, and performance of a human factors review for the
installation of the plant modification as described in the licensee's
application dated December 10, 1997, as supplemented January 28 and
April 9, 1998, and as evaluated in the staff's safety evaluation
attached to this amendment.
Amendment No.: 122.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 14, 1998 (63 FR
2279). The January 28 and April 9, 1998, letters provided clarifying
information and updated TS pages that were within the scope of the
original Federal Register notice and did not change the staff's initial
proposed no significant hazards considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 13, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: April 20, 1998.
Brief description of amendments: The amendments revise Tables 3.3-3
and 4.3-2 of the Technical Specifications of each unit, correcting the
operation mode applicability of the control room area ventilation
actuation logic and relays from ``All'' to ``1, 2, 3, 4.''
Date of issuance: July 9, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 167--Unit 1; 159--Unit 2.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 20, 1998 (63 FR
27761).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas
Date of amendment request: October 2, 1996.
Brief description of amendments: The amendments revise the ANO-1&2
TSs by relocating selected TS requirements related to instrumentation
from the TS to the Updated Final Safety Analysis Report. The NRC
provided guidance to all holders of operating licenses or construction
permits for nuclear power reactors on the proposed TS changes in
Generic Letter 95-10, ``Relocation of
[[Page 40566]]
Selected Technical Specifications Requirements Related to
Instrumentation,'' dated December 15, 1995.
Date of issuance: July 13, 1998.
Effective date: July 13, 1998, to be implemented within 30 days.
Amendment Nos.: 192 and 191.
Facility Operating License Nos. DPR-51 and NPF-6: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 15, 1997 (62 FR
2188).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 13, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 27, 1997, as supplemented by
letters dated April 3, July 21, October 23, November 13, and December
12, 1997, January 21, January 29, March 23, May 1, May 19, and May 21,
May 28, and June 12, 1998.
Brief description of amendment: The amendment changes Appendix A
Technical Specification by increasing the Spent Fuel Pool storage
capacity from 1088 to 2398 fuel assemblies and by increasing the
maximum fuel enrichment from 4.9 w/o (weight percent) to 5.0 w/o U-235.
Date of issuance: July 10, 1998.
Effective date: July 10, 1998.
Amendment No.: 144.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 2, 1997 (62 FR
63732).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 10, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: March 12, 1998.
Brief description of amendments: The amendments revised Turkey
Point Units 3 and 4 Facility Operating Licenses and Technical
Specifications to remove certain license conditions and oudated
references, and to incorporate an organizational change.
Date of issuance: July 9, 1998.
Effective date: July 9, 1998.
Amendment Nos.: 198 and 192.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised Turkey Point Units 3 and 4 Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: April 8, 1998 (67 FR
17225).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Niagara Mohawk Power Corporation, Docket Nos. 50-220 and 50-410, Nine
Mile Point Nuclear Station Unit Nos. 1 and 2, Oswego County, New York
Date of applications for amendments: May 15, 1998 (two letters, one
for each unit).
Brief description of amendment: The amendments change
administrative sections of the Technical Specifications to reflect a
restructuring of licensee's Nuclear Division upper management
organization.
Date of issuance: July 7, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 162 and 83.
Facility Operating License Nos. DPR-63 and NPF-69: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: June 2, 1998 (63 FR
30026).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 7, 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.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Stawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Project Director: S. Singh Bajwa.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: September 28, 1995, and April
23, 1998.
Brief description of amendments: The amendments revise Technical
Specification 3/4.8.1.2, ``Electrical Power Sources--Shutdown,'' by
adding a note to surveillance requirement 4.8.1.2 that identifies those
surveillances which are required to be performed during Modes 5 and 6
(cold shutdown and refueling, respectively).
Date of issuance: July 14, 1998.
Effective date: July 14, 1998.
Amendment Nos.: 212 and 192.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56369).
The April 23, 1998, letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 14,1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: September 29, 1998, as
supplemented February 6, 1998, April 17, 1998, and June 4, 1998.
Brief description of amendment: This amendment revises the
allowable value and trip setpoint for the main steam isolation high
steam flow input into limiting condition for operation.
Table 3.3.2-1, function 4.d.
Date of issuance: July 14, 1998.
Effective date: July 14, 1998.
Amendment No.: 71.
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54876).
The February 6, 1998, April 17, 1998, and June 4, 1998, letters
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 July 14, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
[[Page 40567]]
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 6, 1996, as supplemented
September 26, 1997, January 23, 1998, and May 19, 1998 (TS-372).
Brief description of amendments: Changes to the technical
specifications administrative controls related to quality assurance,
and other administrative and editorial changes.
Date of issuance: July 9, 1998.
Effective date: July 9, 1998.
Amendment Nos.: 233, 252, and 211.
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 25, 1996 (61
FR 50346).
The supplemental letters dated September 26, 1997, January 23, and
May 19, 1998 did not change the original no significant hazards
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 9, 1998.
No significant hazards consideration comments received: None.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: April 24, 1998.
Brief description of amendment: This amendment changed Technical
Specification (TS) Section 3/4.3.1.1, ``Reactor Protection System
Instrumentation,'' TS Section 3/4.3.2.1, ``Safety Features Actuation
System Instrumentation,'' TS Section 3/4.3.2.2, ``Steam and Feedwater
Rupture Control System Instrumentation,'' and the associated TS bases.
The TS tables of response time limits were relocated to the Davis-Besse
Technical Requirements Manual. Other changes in these TS sections were
also made consistent with the relocation.
Date of issuance: July 7, 1998.
Effective date: July 7, 1998.
Amendment No.: 225.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated July 7, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: April 24, 1996, as supplemented
December 15, 1997, and June 22, 1998.
Brief description of amendments: These amendments revise Technical
Specifications (TS) Section 15.7, ``Radiological Effluent Technical
Specifications (RETS).'' Portions of the RETS are moved to licensee-
controlled documents consistent with Nuclear Regulatory Commission
guidance on TS improvements. Other sections of the TSs have also been
revised consistent with the removal of portions of the RETS.
Date of issuance: July 13, 1998.
Effective date: July 13, 1998, with full implementation within 45
days. Implementation shall include relocation of certain Technical
Specification requirements to licensee-controlled documents, as
described in the licensee's application dated April 24, 1996, as
supplemented by letter dated December 15, 1997, and June 22, 1998, and
evaluated in the staff's safety evaluation attached to the amendments.
Amendment Nos.: 184 and 188.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28620) The December 15, 1997, and June 22, 1998, submittals provided
additional clarifying information and updated TS pages. This
information was within the scope of the original Federal Register
notice and did not change the staff's initial no significant hazards
considerations determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 13, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Dated at Rockville, Maryland, this 22nd day of July 1998.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-20111 Filed 7-28-98; 8:45 am]
BILLING CODE 7590-01-P