[Federal Register Volume 62, Number 242 (Wednesday, December 17, 1997)]
[Notices]
[Pages 66133-66151]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-32763]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 21, 1997, through December 5, 1997.
The last biweekly notice was published on December 3, 1997 (62 FR
63970).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By January 16, 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
[[Page 66134]]
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
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.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: February 28, 1997.
Description of amendment request: The proposed amendments would
revise Byron and Braidwood Technical Specifications (TS) Sections 3/
4.4.5, ``Steam Generators,'' and 3/4.4.8, ``Reactor Coolant System
Specific Activity,'' for both the Byron Station, Units 1 and 2, and the
Braidwood Station, Units 1 and 2. The intent of these proposed
revisions is to restore for both Byron, Unit 1, and Braidwood, Unit 1,
the original TS related to steam generator (SG) inspections and the
primary coolant dose equivalent iodine-131 (DEI) concentrations. These
amendments will become effective when the original steam generators
(OSG) which are Westinghouse Model D4 SGs, are removed and the
replacement steam generators (RSG) made by Babcock and Wilcox,
International (BWI), are installed. The RSGs are presently being
installed at Byron, Unit 1, while the RSGs will be installed at
Braidwood, Unit 1, in fall 1998.
The SG inspection methodology, inspection frequency, reporting
requirements and acceptance criteria for the RSGs in both Byron, Unit
1, and Braidwood, Unit 1, will revert to the TSs for the OSGs before
several prior license amendments incorporated into the TSs: (1) The
interim plugging criteria (IPC) consistent with Generic Letter (GL) 95-
05; (2) the F* criteria for the SG tube expansions into the tubesheet;
and (3) the criteria for repairing SG tubes using either Westinghouse
laser welded sleeves or Combustion Engineering tungsten inert gas (TIG)
welded sleeves. The TSs applicable to Byron, Unit 2, and Braidwood,
Unit 2, both of which have Westinghouse Model D5 SGs, remain unchanged
except for designating them in the TSs as model D5 SGs.
With respect to the limiting value of the DEI primary coolant
concentration, both the Byron, Unit 1, TSs and the Braidwood, Unit 1,
TSs will revert from their present TS limit of 0.35 to 1.0 microcuries
per gram. A license amendment request to lower the Byron, Unit 1, TS
DEI limit from 0.35 to 0.20 microcuries per gram was submitted on
January 31, 1997, but this request was
[[Page 66135]]
subsequently withdrawn on November 11, 1997, because the RSGs were
being installed in the Byron, Unit 1, refueling outage which started in
early November 1997. A license amendment request to lower the
Braidwood, Unit 1, TS DEI limit from 0.35 to 0.10 microcuries per gram
was submitted on September 2, 1997. Action on this request is still
pending but in any case, will not affect the subject license amendment
request for Braidwood, Unit 1, because the September 2, 1997, request
is only applicable to the OSGs which are presently using the IPC that
were originally incorporated into the TSs on November 9, 1995. The
applicable bases sections of the Byron, Unit 1, TSs and Braidwood, Unit
1, TS will also be revised to reflect the TS changes discussed above.
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.
Due to design differences between the replacement Steam
Generators (RSGs) and OSGs, the analyses supporting the application
of the F* and voltage-based repair criteria do not apply to the
RSGs. Also, the analyses supporting sleeving repair by the
Westinghouse laser welded or Combustion Engineering Tungsten Inert
Gas (TIG) welded sleeving methodologies do not apply to the RSGs due
to the design differences. The RSG and OSG tube bundle
configurations are similar, however, the RSG tubes are smaller in
diameter, constructed of Inconel Alloy 690 instead of Alloy 600, and
supported by stainless steel lattice grids instead of the drilled
carbon steel plates used in the OSGs. The RSG tubes are
hydraulically expanded into the tube sheet during initial assembly.
The RSG upper tube bundle shape consists of tubes with continuous,
smooth, long radius bends.
The structural analysis demonstrates that the tube integrity is
maintained for a Main Steamline Break (MSLB) occurring during normal
full power operation. The structural evaluation of the tubing for
faulted conditions was performed in accordance with the ASME Boiler
and Pressure Vessel Code Section III requirements. The tube material
selection and size exceed the strength requirements of the existing
steam generators. Comparison of the Alloy 690 tube material used in
the RSGs with the Alloy 600 tube material in the OSGs show that the
RSG material strength characteristics are as good as or better than
those of the existing design. A comparison of the stress margins of
the RSG and OSG show that the stress margin in the RSG tubes exceed
the stress margin in the OSG tubes.
RSG portions of the reactor coolant pressure boundary are
designed to permit periodic inspection and testing of important
areas and features to assess structural and leak-tight integrity.
ASME Section XI, provides the depth of an allowable outside diameter
(O.D.) flaw for tubes in service. The RSG has tubing fabricated from
SB-163 material (Inconel Alloy 690) which is examined by eddy
current methods to the requirements of ASME Section III, NB-2550.
The tubing has a radius to thickness (r/t) ratio less than 8.70. In
accordance with ASME Section XI, for tubing having an r/t ratio of
less than 8.70, the depth of an allowable O.D. flaw shall not exceed
40% of the nominal tube wall thickness.
The potential for tube rupture is not increased from the OSGs as
demonstrated in the qualification analysis and testing for the RSGs.
The program for periodic inservice inspection of the steam
generators monitors the integrity of the SG tubing to ensure that
there is sufficient time to take proper and timely corrective action
if any tube degradation is detected. Therefore, installation of the
RSGs will not increase the probability of the occurrence of primary-
to-secondary leakage or a steam generator tube rupture (SGTR) during
normal or accident conditions.
The design basis doses calculated for postulated accidents
involving degradation of SG tubes, such as SGTR and MSLB accidents,
as presented in UFSAR [Updated Final Safety Analysis Report] Chapter
15 accident analysis have been evaluated and are decreased by
installation of the RSGs and restoration of the RCS activity limit
to 1.0 microcuries/gm. The decrease in offsite dose is primarily due
to the smaller RSG tube diameter and less primary-to-secondary
transfer during the event. The dose calculations are performed
consistent with NUREG-0800, ``Standard Review Plan'' and ensure site
boundary doses are within a small fraction of the Title 10 Code of
Federal Regulations Part 100 (10 CFR 100) requirements. Therefore,
the change does not involve a significant increase in the
consequences of an accident previously evaluated.
Limiting the applicability of TS provisions to a specific cycle
or SG type are administrative changes in that they provide
clarification consistent with current analyses and do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
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.
Restricting application of IPC, F* and sleeving methodologies to
the OSGs and reinstating an RCS activity limit of 1.0 microcuries/gm
upon installation of the RSGs will not introduce significant or
adverse changes to the plant design basis that could lead to a new
or different kind of accident being created. The RSG tubing meets
the requirements of General Design Criteria (GDC) 14, 15, 30, 31,
and 32 of 10 CFR 50, Appendix A. The RSG tubing has been designed
and evaluated consistent with ASME Code Section III criteria and the
inspection criteria for the RSGs is consistent with ASME Code
Section XI criteria. The RSGs have thermally treated Inconel Alloy
690 tubes which are hydraulically expanded into the tube sheet
during initial assembly. Alloy 690 is more resistant to stress
corrosion cracking (SCC) than Alloy 600 which is used in the OSG
tubing. Overall tube bundle structural and leakage integrity is
maintained at a level consistent with or better than the originally
supplied tubing during all plant conditions.
ComEd will continue to apply the TS maximum primary-to-secondary
leakage limit of 150 gpd (0.1 gpm) through any one SG at Byron and
Braidwood to help preclude the potential for excessive leakage
during all plant conditions. The EPRI recommended 150 gpd limit
provides for leakage detection and plant shutdown in the event of an
unexpected tube leak and precludes the potential for excessive
leakage or tube burst in the event of a Main Steam Line Break or
under Loss of Coolant Accident conditions.
Limiting the applicability of TS provisions to a specific cycle
or SG type are administrative changes in that they provide
clarification consistent with current analyses.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Restricting application of IPC, F*, and sleeving methodologies
to the OSGs for which the supporting analyses apply, does not
involve a reduction in a margin of safety. The RSG tubing has been
shown to retain adequate structural and leakage integrity during
normal, transient, and postulated accident conditions consistent
with GDC 14, 15, 30, 31, and 32 of 10 CFR 50 Appendix A. The RSG
tubing has been designed and evaluated consistent with the margins
of safety specified in ASME Code Section III. The proposed program
for periodic inservice inspection of the replacement steam
generators monitors the integrity of the SG tubing to ensure that
there is sufficient time to take proper and timely corrective action
if any tube degradation is present. The proposed program is
consistent with the Standard Technical Specifications.
The Unit 1 RCS dose equivalent I-131 limit is being raised upon
installation of the RSGs to eliminate the compensatory lower limit
that was adopted in conjunction with IPC for the existing
Westinghouse D4 SGs. With the RCS activity limit returned to the
Standard Technical Specification value of 1.0 [mu]Ci/gm, the
assessment of postulated UFSAR Chapter 15 accidents (including SGTR
and MSLB) has concluded that the calculated design basis doses
presented in Chapter 15 are not adversely impacted by the RSGs. This
ensures that the resulting 2-hour dose rates at the Byron and
Braidwood site boundaries will not exceed an appropriately small
fraction of 10 CFR 100 dose guideline values.
Limiting the applicability of TS provisions to a specific cycle
or SG type are
[[Page 66136]]
administrative changes in that they provide clarification consistent
with current analyses.
Therefore, it is concluded that this change does not involve a
significant reduction in a margin of safety with respect to plant
safety as defined in the UFSAR or the Technical Specification.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: March 26, 1997.
Description of amendment request: The proposed amendment would
revise the containment system technical specifications (TS) contained
in TS Sections 3.6 and 4.5. The licensee has classified the changes as
``More Restrictive,'' ``Less Restrictive,'' and ``Administrative.''
``More Restrictive'' changes include reduction of the allowable
containment pressure, addition of an action statement defining action
to be taken when the containment pressure limit is exceeded, addition
of a restriction on containment temperature, and revision of the
applicable conditions for the containment purge valves to require that
the valves be operable above 210 degrees F versus the current
requirement that they be operable above 525 degrees F. ``Less
Restrictive'' changes include addition of an allowance to enter an air
lock through a locked door to perform maintenance, addition of an
allowance to open containment isolation valves under administrative
control, revision of the applicable conditions for containment pressure
to exclude the cold shutdown operating condition, and addition of an
exception to the surveillance requirement requiring verification of the
status of ``locked-closed'' manual isolation valves after a refueling
outage to exclude requiring such verification for valves opened under
administrative control. ``Administrative'' changes include the deletion
of containment isolation valve tables and component identifiers from
the TS in accordance with Generic Letter 91-08 (``Removal of Component
Lists from Technical Specifications'') and editorial restructuring of
the affected TS sections to clarify the remaining requirements.
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:
Each proposed change has been classified as ``Administrative,''
``More Restrictive,'' or ``Less Restrictive.'' ``Administrative''
and ``More Restrictive'' changes are discussed generically; ``Less
Restrictive'' changes are discussed individually.
Five of the proposed changes are classified as being ``Less
Restrictive'':
(G.1) Allowance in LCO [Limiting Condition for Operation] 3.6.1
to enter an air lock to perform maintenance.
(G.2) Allowance in LCO 3.6.1 to open containment isolation
valves under administrative control.
(I.2) Revising the applicable conditions of LCO 3.6.2,
Containment Pressure to exclude Cold Shutdown.
(J.2) Exception in SR [Surveillance Requirement] 4.5.3d for
valves opened under administrative control as allowed by LCO 3.6.1.
(P) Allowance in SR 4.5.2 to enter an air lock to perform
maintenance.
Four of the proposed changes are classified as being ``More
Restrictive'':
(I.1) Revising LCO 3.6.2 to reduce the allowable containment
pressure.
(I.3) Addition of an action statement to LCO 3.6.2, Containment
Pressure.
(K) Addition of a new LCO which restricts Containment
Temperature.
(M.2) Revising the applicable conditions for LCO 3.6.5, Purge
Valves.
The remaining changes are all classified as being
``Administrative''.
Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P
allow limited access through the operable door of an air lock when
the other door is inoperable; current Technical Specifications [TS]
do not. Proposed changes G.2 and J.2 allow unisolating containment
penetration flow paths intermittently under administrative control;
current TS do provide a similar allowance, but only for one specific
penetration. These changes cannot significantly increase the
probability of an accident because opening an air lock door or a
containment penetration is not, itself, an initiator and does not
affect the items which are initiators of any analyzed accident.
The ability to open the operable door or to open a containment
penetration, even if it means the containment boundary is
temporarily not intact, does not significantly increase the
consequences of an accident previously evaluated because of the low
probability of an event that could pressurize the containment
occurring during the short time the operable door or containment
penetration is expected to be open. In a case where containment
integrity (or containment operability) is lost due to excessive
leakage, both the Palisades Technical Specifications and the
Standard Technical Specifications [STS] allow one hour of continued
operation for its restoration. That time period is allowed without
regard to the magnitude of the potential leakage, and would be
allowed even if both personnel air lock doors [were] leaking
excessively. The additional allowance of permitting the operable
door to be opened momentarily for entry or egress when the other
door is inoperable due to excessive leakage would not significantly
add to the probability of containment leakage and the resultant
consequences of an accident. Similarly, the allowance to open any
containment penetration intermittently under administrative control,
which currently is allowed for one penetration, would not
significantly add to the probability of containment leakage and the
resultant consequences of an accident.
Therefore, operation of the Facility in accordance with proposed
changes G.1, G.2, J.2, and P would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment
Pressure so that it no longer applies during Cold Shutdown. LCO
3.6.2 is intended to limit containment pressure to that value used
as an initial condition in the safety analysis. Containment pressure
is an initial condition in analyses which assure that containment
internal pressure will not exceed the containment design values
during a LOCA or MSLB. Containment pressure is not an initiator of
any accident previously evaluated. Neither a LOCA [loss-of-coolant
accident] nor a MSLB [main steam line break] occurring during Cold
Shutdown would pressurize the containment. Therefore, a containment
pressure LCO is not necessary, during Cold Shutdown, to assure that
containment design pressure and temperature is not exceeded. The STS
Containment pressure LCO is not applicable in Cold Shutdown.
Therefore, operation of the Facility in accordance with proposed
change I.2 would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
3. More Restrictive Changes: ``More Restrictive'' changes only
add new requirements, or revise existing requirements to result in
additional operational restrictions. The TS, with all ``More
Restrictive'' changes incorporated, will still contain all of the
requirements which existed prior to the changes. Therefore, ``More
Restrictive'' changes cannot involve a significant increase in the
probability or consequences of an accident previously evaluated.
4. ``Administrative'' changes make wording changes which clarify
existing TS requirements, without affecting their
[[Page 66137]]
technical content. Since ``Administrative'' changes do not alter the
technical content of any requirements, they cannot involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P
allow limited access through the operable door of an air lock when
the other door is inoperable; current Technical Specifications do
not. Proposed changes G.2 and J.2 allow unisolating containment
penetration flow paths intermittently under administrative control;
current TS do provide a similar allowance, but only for one specific
penetration. Opening an air lock door or a containment penetration
does not affect the operating conditions or operation of any plant
systems (other than the containment); it does not create a threat to
the integrity of any operating system or alter any system operating
practice or settings.
Since the opening of an air lock door or a containment
penetration only affects the potential leakage from the containment,
and does not affect any of the operating plant systems, operation of
the Facility in accordance with the proposed Technical
Specifications change would not create the possibility of a new or
different kind of accident from any previously evaluated.
2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment
Pressure so that it no longer applies during Cold Shutdown. LCO
3.6.2 is intended to limit containment pressure to that value used
as an initial condition in the safety analysis. Containment pressure
is an initial condition in analyses which assure that containment
internal pressure will not exceed the containment design values
during a LOCA or MSLB. Neither a LOCA nor a MSLB occurring during
Cold Shutdown would pressurize the containment. Therefore, a
containment pressure LCO is not necessary, during Cold Shutdown, to
avoid creation of a new or different kind of accident. The STS
Containment pressure LCO is not applicable in Cold Shutdown.
Therefore, operation of the Facility in accordance with proposed
change I.2 would not create the possibility of a new or different
kind of accident from any previously evaluated.
3. More Restrictive Changes: ``More Restrictive'' changes only
add new requirements, or revise existing requirements to result in
additional operational restrictions. The TS, with all ``More
Restrictive'' changes incorporated, will still contain all of the
requirements which existed prior to the changes. Therefore, ``More
Restrictive'' changes cannot create the possibility of a new or
different kind of accident from any previously evaluated.
4. ``Administrative'' changes make wording changes which clarify
existing TS requirements, without affecting their technical content.
Since ``Administrative'' changes do not alter the technical content
of any requirements, they cannot create the possibility of a new or
different kind of accident from any previously evaluated.
Do the proposed changes involve a significant reduction in a
margin of safety?
1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P
allow limited access through the operable door of an air lock when
the other door is inoperable; current Technical Specifications do
not. Proposed changes G.2 and J.2 allow unisolating containment
penetration flow paths intermittently under administrative control;
current TS do provide a similar allowance, but only for one specific
penetration. The ability to open the operable door or a containment
penetration, even if it means the containment boundary is
temporarily not intact, does not involve a significant reduction in
a margin of safety because of the low probability of an event that
could pressurize the containment occurring during the short time the
operable door or penetration is expected to be open.
Therefore, operation of the Facility in accordance with the
proposed Technical Specifications change would not involve a
significant reduction in a margin of safety.
2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment
Pressure so that it no longer applies during Cold Shutdown. LCO
3.6.2 is intended to limit containment pressure to that value used
as an initial condition in the safety analysis. Containment pressure
is an initial condition in analyses which assure that containment
internal pressure will not exceed the containment design values
during a LOCA or MSLB. Neither a LOCA nor a MSLB occurring during
Cold Shutdown would pressurize the containment. Therefore,
elimination of a Cold Shutdown LCO for containment pressure would
not affect the post-accident pressure or temperature. Since peak
post accident [pressure] and temperature would be unaffected by the
proposed change, operation of the Facility in accordance with
proposed change I.2 would not involve a significant reduction in a
margin of safety.
3. More Restrictive Changes: ``More Restrictive'' changes only
add new requirements, or revise existing requirements to result in
additional operational restrictions. The TS, with all ``More
Restrictive'' changes incorporated, will still contain all of the
requirements which existed prior to the changes. Therefore, ``More
Restrictive'' changes cannot involve a significant reduction in a
margin of safety.
4. ``Administrative'' changes make wording changes which clarify
existing TS requirements, without affecting their technical content.
Since ``Administrative'' changes do not alter the technical content
of any requirements, they cannot involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: John N. Hannon.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: October 29, 1997.
Description of amendments request: The proposed amendments to the
Technical Specifications (TS) for the Brunswick Steam Electric Plant
(BSEP) Units 1 and 2 would revise the description of the control rod
assemblies (CRAs) in TS 5.3.2. The proposed revision was requested to
support replacement of a portion of the BSEP Unit 1 CRAs during that
unit's next refueling outage with assemblies of a different design.
Carolina Power & Light Company, the licensee, has proposed adopting the
description of CRAs used in NUREG-1433, Revision 1, ``Standard
Technical Specifications General Electric Plants, BWR/4,'' which
includes the number and shape of CRAs and a stipulation that NRC-
approved absorber material be used in CRAs. The more detailed
description in the current TS of CRAs would be relocated to the Updated
Final Safety Analysis Report. The licensee has stated that the CRA
description proposed for TS 5.3.2 will be sufficient to ensure that any
future changes in CRA design that may affect safety will require prior
NRC review and approval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Relocation of the control rod assembly descriptive information
from the Technical Specifications to the Updated Final Safety
Analysis Report will ensure that adequate control of the information
is maintained. Any changes to this design information must conform
with the requirements of 10 CFR 50.59. Restricting use of control
rod assembly absorber materials to those listed, or to materials
that have been approved by the NRC, will ensure any changes which
may affect safety to require prior NRC review and approval. Since
the information with a potential to affect safety is sufficiently
addressed by the Technical Specifications, the criteria of 10 CFR
50.36(c)(4) for
[[Page 66138]]
including the relocated information as Design Features are not met.
Because the relocated information is not required to be in the
Technical Specifications to provide adequate protection of the
public health and safety, relocation of control rod assembly
descriptive information will not increase either the probability or
the consequences of an accident previously evaluated.
2. The proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Relocation, to the Updated Final Safety Analysis Report, of the
information pertaining to the control rod assembly designs ensures
that adequate control of the information will be maintained. Since
the information with a potential to affect safety is sufficiently
addressed by the Technical Specifications, the criteria of 10 CFR
50.36(c)(4) for including the relocated information as Design
Features are not met. Because the relocated information is not
required to be in the Technical Specifications to provide adequate
protection of the public health and safety, the proposed Technical
Specification changes to relocate the control rod assembly design
information to the Updated Final Safety Analysis Report does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
As discussed in Items 1 and 2 above, relocation of the control
rod assembly descriptive information from the Technical
Specifications to the Updated Final Safety Analysis Report will
ensure that adequate control of the information is maintained. Any
changes to this design information must conform with the
requirements of 10 CFR 50.59. Restricting use of control rod
assembly absorber materials to those listed, or to materials that
have been approved by the NRC, will ensure any changes which may
affect safety to require prior NRC review and approval. The
information with a potential to affect safety is sufficiently
addressed by the Technical Specifications, therefore, the proposed
Technical Specification changes to relocate control rod assembly
design information to the Updated Final Safety Analysis Report do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: James E. Lyons.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: July 16, 1997, as supplemented October
30, 1997.
Description of amendment request: The amendment would update
License condition 2.C(4) to reflect the latest revision levels of the
Oyster Creek Security Training and Qualification Plan, License
Amendment Request No. 252.
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:
GPU Nuclear has concluded that the proposed changes to the
Security Plan do not involve a significant hazard consideration. In
support of this determination, an evaluation of each of the three
standards set forth in 10 CFR 50.92 is provided below.
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Security Plan provisions are not associated with design basis
accident initiators nor do they constitute part of any mitigation
system. Therefore, the probability and consequences of accidents are
not increased.
(2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The Security Plan changes do not create new or change existing
physical interfaces with plant equipment. Therefore, the changes do
not create the possibility of a new or different kind of accident.
(3) The proposed changes do not involve a significant reduction
in a margin of safety.
Margins associated with reactor and fuel storage nuclear safety
are not affected by the proposed Security Plan changes since neither
physical nor procedural changes to associated systems, structures
and components are involved. Vital area security measures, which are
reduced, are compensated by commitments to hold contingency drills
at a frequency sufficient to maintain response capability for
response personnel and to use organic-type X-ray equipment.
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: Ronald B. Eaton, Acting Director.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: November 14, 1997.
Description of amendment request: The proposed change to Technical
Specification 4.5.2.d.1 will clarify the wording and increase the
setpoint for the open pressure interlock (OPI).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with 10
CFR 50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10 CFR 50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
Increasing the Technical Specification Open Pressure Interlock
(OPI) pressure to 412.5 psia [pounds per square inch--atmospheric]
will still maintain the required function of preventing the MOVs
[motor operated valves] from opening inadvertently. The increased
pressure is within the design limits of the RHR [residual heat
removal] piping system and components. The pressure signal is
generated from a transmitter and results in an electronic input to
the bistable. This is a clarification of the conditions under which
the OPI is tested.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
There is no change to the function of the OPI. The protection
provided by the interlock remains intact. The Technical
Specification OPI pressure has been raised to take into account
instrument accuracies and reset deadbands. The RHR system design
pressure remains protected from being exceeded by inadvertent
opening of the isolation MOVs. The method for the OPI surveillance
is clarified by clearly stating that the bistable receives a
simulated transmitter signal representative of the process pressure.
[[Page 66139]]
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The design pressure of the RHR system is 600 psig [pounds per
square inch--gauge]. The most limiting case is to prevent the RHR
pump developed head pressure from exceeding the design pressure when
aligned to the RCS [reactor coolant system] as suction pressure. RHR
pump testing has determined that a maximum pump differential
pressure of 195 psi [pounds per square inch] exists for deadhead/no
flow conditions. Therefore, to maintain the 600 psig design pressure
limit, RCS/suction pressure must be limited to 405 psig (420 psia,
assuming a 15 psi conversion from psig to psia). The proposed
maximum pressure, including setpoint tolerances and reset deadbands,
is less than this value; i.e. 412.5 psia. Head corrections due to
elevation differences are considered to be insignificant.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: Phillip F. McKee.
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: November 3, 1997.
Description of amendment request: The proposed amendment would
revise Sections 1, 3.1, 3.3, 4.3, and 6 of Appendix A of the Indian
Point 3 Technical Specifications. These revisions extend the Heatup-
Cooldown limits from 11 to 13 effective full power years (EFPYs),
provide the corresponding Overpressure Protection System (OPS) limits,
relocate the new pressure temperature limit curves and low-temperature
overpressurization protection (LTOP) system limits to the pressure
temperature limit report (PTLR) and include some minor revisions which
ensure specification clarity and conservatism.
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: The proposed license amendment does not involve a
significant increase in the probability or consequences of a
previously analyzed accident. The pressure-temperature limit changes
proposed by this amendment are based on supporting data and
evaluation methodologies previously submitted to the NRC in
Reference 3 [see application dated November 3, 1997] and approved as
Amendments 109 and 121 (References 4 and 5) [see application dated
November 3, 1997]. These limits are based upon the irradiation
damage prediction methods of Regulatory Guide 1.99, Revision 2. The
LTOPs changes contained in this submittal have been conservatively
adjusted in accordance with the new pressure-temperature limits, in
accordance with the methodology contained in Reference 3 and ASME
Code Case N-514.
The relocation of the pressure-temperature and LTOPs limits from
the Technical Specifications to the PTLR does not eliminate the
requirement to operate in accordance with the limits specified in 10
CFR [Part] 50, Appendix G. The requirement to operate within the
limits in the PTLR is specified in and controlled by the Technical
Specifications.
The revised version of Section 3.1.A.8 clarifies existing
requirements related to the OPS system and adds an eight hour
completion time for compensating actions, consistent with the STS
[standard technical specifications]. The changes to Section
3.1.A.1.h, i, and j revise the requirements associated with the
start of an RCP [reactor coolant pump]. These changes improve
specification clarity and do not increase the probability or
consequences of an accident.
The Technical Specification changes associated with the
restriction on SI [safety injection] pumps provides added
conservatism to the Technical Specifications and limits the
likelihood of an RHR [residual heat removal] overpressurization
event. Current plant procedures prohibit actuation of any SI pumps
when RHR is in service, except during testing, loss of RHR cooling,
or reduced inventory operations. Therefore, the change to the
Technical Specifications will not alter current plant operation.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously analyzed. The pressure-temperature limits are updating
the existing limits by taking into account the effects of radiation
embrittlement, utilizing criteria defined in Regulatory Guide 1.99,
Revision 2, and extending the effective period to 13 EFPYs. The
updated OPS limits have been adjusted to account for the effect of
irradiation on the limiting reactor vessel material. These changes
do not affect the way the pressure-temperature or OPS limits provide
plant protection and no physical plant alterations are necessary.
The relocation of the pressure-temperature and OPS limits from the
Technical Specifications to the PTLR does not alter the requirements
associated with these limits.
The revisions to Section 3.1.A.8 concerning the OPS system
improve on the clarity of existing specifications and add a
completion time for compensating actions that is consistent with the
STS. These changes do not involve any hardware modifications and do
not affect the function of the OPS system.
The revisions concerning the operation of SI pumps bring the
Technical Specifications into line with current operating
procedures. The changes to Specification 3.1.A.1.h, i, and j provide
specification clarity and are more conservative than existing
Technical Specifications. Therefore, the changes cannot create the
possibility of a new or different kind of accident.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: The proposed amendment does not involve a significant
reduction in a margin of safety. The margins of safety against
fracture provided by the pressure-temperature limits are those
limits specified in 10 CFR Part 50, Appendix G and ASME Boiler and
Pressure Vessel Code Section XI, Appendix G. The guidance in these
documents has been utilized to develop the pressure-temperature
limits with the requisite margins of safety for the heatup and
cooldown conditions. The new LTOP limits are based upon Reference 3
and ASME Code Case N-514. The relocation of the pressure-temperature
and OPS limits to the PTLR does not alter the requirements
associated with these limits.
The revisions to Section 3.1.A.8 clarify the requirements
associated with the OPS system. The revisions associated with the
operation of SI pumps with RHR in service (Sections 3.3.A.8, 9 and
10) and the changes regarding RCP starts (Section 3.1.A.1.h, i, and
j) are more conservative than the current Technical Specifications,
and are consistent with plant operating procedures. Therefore, they
do not reduce a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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,
[[Page 66140]]
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New
York, New York 10019.
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 amendment request: October 24, 1997.
Description of amendment request: The amendments would increase the
containment hydrogen analyzer surveillance frequency in Technical
Specification 4.6.4.1 from once per refueling outage to quarterly.
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 containment hydrogen analyzers provide control room
indication of hydrogen concentration in the containment atmosphere.
They do not affect the probability of any previously evaluated
accident. The proposed change would increase the calibration
frequency specified in TS 4.6.4.1 to make it consistent with
manufacturer's recommendations and the current calibration frequency
at [Salem Generating Station] SGS as imposed by administrative
controls. The change in TS-required calibration frequency is in the
conservative (more frequent) direction, to ensure that potential
degradation of the sensor electrolyte over time would not result in
unacceptable performance of the hydrogen analyzers. The change in
specified frequency would not adversely affect the consequences of
any previously evaluated accident.
2. Proposed change does not create the possibility of a new or
different kind of accident from any accident previously analyzed
[evaluated].
The proposed change affects only the specified calibration
frequency of the containment hydrogen analyzers. The proposed change
does not affect the design of any SGS structure, system or
component, nor would it result in any new plant configuration.
Therefore, it does not create the possibility of a new or different
kind of accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change to the containment hydrogen analyzer
calibration frequency does not affect the design or operating limits
of any SGS structure, system or component. The change would make the
specified calibration frequency more conservative, to ensure the
hydrogen analyzers perform as designed over time. The proposed
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.
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 amendment request: October 24, 1997.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3/4.7.7, ``Auxiliary Building Exhaust Air
Filtration System.'' The revisions would: (1) Require both Auxiliary
Building Ventilation (ABVS) supply fans to be operable, (2) require all
three ABVS exhaust fans to be operable, (3) align ABVS TSs to be
consistent with current TS bases and recently revised system
descriptions in the Salem Updated Final Safety Analysis Report (UFSAR),
(4) assure that negative pressure is maintained in the Auxiliary
Building under all postulated single active failures, (5) clarify
required Engineered Safety Feature filter testing, (6) provide
consistency between Unit 1 and Unit 2 TSs, and (7) for Unit 2 only,
remove the requirement to verify safety injection auto-start
capabilities.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change alters the number of fans which must be
OPERABLE to ensure that a sufficient number of supply and exhaust
fans will be operable, following a most limiting single failure, to
mitigate the consequences of design basis accidents. The changes to
the ABVS surveillance requirements still provide an appropriate
means for demonstrating the operability of the ABVS.
The ABVS cannot initiate or otherwise cause any accident or
operational transient evaluated in the UFSAR. Consequently, the
probability of such events is not increased. The ABVS cannot
increase the consequences of a design basis LOCA unless: (1)
Auxiliary Building negative pressure is lost, resulting in
uncontrolled, ground level release of radioactive material; (2) ABVS
carbon adsorbers are bypassed, resulting in uncontrolled release of
radioactive iodine from the plant vent; or (3) Auxiliary Building
temperatures are not controlled, resulting in failure of accident
mitigating equipment.
By requiring OPERABILITY of all ABVS supply and exhaust fans,
the proposed changes contained in this submittal assures Auxiliary
Building negative pressure is maintained under all postulated post-
accident, single-failure scenarios. The proposed changes to ABVS
will not affect the elemental iodine adsorption capability of the
system. Finally, engineering analyses conclude that these fan
combinations, with single-active failures of the fans or their
support systems considered, provide sufficient Auxiliary Building
ventilation. Under the most limiting temperature conditions, the
fans will maintain room temperatures within design limits.
Accordingly, the consequences of a design basis LOCA, hence
applicable design basis accidents or operational transients, are not
increased.
Therefore, the proposed amendment 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.
ABVS supply fans are not considered essential to the primary
safety-function of preventing or mitigating radioactive releases,
nor are they currently required to be OPERABLE. Similarly, accident
analyses take no credit for operation of supply fans. Accordingly,
malfunctions of vital buses and ABVS exhaust fans are the only
malfunctions of active ABVS related equipment important to safety
that are previously evaluated.
The probability of failure of a vital bus is not increased by
this proposal since the proposal has no direct effect on electrical
power. Neither is the probability of exhaust fan failure increased
by the proposal, since exhaust fans are not affected by this
proposal, except that the number that must be OPERABLE is increased
from two to three.
By requiring additional supply fans and exhaust fans to be
OPERABLE, no single failure of either a vital bus or ABVS fan
prevents (1) maintenance of negative Auxiliary Building pressure or
(2) maintenance of temperatures within design limits. Since ABVS
supply and exhaust fans cannot initiate accidents, increasing the
number of fans required to be OPERABLE cannot create the possibility
of a new or different kind of accident from any accident previously
evaluated. In addition, the proposed changes to the ABVS
surveillance testing concern ABVS leakage, HEPA filter and carbon
adsorber capabilities, and laboratory test methods. Therefore, the
proposed surveillance requirement changes would have no impact on
the initiation of accidents.
Thus, the proposed changes do not create the possibility of a
new or different kind of
[[Page 66141]]
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety is dependent upon the maintenance of
specific operating parameters within designated design limits. Since
iodine removal capability is not affected by the proposed changes,
and negative Auxiliary Building pressure and temperatures will
continue to be maintained within existing design limits under post-
accident conditions, including consideration of the most limiting
single active failure, the margin of safety is not reduced. By
imposing new restrictions on the allowed outage times of ABVS
components, the margin of safety is increased with the proposed
changes to the ABVS Technical Specification Limiting Condition for
Operation (LCO).
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: Salem Free Public library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.
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 amendment request: November 4, 1997.
Description of amendment request: The amendments would change
Technical Specification (TS) 3/4.6.2, ``Containment Spray System,'' to
verify on recirculation flow that the containment spray pumps develop a
differential pressure of at least 204 psi.
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 revises the CS [containment spray] pump
technical specification surveillance test acceptance from pump
discharge pressure to pump differential pressure. This will account
for the effect of RWST [refueling water storage tank] level on test
results and provide acceptance criteria that verifies each CS pump
performs as assumed in the accident analyses. This surveillance test
is also being added to the Salem Unit 1 TS. The proposed change does
not alter the physical plant arrangement or the method of CS pump
inservice testing. Therefore it does not increase the probability of
an accident. There is no change to pump performance requirements as
assumed in the accident analyses. There is no change to CS system
performance in response to an accident. Therefore, the proposed
change does not involve an increase in the consequences of an
accident previously evaluated.
The proposed change also corrects a typographical error by
removing a repeated word. This change does not involve an increase
in the consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change revises the Salem Unit 2 CS pump
surveillance test acceptance criteria from pump discharge pressure
to pump differential pressure. This will account for the effect of
RWST level on test results and provide acceptance criteria that
verify the CS pumps perform as assumed in the accident analyses.
This surveillance test is also being added to the Salem Unit 1 TS.
The proposed change does not alter the plant configuration. The
change does not alter the method of performing inservice testing on
the CS pumps. The change does not alter the CS pump performance
assumed in the accident analyses. Therefore, the change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
The proposed change also corrects a typographical error by
removing a repeated word. This change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change ensures the CS pump Salem Unit 2 TS
surveillance test acceptance criteria verify CS pump performance as
assumed in the accident analyses accounting for RWST level effects.
This surveillance test is also being added to the Salem Unit 1 TS.
The proposal does not change the CS pump performance requirements
assumed in the accident analyses and thus does not reduce the margin
of safety.
The proposed change also corrects a typographical error by
removing a repeated word. This 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: Salem Free Public library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.
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 amendment request: November 14, 1997.
Description of amendment request: The proposed changes to the
Technical Specifications (TSs) include administrative and editorial
changes to correct errors in the TSs that have either existed since
initial issuance or were introduced during subsequent changes. In
addition, surveillance requirements are added that are considered
administrative changes since the surveillances should have been
incorporated with the TS when the applicable amendment to the TSs was
approved by the NRC.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the TS are administrative or editorial
changes to the TS and do not involve any physical changes to the
plant. The administrative changes and editorial changes do not
delete any existing surveillance requirements or delete any
requirements from the Limiting Condition for Operations (LCOs) or
Action Statements and therefore do not reduce the actions that are
currently taken in the TS to demonstrate operability of plant
structures, systems, or components (SSCs). The additional
surveillance requirements that are being added to the TS including
the new surveillances correct past administrative errors and should
have been incorporated within the TS as part of the approved
Amendments to the TS. These changes will provide additional
assurance that SSCs perform their intended safety functions.
Surveillance testing has been and is currently being performed for
the surveillance requirements that should have been incorporated and
are now administratively being added to the TS. Since these changes
do not modify any SSCs or reduce the current requirements for
demonstrating operability of these SSCs or reduce the current
requirements for demonstrating operability of these SSCs, the
proposed changes to the TS do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
[[Page 66142]]
The proposed changes to the TS are administrative and editorial
corrections to the TS that do not affect the ability of the plant
systems to meet their current TS requirements or design basis
functions. There is no reduction in the current surveillance
requirements required to demonstrate the operability of plant SSCs.
These changes also do not involve any physical changes to plant
SSCs. Therefore the proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes are administrative and editorial
corrections to the TS that do not affect the ability of plant SSCs
to perform their design basis accident functions. There is no
reduction in the current surveillance requirements required to
demonstrate the operability of plant SSCs. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.
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 amendment request: November 14, 1997.
Description of amendment request: The proposed changes to the
Technical Specifications (TSs) include administrative and editorial
changes to correct errors in the TSs that have either existed since
initial issuance or were introduced during subsequent changes. In
addition, surveillance requirements are added that are considered
administrative changes since the surveillances should have been
incorporated with the TS when the applicable amendment to the TSs was
approved by the NRC.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the TS are administrative or editorial
changes to the TS and do not involve any physical changes to the
plant. The administrative changes and editorial changes do not
delete any existing surveillance requirements or delete any
requirements from the Limiting Condition for Operations (LCOs) or
Action Statements and therefore do not reduce the actions that are
currently taken in the TS to demonstrate operability of plant
structures, systems, or components (SSCs). The additional
surveillance requirements that are being added to the TS including
the new surveillances correct past administrative errors and should
have been incorporated within the TS as part of the approved
Amendments to the TS. These changes will provide additional
assurance that SSCs perform their intended safety functions.
Surveillance testing has been and is currently being performed for
the surveillance requirements that should have been incorporated and
are now administratively being added to the TS. Since these changes
do not modify any SSCs or reduce the current requirements for
demonstrating operability of these SSCs or reduce the current
requirements for demonstrating operability of these SSCs, the
proposed changes to the TS do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to the TS are administrative and editorial
corrections to the TS that do not affect the ability of the plant
systems to meet their current TS requirements or design basis
functions. There is no reduction in the current surveillance
requirements required to demonstrate the operability of plant SSCs.
These changes also do not involve any physical changes to plant
SSCs. Therefore the proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes are administrative and editorial
corrections to the TS that do not affect the ability of plant SSCs
to perform their design basis accident functions. There is no
reduction in the current surveillance requirements required to
demonstrate the operability of plant SSCs. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.
Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear
Generating Station, Unit No. 2, Salem County, New Jersey
Date of amendment request: October 29, 1997.
Description of amendment request: The proposed amendment would make
a one-time change to Technical Specification 3/4.4.6, ``Steam
Generators,'' to require that the next inspection be performed within
24 months of criticality for fuel cycle 10, rather than within 24
months from the previous inspection. The previous inspection was
performed in May 1996; thus, adhering to the current Technical
Specification would require inspection by May 1998 and would require a
forced outage. It would also eliminate description of an alternate
sampling plan that was applicable only to Unit 2's fourth refueling
outage.
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 Design Basis Accident (DBA) analyzed in UFSAR Chapter
15.4.4, is Steam Generator Tube Rupture. The Technical Specification
steam generator tube inspection attempts to avoid this DBA by
maintenance of the integrity of the primary to secondary coolant
boundary represented by steam generator tubes. The process by which
this integrity is maintained is inspection of steam generator tubes
at prescribed intervals, and the removal of defective tubes from
service. Inspection intervals are based on preventing corrosion
growth from exceeding tube structural strength, thereby preventing
tube failure. An extensive steam generator inspection in May of 1996
characterized existing steam generator tube degradation, and
degraded tubes were removed from service at that time. Degradation
growth rates were evaluated for the next operating interval and it
was determined that full cycle operation would not challenge tube
structural integrity. Because degraded tubes were plugged, the
integrity of the steam generators has been restored, and, because
further degradation was prevented by a strictly controlled wet lay-
up program in place since the inspection, steam generator integrity
has since been maintained at the May 1996 level. This is the level
normally expected for commencement of full power operations at the
beginning of a fuel cycle. Thus, it can be reasonably
[[Page 66143]]
concluded that this request to extend the inspection interval to
conclude 24 months after the start of Unit 2 fuel cycle 10 does not
involve an increase in the probability of an accident previously
analyzed.
Salem UFSAR Chapter 15, Section 15.4.4., discusses the Design
Basis Accident involving steam generator tube rupture. Since the
Salem Unit 2 steam generators were extensively inspected and all
degraded tubes were removed from service by plugging, integrity of
the generators was restored to fully serviceable condition at that
time. Degradation of steam generator tubes has been prevented since
the inspection by a carefully controlled, EPRI Guidelines based,
corrosion prevention program. It follows, then, that the Unit 2
steam generators were in the same condition immediately prior to
fill and vent as if the inspection had just been concluded. This is
the condition assumed for commencement of normal operation. Thus, it
is reasonable to conclude that this proposal to extend the current
steam generator inspection interval to end 24 months after start of
Unit 2 fuel cycle 10 represents no significant increase in the
consequences of an accident previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Steam generator tube inspections determine tube integrity and
provide reasonable assurance that a tube rupture or primary to
secondary leak will not occur. Accidents involving steam generator
tube rupture are analyzed in Salem UFSAR Section 15.4.4, Steam
Generator Tube Rupture. The only type of accident that can be
postulated from extending the steam generator inspection interval
would be a tube leak or rupture. Thus, it can be concluded that
extending the steam generator inspection interval on a one-time
basis cannot create the possibility of a 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 margin of safety, as with any TS, depends upon maintenance
of specific operating parameters within design limits. In the case
of steam generators, that margin is maintained through assurance of
tube integrity as the primary to secondary boundary. Assurance of
tube integrity is provided through periodic inservice testing of
tube integrity and removal from service of defective tubes.
Additional margin is provided through protection from possible
consequences of steam generator tube failure by detection and
mitigation systems. As discussed in 1., above, there was an
extensive steam generator inspection, and the steam generators have
been maintained since the inspection, using a lay-up program that
complies with EPRI Guidelines, to prevent further tube degradation.
Also, N-16 monitors were added, enhancing detection capabilities.
The margin as established by the latest inspection has been
maintained by the corrosion control program of EPRI Primary and
Secondary Guidelines based on wet lay-up conditions. Thus, it can be
reasonably concluded that this proposal to amend the Salem Unit 2
Technical Specifications, on a one-time basis, to extend the steam
generator inspection interval to end 24 months after start of Unit 2
fuel cycle 10 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: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: August 8, 1997.
Description of amendment request: The proposed amendment would
revise the surveillance requirements (SR) of Technical Specification
(TS) 3/4.7.4 ``Essential Service Water System'' by removing the
requirement to perform SR 4.7.4.b.1, 4.7.4.b.2 and 4.7.4.c during
shutdown.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change to TS has no adverse impact on the
probability of occurrence or the consequences of an accident. The
proposed amendment does not change or alter the design assumptions
for the systems or components used to mitigate the consequences of
an accident and the methodologies used in the accident analysis
remain unchanged. The operating limits and the radiological
consequences will not be changed. No design basis accidents will be
affected by this change since the required TS surveillances will
continue to be performed on an 18 month frequency.
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.
All design and performance criteria continue to be met and no
new failure mechanisms have been identified. The proposed change
does not affect the design or operation of any system or component
in the plant since the required TS surveillances will continue to be
performed on an 18 month frequency. The safety functions of the
related structures, systems or components are not changed in any
manner, nor is the reliability of any structure, system or component
reduced. Conducting these surveillances online will not increase the
possibility of plant transients. Since the safety functions and
reliability are not adversely affected, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will not affect or change a safety limit or
affect plant operations since the required TS surveillances will
continue to be performed on an 18 month frequency. This change will
not reduce the margin of safety assumed in the accident analysis nor
reduce any margin of safety as defined in the basis for any TS.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: August 8, 1997.
Description of amendment request: The proposed amendment would
revise Table 3.3-3, Functional Units 4.b.2 and 5.a.2 of the Callaway
Technical Specifications (TS) by (1) changing the main steam and
feedwater isolation system (MSFIS) channels to be consistent with the
requirements for the solid state protection system (SSPS), (2) adding a
clarifying note, and (3) deleting and replacing Action Statements 27a
and 34a with Action Statements 27 and 34. In addition, Table 4.3-2,
Functional Units 4.b and 5.a are proposed to be revised by changing the
slave relay quarterly surveillance to a quarterly actuation logic test
for the MSFIS actuation and relays.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the
[[Page 66144]]
issue of no significant hazards consideration, which is presented
below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to Technical Specifications (TS) have no
adverse impact on the probability of occurrence or the consequences
of an accident. The proposed amendment does not change or alter the
design assumptions for the systems or components used to mitigate
the consequences of an accident and the methodologies used in the
accident analysis remain unchanged. The operating limits and the
radiological consequences will not be changed. No design basis
accidents will be affected by these changes. The proposed changes do
not result in any hardware changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in parameters governing normal plant operation. All design
and performance criteria continue to be met and no new failure
mechanisms have been identified. The proposed changes do not affect
the design or operation of any system or component in the plant. The
safety functions of the related structures, systems or components
are not changed in any manner, nor is the reliability of any
structure, system or component reduced. However, these changes are
consistent with the requirements for the SSPS. Since the safety
functions and reliability are not adversely affected, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes will not affect or change a safety limit or
affect plant operations. These changes will not reduce the margin of
safety assumed in the accident analysis nor reduce any margin of
safety as defined in the basis for any TS. The proposed changes do
not affect the acceptance criteria for any analyzed event. No
setpoints are revised and the system response time will not be
affected.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: August 8, 1997.
Description of amendment request: The proposed amendment would
revise Table 3.7-2 of the Technical Specifications to specify that the
lift setting tolerance for the main steam line safety valves be +3/-1%
as-found and plus or minus 1% as-left. Table 2.2-1 would be revised by
reducing the sensor error for the pressurizer pressure-high trip.
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 main steam line safety valves are designed to mitigate
transients by preventing overpressurization of the main steam
system. The proposed change does not alter this design basis. The
revised analysis shows that the probability or consequences of all
previously analyzed accidents are not changed by increasing the
setpoint tolerance of the safety valves. Therefore, there is no
increase in the probability of occurrence or the consequences of any
accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There is no new type of accident or malfunction created, the
method and manner of plant operation will not change nor is there a
change in the method in which any safety related system performs its
function. Any main steam safety valve lifting at the extremes of the
proposed tolerance will not result in a low lift setpoint that is
less than the normal no load system pressure or a high lift setpoint
that allows main steam system overpressurization.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This is based on the fact that no plant design changes are
involved and the method and manner of plant operation remains the
same. With the increased setpoint tolerance, the main steam safety
valves will still prevent pressure from exceeding 110 percent of
design pressure in accordance with the ASME code. All FSAR accident
analysis conclusions remain valid and unaffected by this change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: August 8, 1997.
Description of amendment request: The proposed amendment
application would revise feedwater isolation engineered safety feature
actuation system (ESFAS) functions in Technical Specification Tables
3.3-3, 3.3-4 and 4.3-2 as follows:
(1) The Applicable MODES for Functional Units 5.a.1), Automatic
Actuation Logic and Actuation Relays, and 5.a.2), Automatic Actuation
Logic and Actuation Relays, in Tables 3.3-3 and 4.3-2 would be revised
to add MODE 3.
(2) A new Functional Unit 5.d, Steam Generator (SG) Water Level
Low-Low (for feedwater isolation only), would be added to Tables 3.3-3,
3.3-4, and 4.3-2.
(3) In conjunction with the changes under item (2), the Applicable
MODES in Table 3.3-3 for AFW SG Water Level Low-Low Functional Units
6.d.1).c), Start Motor-Driven Pumps Vessel delta T (Power-1, Power-2),
would be revised to delete MODE 3. Functional Unit 6.d.3) in Table 4.3-
2 would also be revised to delete MODE 3.
(4) The Bases for Functional Unit 11.b, Reactor Trip P-4, in Table
3.3-3 would be revised to add a note allowing the feedwater isolation
function on P-4 coincident with low Tavg to be blocked.
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:
Actuation Logic Applicability and New SG Water Level Low-Low Functional
Unit
1. The proposed change does not involve a significant increase
in the probability or
[[Page 66145]]
consequences of an accident previously evaluated.
The proposed changes impose more stringent requirements and have
been reviewed to ensure no previously evaluated accident has been
adversely affected. The more stringent requirements are imposed to
ensure the plant's operation and testing are consistent with the
safety analysis and licensing basis. Therefore, the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed other
than the bypass switch addressed in a separate 50.92 evaluation
below) or changes in controlling parameters. The proposed changes do
impose different requirements; however, these changes are consistent
with assumptions made in the safety analysis and licensing basis.
Actuation logic applicability is extended to MODE 3 and the SSPS
slave relays that implement feedwater isolation on SG water level
low-low will continue to be surveilled quarterly as they have always
been tested. Thus, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The imposition of more stringent requirements does not reduce
the margin of safety. The margin of safety would be increased since
the scope of the Technical Specifications has been increased to
include additional plant equipment and add additional Applicability
requirements. The changes are consistent with the safety analysis
and licensing basis. Therefore, the proposed changes do not involve
a reduction in a margin of safety.
TTD Applicability
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The proposed change adds a relaxation
to the Applicability for the SG Water Level Low-Low Vessel delta T
channels. The proposed change in the Applicability will not affect
any of the analysis assumptions for any of the accidents previously
evaluated. The proposed change will not affect the probability of
any event initiators nor will the proposed change affect the ability
of any safety-related equipment to perform its intended function. A
Vessel delta T channel should only be tripped if it is inoperable
and the reactor is operating, when the need to restrict trip time
delays is applicable. There will be no degradation in the
performance of nor an increase in the number of challenges imposed
on safety-related equipment assumed to function during an accident
situation. Accident analyses have been performed with the maximum
trip time delays enabled at power levels up to 19% RTP (10% RTP plus
uncertainty). Therefore, operation in MODE 3 with the maximum trip
time delays is enveloped. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. The change in Applicability will not impact the normal
method of plant operation. The maximum trip time delay should be
enabled in MODE 3 to preclude an unnecessary feedwater isolation or
auxiliary feedwater actuation from occurring prior to the expiration
of the trip time delay previously analyzed for MODE 1 operation. No
new accident scenarios, transient precursors, failure mechanisms, or
limiting single failures are introduced as a result of this change.
Therefore, the proposed change does not create the possibility of a
new of different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not affect the acceptance criteria for
any analyzed event. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection functions. There will be no impact
on any margin of safety.
Feedwater Isolation on P-4/Low Tavg Bypass Switch
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses. The P-4/Low
Tavg Bypass Switch design change will not impact any
accidents previously evaluated in the FSAR since feedwater isolation
upon reaching this function was never credited.
The ESFAS will continue to function in a manner consistent with
the accident analysis assumptions and the plant design basis. As
such, there will be no degradation in the performance of nor an
increase in the number of challenges to equipment assumed to
function during an accident situation.
This Technical Specification change does not affect the
probability of any event initiators. There will be no change to
normal plant operating parameters or accident mitigation
capabilities. Therefore, there will be no increase in the
probability or consequences of any accident occurring due to this
change.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no changes in the method by which any safety-related
plant system performs its safety function and the normal manner of
plant operation is unaffected, other than the proposed allowance to
bypass feedwater isolation on P-4 coincident with low
Tavg. This bypass switch modification will be performed
under the design standards applicable to all safety system bypasses
at Callaway, except for Section 4.12 of IEEE 279-1971. Section 4.12
of IEEE 279-1971 requires that an operating bypass of a protective
function be automatically removed whenever permissive conditions are
not met. However, the subject circuitry does not provide a
protective function. It is not assumed or credited in any safety
analysis. In addition, plant conditions that would call for the
restoration of the feedwater isolation function cannot occur without
operator action to close the reactor trip breakers. Administrative
controls will govern the proper use of and restoration from the
proposed bypass. Although the addition of the bypass switch
introduces the potential for an equipment malfunction of a different
type from any previously evaluated in the FSAR, the possibility of a
new or different type of accident is not created. The switch
functions only to allow a manual bypass of feedwater isolation. The
failure of the switch or its improper use will not be an event
initiator for the previously analyzed Loss of Normal Feedwater event
in FSAR Section 15.2.7 since it cannot fail in such a manner as to
cause feedwater isolation.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this change.
Therefore, the possibility of a new or different type of accident is
not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on DNBR limits,
FQ, F-delta-H, LOCA PCT, peak local power density, or any
other margin of safety.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
[[Page 66146]]
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: November 5, 1997.
Description of amendment request: The current Technical
Specifications requirements prohibit loads in excess of 2500 pounds
from traveling over irradiated fuel assemblies in the spent fuel pit.
Due to the number of irradiated fuel assemblies currently stored in the
spent fuel pit over years of operation, additional flexibility is
needed to accomplish the movement of the spent fuel pit gates during
refueling activities and to reduce fuel handling activities in
preparation for refueling outages. In order to perform gate seal
maintenance prior to each outage, a gate is moved across the irradiated
fuel storage area to the cask handling area where it can be lifted out
of the spent fuel pit. When a clear path of empty fuel storage cells
cannot be established, seal maintenance cannot be performed unless
relief from the current Limiting Condition of Operation is granted. The
proposed changes will exempt these requirements for the movements of
the spent fuel gates provided specific administrative controls are
satisfied.
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:
Specifically, operation of the North Anna Power Station in
accordance with the proposed changes will not:
1. Involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated.
The accident in question is a fuel handling accident in the
spent fuel pit. The proposed changes will actually reduce the
probability of a fuel handling accident by eliminating unnecessary
fuel assembly movements. After this change is implemented, only
those assemblies containing control rod assemblies will be subjected
to such moves prior to movement of the gates instead of the current
practice of moving all the fuel necessary to establish a load path
of empty cells. A redundant rigging system will be provided which
eliminates the possibility of a load drop due to a hoist failure.
Furthermore, even though the double rigging system makes a load drop
due to a hoist failure an incredible event, a calculation was
performed to determine the effects of a direct impact load on a
single fuel storage cell or the SFP [spent fuel pit] structure. The
calculation concludes that there will be no adverse consequences to
either irradiated fuel or the SFP structure. The plant design basis
fuel handling accident will not be violated. Therefore, with the
administrative controls in place to eliminate the possibility of a
gate drop the probability of occurrence or the consequences of a
fuel handling accident are not increased.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes establish adequate administrative controls
over the spent fuel pit gate movements to prevent damage to stored
irradiated fuel and fuel racks thereby ensuring the design basis
fuel handling accident remains bounding and that fuel spacing is
maintained in the racks precluding criticality.
3. Involve a significant reduction in any margin of safety.
The new administrative controls ensure that a postulated gate
drop will not occur due to compliance with our licensing commitments
to NUREG-0612 and the requirement to install a redundant rigging
system to eliminate the possibility of a load drop initiated by
hoist failure. Analysis has determined that in the event the gate
was to be dropped from its controlled lift height: (1) There will be
no damage to irradiated fuel caused by the direct impact loading on
a single storage cell and (2) the fuel storage rack will maintain
fuel in a non-critical array. A new criteria, demonstrating the
ability of the pool floor to remain intact after a gate drop has
been shown by analysis. New controls prevent the degradation of the
existing margin of safety and ensure an adequate safety margin for
the new criteria. The administrative controls added for the gate
lift preclude the possibility of a load drop induced by a hoist
failure and, therefore ensure the potential for radioactivity
release and inadvertent criticality remain bounded by the present
design basis. Therefore, the margin of safety is not reduced by the
proposed change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: James E. Lyons.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: November 18, 1997.
Description of amendment request: The Technical Specifications
surveillance requirements currently require testing and inspection of
the Turbine Overspeed Protection System control valves, at least once
per 31 days, to ensure their ability to prevent overspeeding of the
turbine. Based on an analysis of Westinghouse BB-296 turbines with
steam chests, the proposed change would increase the surveillance test
interval from at least once per 31 days to at least once per 92 days.
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:
Specifically, operation of the North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No new or unique accident precursors are introduced by these
changes in surveillance requirements. The probability of turbine
missile ejection with an extended test interval to 92 days for the
turbine governor and throttle valves has been determined to remain
within the applicable NRC acceptance criteria. The heavy hub design
of the turbine rotors provides further assurance that the
probability of ejection of turbine missiles due to destructive
overspeed remains within the acceptance criteria. Therefore, these
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The demonstrated high reliability of the turbine governor and
throttle valves and the verification of the operability of the other
turbine control valves provide adequate assurance that the turbine
overspeed protection system will operate as designed, if needed.
Turbine governor and throttle valve testing performed to date has
demonstrated the reliability of these valves. In addition, the
operability of the other turbine valves (i.e., reheat and intercept
stop valves) will continue to be verified every 18 months as
required by the Technical Specifications.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Since the implementation of the proposed change to the
surveillance requirements will not require hardware modifications
(i.e., alterations to plant configuration), operation of the
facilities with these proposed Technical Specifications does not
create the possibility for any new or different kind of accident
which has not been already been evaluated in the Updated Final
Safety Analysis Report (UFSAR). In addition, the results of the
probabilistic evaluation indicate that no additional transients have
been introduced.
The proposed revision to the Technical Specifications will not
result in any physical alteration to any plant system, nor would
there be a change in the method by which any safety-related system
performs its function. The design and operation of the
[[Page 66147]]
turbine overspeed protection and turbine control systems are not
being changed.
The proposed Technical Specifications changes do not affect the
design, operation, or failure modes of the valves and other
components of the turbine overspeed protection system. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes do not reduce the margin of safety as
defined in the basis for any Technical Specifications. Furthermore,
the total turbine missile ejection probability continues to be
enveloped by the applicable acceptance criteria of 1E-5. The design
and operation of the turbine overspeed protection and turbine
control systems are not being changed and the operability of the
turbine governor and throttle valves will be demonstrated on a
refuelling outage basis. In addition, the results of the accident
analyses, which are documented in the UFSAR, continue to bound
operation with the proposed change in surveillance interval for the
turbine throttle and governor valves, so that there is no safety
margin reduction. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: James E. Lyons.
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.
Tennessee Valley Authority, Docket Nos. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendments: November 21, 1997.
Description of amendments request: Amend Technical Specifications
to add a one-time allowance through Operating Cycle 9 to Surveillance
Requirement 4.4.3.2.1.b to perform stroke testing of the power-operated
relief valve in Mode 5 rather than in Mode 4.
Date of publication of individual notice in the Federal Register:
December 1, 1997 (62 FR 63565).
Expiration date of individual notice: December 31, 1997.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of application for amendments: November 30, 1995, as
supplemented March 15, 1996, March 6, 1997, and June 27, 1997.
Brief description of amendments: The amendments incorporate
references to a new Combustion Engineering, Inc. topical report
describing steam generator tube sleeves, delete references to the
previous CE topical report, incorporate sleeve/tube inspection scope
and expansion criterion, revise the plugging limit for a CE sleeve to
28% of the nominal sleeve wall thickness, and incorporate a post weld
heat treatment for free span welds.
Date of issuance: November 18, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 223 and 199.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
176). The March 15, 1996, March 6, 1997, and June 27, 1997, letters
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated November 18, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: August 6, 1997.
Brief description of amendments: The amendments address an
unreviewed safety question associated with the handling of the spent
fuel shipping cask at the Brunswick Steam Electric Plant, Units 1 and
2.
Date of issuance: December 2, 1997.
Effective date: December 2, 1997.
[[Page 66148]]
Amendment Nos.: 190 and 221.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
authorize changes to the facility's Updated Final Safety Analysis
Report.
Date of initial notice in Federal Register: September 17, 1997 (62
FR 48897) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 2, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: February 18, 1997.
Brief description of amendment: This amendment revises the maximum
allowable power range neutron flux high setpoints (percent of rated
thermal power) shown in Technical Specification Table 3.7-1.
Date of issuance: November 25, 1997.
Effective date: November 25, 1997.
Amendment No.: 75.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17225) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 25, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendments: April 7, 1997, as supplemented
on August 7, 1997.
Brief description of amendments: The amendments revise the
technical specifications to permit installation and use of C&D Charter
Power Systems, Inc., batteries.
Date of issuance: November 25, 1997.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 93 and 93.
Facility Operating License Nos. NPF-37 and NPF-66: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54868). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 25, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: October 13, 1997, as
supplemented by letters dated October 28 and November 5, 1997.
Brief description of amendments: The amendments revise TS Table
3.3-4, ``Engineered Safety Features [ESF] Actuation System Instrument
Trip Setpoints.'' Specifically, the amendments support the replacement
of three safety-related narrow range Refueling Water Storage Tank level
instruments with three safety-related wide range level instruments. The
ESF trip setpoint for the refueling water automatic switchover to
recirculation is revised to account for the difference in instrument
uncertainty associated with wide range level instruments and provides
additional operator response time margin.
Date of issuance: November 25, 1997.
Effective date: Unit 1--As of the date of issuance to be
implemented consistent with the refueling outage scheduled for June
1998; Unit 2--As of the date of issuance to be implemented within 30
days from the date of issuance.
Amendment Nos.: 177 (Unit 1); 159 (Unit 2).
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54859). The October 28 and November 5, 1997, letters provided
additional and clarifying information that did not change the scope of
the October 13, 1997, application and the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 25, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application for amendments: October 10, 1997, as
supplemented by letters dated November 3, 6, and 10, 1997.
Brief description of amendments: The amendments revise Technical
Specifications to implement alternate repair criteria for steam
generator tubes that have degraded roll joints inside of the upper
tubesheet. The alternate repair criteria would allow new roll joints to
be installed below the degraded roll joints in the upper tubesheet.
Date of issuance: November 21, 1997.
Effective date: November 21, 1997.
Amendment Nos.: Unit 1--227; Unit 2--227; Unit 3--224.
Facility Operating License Nos. DPR-38, DPR-47, AND DPR-55: The
amendments revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. (62 FR 55835 dated October 28, 1997). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by November 28, 1997, but indicated that if the Commission
makes a final no significant hazards consideration determination, any
such hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and a final determination of no significant
hazards consideration are contained in a Safety Evaluation dated
November 21, 1997.
Attorney for licensee: M. J. Michael McGarry, III, Winston and
Strawn, 1200 17th Street, NW., Washington, DC.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power
Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: March 10, 1997, as supplemented
July 28 and September 17, 1997.
Brief description of amendment: The amendment modifies Technical
Specification 3/4.4.5, ``Steam Generators,'' and its associated Bases
[[Page 66149]]
and adds a new license condition to Appendix C for Beaver Valley Power
Station, Unit No. 1 (BVPS-1) to allow repair of steam generator tubes
by installation of sleeves developed by ABB Combustion Engineering. In
addition, the amendment deletes the option for using the kinetic
sleeving methodology previously approved for use at BVPS-1.
Date of issuance: November 25, 1997.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No.: 208.
Facility Operating License No. DPR-66: Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19829). The July 28 and September 17, 1997, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the amendment request
beyond the scope of the April 23, 1997, Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 25, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: September 12, 1997, as
supplemeneted November 7, 1997.
Brief description of amendment: The proposed amendment involves a
revision to the Emergency Diesel Generator protective relaying scheme
at Crystal River Unit 3, to be reflected in the next revision to the
Final Safety Analysis Report (FSAR).
Date of issuance: December 1, 1997.
Effective date: Effective upon issuance.
Amendment No.: 159.
Facility Operating License No. DPR-72:. Amendment revises the FSAR.
Date of initial notice in Federal Register: September 30, 1997 (62
FR 51165). By letter dated November 7, 1997, the licensee provided
additional information which did not affect the original no significant
hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 1, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: October 10, 1996, as
supplemented March 25, June 6, and August 29, 1997.
Brief description of amendment: The amendment extends the
instrumentation surveillances for the condenser low vacuum, high
temperature main steamline tunnel, recirculation flow, and reactor
coolant leakage. Additionally, the change extends the equipment test/
operability checks for containment vent and purge isolation,
electromagnetic relief valve operability, and drywell to torus leakage
test.
Date of Issuance: November 26, 1997.
Effective date: November 26, 1997, with full implementation within
60 days.
Amendment No.: 193.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 6, 1996 (61 FR
57485). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated November 26, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit 2, New London County, Connecticut
Date of application for amendment: May 20, 1997, as supplemented on
September 23, 1997.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) by relocating the containment isolation valve
(CIV) list from the TSs to the Technical Requirements Manual in
accordance with Generic Letter 91-08, ``Removal of Component Lists from
the Technical Specifications.'' The amendment also changes the
surveillance requirement for valves, blind flanges, and deactivated
automatic valves located inside containment that are locked, sealed, or
otherwise secured in the closed position from once every 31 days to
during each cold shutdown, but no more than once per 92 days. The TS
Bases is changed to reflect the relocation of the containment isolation
valve list from the TSs to the Technical Requirements Manual and
dicusses administrative controls for CIV operation in Modes 1 through
4. Also, a license condition has been added to paragraph 2.C. of the
Operating License to ensure enforceability and to provide a method of
tracking the license condition back to the license amendment.
Date of issuance: November 19, 1997.
Effective date: As of the date of issuance, to be implemented
within 90 days.
Amendment No: 210
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications and License Conditions.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33128). The September 23, 1997, letter provided clarification of the
initial submittal and did not affect the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated November 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut.
Date of application for amendment: September 16, 1997.
Brief description of amendment: The amendment changes the main
steam line American Society of Mechanical Engineers Code (Code) safety
valves Technical Specifications (TSs) by: (1) Deleting TS Table 3.7.1,
``Maximum Allowable Power Level-High Trip Setpoint with Inoperable
Steam Line Safety Valves During Operation with Both Steam Generators,''
by not allowing operation in Mode 1 or 2 with inoperable Code safety
valves while allowing operation in Mode 3 with up to three Code safety
valves inoperable per steam generator, (2) modifing the associated
action statement in TS 3.7.1.1 to reflect the operational changes, and
(3) updating the TS Bases to reflect the proposed changes and include
the correct amendment history numbers to
[[Page 66150]]
reflect previously approved amendments.
Date of issuance: November 19, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 211.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 8, 1997 (62 FR
52582). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: September 26, 1997.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.4.B, ``Auxiliary Feedwater System,'' to provide
specific guidance for conducting post-maintenance operational testing
of the turbine-driven auxiliary feedwater pump and associated system
valves to meet operability requirements and limiting conditions for
operation during unit startup. Additionally, the amendments revise
Table TS.3.5.2B to allow the auxiliary feedwater pump auto-start
actuation instrumentation to be bypassed during startup and shutdown
operations when the main feedwater pumps are not required to supply
feedwater to the steam generators.
Date of issuance: November 25, 1997.
Effective date: November 25, 1997, with full implementation within
30 days.
Amendment Nos.: 134 and 126.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54874). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 25, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 20, 1996.
Brief description of amendment: The amendment revises the technical
specifications to correct and clarify surveillance test requirements
for the reactor protective system and other plant instrumentation and
control systems.
Date of issuance: November 24, 1997.
Effective date: November 24, 1997, to be implemented within 120
days of the date of issuance.
Amendment No.: 182.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44361). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 24, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear
Generating Station, Unit No. 1, Salem County, New Jersey
Date of application for amendment: May 10, 1996, as supplemented
March 19 and August 29, 1997.
Brief description of amendment: The amendment incorporates into the
Technical Specifications the Margin Recovery portion of the Fuel
Upgrade Margin Recovery Program and support increased steam generator
plugging, improved fuel reliability, reduced fuel costs, longer fuel
cycles, reduced spent fuel pool storage, and enhanced reactor safety.
Date of issuance: November 26, 1997.
Effective date: As of date of issuance. To be implemented on Unit 1
prior to entry into Mode 2 from the current outage.
Amendment No.: 201.
Facility Operating License No. DPR-70: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34898). The March 19 and August 29, 1997, 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
November 26, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: January 4, 1996.
Brief description of amendments: These amendments delete License
Condition 2.C(26) for SONGS Unit 2 and License Condition 2.C(27) for
SONGS 3. These license conditions require that Southern California
Edison implement and maintain a plan for scheduling all capital
modifications based on an NRC approved Integrated Implementation
Schedule Program Plan.
Date of issuance: December 3, 1997.
Effective date: December 3, 1997.
Amendment Nos.: Unit 2--137; Unit 3--129.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15997). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 3, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: January 20, 1997.
Brief description of amendment: This amendment revises Technical
Specification (TS) Section 3/4.5.2, ``Emergency Core Cooling Systems,
ECCS Subsystems-Tavg greater than or equal to 280 deg.F,''
TS Section 3/4.5.3, ``Emergency Core Cooling Systems, ECCS
Subsystems-Tavg less than 280 deg.F,'' and TS Section 3/
4.7, ``Plant Systems.'' Several surveillance intervals were changed
from 18 months to once each refueling interval.
Date of issuance: December 2, 1997.
[[Page 66151]]
Effective date: December 2, 1997.
Amendment No.: 216
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 12, 1997 (62 FR
11498). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 2, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: September 17, 1996, as
supplemented by letters dated November 27, 1996, and October 14, 1997.
Brief description of amendment: This amendment revises the
surveillance interval from 18 months to less than or equal to 730 days,
nominally 24 months, for Technical Specification (TS) 3/4.5.2,
``Emergency Core Cooling Systems--ECCS Subsystems--Tavg
greater than or equal to 280 degrees F''; TS 3/4.6.5.1, ``Containment
Systems--Shield Building--Emergency Ventilation System''; TS 3/4.7.6.1,
``Plant Systems--Control Room Emergency Ventilation System''; TS 3/
4.7.7, ``Plant Systems--Snubbers''; TS 3/4.9.12, ``Refueling
Operations--Storage Pool Ventilation''; and TS Bases 3/4.7.7--
``Snubbers.''
Date of issuance: December 2, 1997.
Effective date: Immediately, and shall be implemented no later than
120 days after issuance.
Amendment No.: 217.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52972). The supplemental information submitted by the licensees did not
impact the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 2, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
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: December 11, 1996 (as
supplemented by letter dated January 6, 1997), January 30, 1997 (as
supplemented by letter dated September 15, 1997), and April 18, 1997.
Brief description of amendment: This amendment extends surveillance
requirement intervals from 18 to 24 months, revises setpoints, and
revises TS 2.2, ``Limiting Safety System Settings.'' Administrative
changes have also been made.
Date of issuance: December 2, 1997.
Effective date: December 2, 1997.
Amendment No.: 218.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Dates of initial notice in Federal Register: January 15, 1997 (62
FR 2194), March 12, 1997 (62 FR 11498) and June 4, 1997 (62 FR 30654).
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated December 2, 1997.
No significant hazards consideration comments received: No. The
supplemental information provided by the licensees did not affect the
proposed no significant hazards consideration.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Dated at Rockville, Maryland, this 10th day of December 1997.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 97-32763 Filed 12-16-97; 8:45 am]
BILLING CODE 7590-01-P