[Federal Register Volume 61, Number 60 (Wednesday, March 27, 1996)]
[Notices]
[Pages 13521-13540]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-7259]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 4, 1996, through March 15, 1996. The
last biweekly notice was published on March 13, 1996.
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 Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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 April 26, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing
[[Page 13522]]
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, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, 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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendments request: February 1, 1996
Description of amendments request: The proposed amendment would (1)
revise Technical Specifications (TS) Sections 3/4.1.1.1, 6.9.1.9, and
6.9.1.10 to relocate the shutdown margin (reactor trip breakers open)
to the Core Operating Limits Report (COLR); (2) revise TS 3/4.3.2
(Tables 3.3-3 and 3.3-4), to specify an additional restriction for the
allowed low pressurizer pressure trip setpoint when reducing reactor
coolant system (RCS) pressure in Mode 3; (3) revise TS Section 2.2.1
(Table 2.2-1) to make it consistent with the footnote in TS Tables 3.3-
3 and 3.3-4; and (4) revise TS Sections 3/4.5.2 and 3/4.5.3 to specify
an additional restriction to require that two emergency core cooling
system (ECCS) subsystems be operable in Mode 3 whenever the RCS cold
leg temperature is equal to or above 485 degrees F. In addition, the
Table of Contents and the Bases would be revised to be consistent with
these changes.
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 do not significantly increase the
probability or consequences of an accident previously evaluated in
the Updated Final Safety Analysis Report (UFSAR). The proposed
changes to TS Tables 2.2-1, 3.3-3, and 3.3-4 to add additional
restrictions to the pressurizer pressure - low trip setpoint
requirements are more conservative than the current Technical
Specifications and will reflect the updated Mode 3 steam line break
safety analyses assumptions. The proposed changes to TS sections 3/
4.5.2 and 3/4.5.3 to add additional restrictions to the requirement
to have two ECCS Subsystems operable are also more conservative than
the current Technical Specifications and will reflect the updated
Mode 3 steam line break safety analyses assumptions. Since these
changes are more restrictive, they would not contribute to the
initiation of any accident, nor would they increase the consequences
of an accident, but
[[Page 13523]]
they would enhance the plant response to a steam line break in Mode
3 to reduce consequences. The proposed changes to relocate the
shutdown margin - reactor trip breakers open to the COLR will have
no effect on the initiation or consequences of an accident. The
shutdown margin-reactor trip breakers open, which would be
determined using NRC approved analytical methods, as required by the
proposed changes, would ensure that the probability and consequences
of an accident would not increase. The changes to the titles of TS
3/4.5.2 and 3/4.5.3, and to the Table of Contents, are editorial and
have no effect on the operation of the plant or on any structures,
systems or components.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed TS changes do not create the possibility of an
accident of a new or different kind. The proposed changes to TS
Tables 2.2-1, 3.3-3, and 3.3-4, and TS section 3/4.5.2 and 3/4.5.3,
to add additional restrictions to the pressurizer pressure - low
trip setpoint requirement and add additional restrictions to the
requirement to have two ECCS Subsystems operable are more
conservative than the current Technical Specifications and will
reflect the updated Mode 3 steam line break safety analyses
assumptions. Since these changes are more restrictive, and therefore
bounded by the current TS, they would not contribute to the
initiation of any kind of new or different accident. The proposed
changes to relocate the shutdown margin -reactor trip breakers open
to the COLR will have no effect on the possibility of a new or
different kind of accident. The shutdown margin-reactor trip
breakers open, which would be determined using NRC approved
analytical methods as required by the proposed changes, would ensure
that there would be no possibility of a new or different kind of
accident from any accident previously evaluated. The changes to the
titles of TS 3/4.5.2 and 3/4.5.3, and to the Table of Contents, are
editorial and have no effect on the operation of the plant or on any
structures, systems or components.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed TS changes do not involve a reduction in any margin
of safety. The proposed changes to TS Tables 2.2-1, 3.3-3, and 3.3-
4, and TS section 3/4.5.2 and 3/4.5.3, to add additional
restrictions to the pressurizer pressure - low trip setpoint
requirement and add additional restrictions to the requirement to
have two ECCS Subsystems operable are more conservative than the
current Technical Specifications and will reflect the updated Mode 3
steam line break safety analyses assumptions. Since these changes
are more restrictive, they do not involve a reduction in any margin
of safety as currently established by the existing TS. The proposed
changes to relocate the shutdown margin - reactor trip breakers open
to the COLR will have no effect on any margin of safety. The
shutdown margin - reactor trip breakers open would be determined
using NRC approved analytical methods as required by the proposed
changes, thus ensuring that there would be no reduction in any
margin of safety. The changes to the titles of TS 3/4.5.2 and 3/
4.5.3, and to the Table of Contents, are editorial and have no
effect on the operation of the plant or on any structures, systems
or components.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: William H. Bateman
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: February 15, 1996
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.7 to add operability requirements
for the Keowee Hydro units during periods of commercial power
generation. These requirements are based on lake level and power level
of the Keowee Hydro units. Also, two surveillance requirements would be
added to TS 4.6 to (1) address periodic testing of the circuitry that
was added by the modification approved in NRC's SER dated August 15,
1995, and (2) add a load rejection surveillance to ensure that the
response of the Keowee Hydro units is bounded by the design criteria
used to develop the Keowee operating restrictions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) [Does not] involve a significant increase in the probability
or consequences of an accident previously evaluated:
Each accident analysis addressed within the Oconee Final Safety
Analysis Report (FSAR) has been examined with respect to the change
proposed within this amendment request. The probability of any
Design Basis Accident (DBA) is not significantly increased by this
change. In addition, the consequences of the accidents are within
the bounds of the FSAR analyses.
The design basis of the auxiliary electrical systems is to
supply the required engineered safeguards (ES) loads of one unit and
the safe shutdown loads of the other two units. The systems are
arranged so that no single failure will jeopardize plant safety. The
addition of the operability requirement and surveillances for the
Keowee Hydro units will ensure that the electrical systems can meet
their design basis.
(2) [Does not] create the possibility of a new or different kind
of accident from any kind of accident previously evaluated:
Addition of the operability requirement and surveillances will
not create a new or different kind of accident. The addition of the
circuitry which is covered by the operability requirement and
surveillances has been reviewed and approved by the NRC. Therefore,
operation of ONS [Oconee Nuclear Station] in accordance with this
Technical Specification amendment will not create any failure modes
not bounded by previously evaluated accidents. Consequently, this
change will not create the possibility of a new or different kind of
accident from any kind of accident previously evaluated.
(3) [Does not] involve a significant reduction in a margin of
safety:
The design basis of auxiliary electrical systems is to supply
the required ES loads of one Unit and safe shutdown loads of the
other two units. The ability of the Keowee Hydro units to provide
emergency power following an accident during a period of Keowee
Hydro commercial power generation was reviewed and approved by the
NRC in [an] SER dated August 15, 1995. Therefore, there will be no
significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: February 20, 1996
Description of amendment request: The proposed amendments would
revise Technical Specifications (TS) 3.1.5, 3.1.10, and 4.1. The TS
changes would: (1) reduce the frequency for the concentrated boric acid
storage tank boron concentration surveillance, (2) delete the chemical
and radiochemical surveillance requirements for the reactor
[[Page 13524]]
coolant for Sr189 and Sr190, gross beta
activity, gross alpha activity, dissolved gas concentration in the
reactor coolant, and gross beta activity in the steam generator
feedwater, and (3) relocate the surveillance requirements for tritium,
chloride, fluoride and oxygen to the Selected Licensee Commitments
(SLC) Manual. The proposed changes would also delete some temperature
and pressure requirements on control rod operation.
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. The licensee has determined that operation of the
facility in accordance with the proposed amendments would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated:
Each accident analysis addressed within the Oconee Final Safety
Analysis Report (FSAR) has been examined with respect to the
proposed amendment request. The probability of any Design Basis
Accident (DBA) is not significantly increased by the proposed
amendment due to the fact that the identified cause in the FSAR
accidents is not impacted. In addition, the consequences of the
accidents are within the bounds of the FSAR analyses since the
proposed amendment does not change the accident analysis methods or
assumptions described in the FSAR.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
The proposed amendment revises and eliminates several of the RCS
[Reactor Coolant System] chemistry Technical Specification
surveillance requirements. The changes in the surveillance
requirements do not alter the plant safety features or the method of
operation at ONS [Oconee Nuclear Station]. Therefore, operation of
ONS in accordance with the proposed Technical Specification will not
create any failure modes not bounded by previously evaluated
accidents.
(3) Involve a significant reduction in a margin of safety.
The proposed amendment does not impact the mitigation of any of
the accidents analyzed in the FSAR. Therefore, there is not a
significant reduction in the margin of safety associated with the
proposed amendment.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: February 22, 1996
Description of amendment request: The licensee has proposed to
increase the safety function lift setpoint tolerances for the safety
and relief valves that are listed in Surveillance Requirement 3.4.4.1
(Page 3.4-10) of the Technical Specifications TSs) for the Grand Gulf
Nuclear Station, Unit 1. The tolerances would be increased from the
current plus/minus 1 percent of the safety function (i.e., safety
relief valve) lift setpoint to plus/minus 3 percent.
The frequency of verifying these setpoints would not be changed by
this amendment request. Also, the other surveillance requirements in
the TSs on these valves and the number of these valves required to be
operable are not being changed by this amendment request.
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 (NSHC) in Attachment 2 to its application of February 22,
1996.
In its application, the licensee stated that it has used the NRC
staff's safety evaluation report (SER), NEDC 31753-P-A, issued in the
NRC letter of March 8, 1993, which evaluated General Electric (GE)
topical report NEDC-31753P, ``BWROG [BWR Owners' Group] In-Service
Pressure Relief Technical Specification Revision Licensing Topical
Report,'' dated February 1990.
The licensee's NSHC analysis is presented below:
Entergy Operations, Inc. is proposing that the Operating License
for Grand Gulf Nuclear Station (GGNS) be amended to increase the
tolerance of the safety function lift setpoints [from plus/minus 1%]
to plus/minus 3%. The GGNS Inservice Testing (IST) program controls
the frequency of safety relief valve (S/RV) testing as required by
the GGNS Operating License; therefore, this proposal will also
incorporate changes [concerning the setpoint tolerances] to
applicable IST procedures. GGNS will incorporate the recommendations
of the NEDC-31753-P-A [NRC staff's] SER, by resetting the safety
function [S/RV] lift setpoints for all tested valves to within plus/
minus 1% of the design lift setpoint and increasing the test sample
size by two valves for each valve found outside the plus/minus 3%
safety function lift setpoint. S/RV test sample population will be
determined based upon the currently licensed ASME [American Society
of Mechanical Engineers] Boiler and Pressure Vessel Code.
The commission has provided standards for determining whether a
no significant hazards consideration exists as stated in
10CFR50.92(c). A proposed amendment to an operating license involves
no significant hazards if the 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.
Entergy Operations, Inc. has evaluated the no significant
hazards considerations in its request for a license amendment. In
accordance with 10CFR50.91(a), Entergy Operations, Inc. is providing
the following analysis of the proposed amendment against the three
standards in 10CFR50.92(c):
a. No significant increase in the probability or consequences of
an accident previously evaluated results from this change.
The GGNS safety design bases for the S/RVs are:
) Prevent overpressurization of the nuclear system that could
lead to failure of the reactor coolant pressure boundary,
) Provide automatic depressurization for small breaks in the
nuclear system,
) Permit verification of operability,
) Withstand adverse combinations of loadings and forces during
abnormal, accident, or special event conditions.
The most limiting vessel overpressurization event is a closure
of all main steam isolation valves with a high flux scram. This
event was analyzed for GGNS using the minimum number of S/RVs
required by the GGNS Operating License. The safety function lift
setpoint tolerance used in the analysis bounds the proposed plus/
minus 3% setpoint tolerance. The analysis indicates that the S/RVs
are capable of maintaining adequate margin below the Operating
License Reactor Coolant System Pressure of 1325 psig.
Anticipated operational transients can also challenge the
operation of the S/RVs, for instance, Generator Load Reject without
Bypass. Analyses have been performed on the limiting events that
bound other pressure transient events using safety function limit
setpoint tolerances that bound the proposed plus/minus 3% tolerance
request. Fuel operating limits are based on the results of these
analyses; therefore, adequate fuel thermal margin is maintained.
Plant transients and events that require the use of automatic
depressurization and the low-low set feature utilize the relief mode
of S/RV operation. This proposed change does not affect the relief
mode of S/RV operation.
The verification of valve operability will still be performed in
accordance with the GGNS Inservice Testing Program, and S/RV safety
mode operability will be verified prior to reinstallation. Analysis
of the loads placed on each S/RV sub-system (discharge piping,
spargers and associated components) verifies that adequate margin
exists to ensure that the
[[Page 13525]]
overpressurization system can perform its designed function.
The negative tolerance of the safety function lift setpoint
remains above the highest setpoint of the S/RV relief mode, and
therefore normal vessel pressure. This margin provides reasonable
assurance that inadvertent opening of an S/RV will not occur during
power operations.
GGNS will replace each S/RV removed for IST program testing with
an S/RV that has been reset to within plus/minus 1% of the designed
safety function lift setpoint. During each refueling outage, at
least six of the installed S/RVs will be tested for safety lift
setpoint in accordance with the current IST program plant
procedures. This sample population is in agreement with the current
ASME Boiler and Pressure Vessel Code requirements for the GGNS IST
program, and is more restrictive than the ANSI/ASME OM-1-1981
requirement upon which the setpoint tolerance was based. For S/RV
setpoint testing ([the] as-found [setpoint]), additional valves will
be tested if the as-found setpoint is outside plus/minus 3% of its
designed safety function lift setpoint. Sample expansion will be
consistent with the NEDC 31753-P-A SER requirement of two additional
valves per valve failure.
The GGNS UFSAR currently requires at least fifty percent of the
installed valves to be removed and tested during each refueling
outage. GGNS FSAR Questions & Responses 211.49 discusses
the bases for this requirement. The concern regarded the performance
of S/RVs installed in operating plants at the time of GGNS
construction and licensing, and that new plants should have
significantly better performing S/RVs. The fifty percent requirement
provides a very conservative margin of testing to demonstrate that
no common cause of S/RV failure occurs within any one operating
cycle. The minimum testing of six valves proposed for each outage,
with additional testing for each failure from the initial test
population, provides reasonable assurance that no common cause
failure is occurring without early detection. [The minimum testing
of six valves is in agreement with the current ASME Code
requirements and is consistent with the current industry practices
that was accepted in the NRC staff's safety evaluation report, NEDC
31753-P-A.]
One of the major factors in the requirement of additional
testing population beyond ASME Boiler and Pressure Vessel Code is
many of the older plants were experiencing failures with multiple
stage pilot operated S/RVs. The safety function of this type of S/RV
requires operation of a pilot valve that is susceptible to excessive
leakage and corrosive bonding to cylinder walls; thereby preventing
proper safety function operation. The GGNS Dikkers S/RVs are direct
acting, and do not require the operation of a pilot valve for the
safety function. The Dikkers S/RV Instruction Manual recommends ``to
replace part of the installed valves each maintenance stop
(refueling outage)'', and does not prescribe any particular [number
of valves to be tested].
Therefore, no significant increase in the probability or
consequences of an accident previously evaluated results from this
proposed change.
b. This change would not create the possibility of a new or
different kind of accident from any previously analyzed.
The plant specific analyses verify that each S/RV will still
perform the intended function of preventing overpressurization of
the nuclear system. The vessel will have adequate margin below the
Operating License Reactor Coolant System Pressure of 1325 psig, and
plant system response will not deviate from the expected sequence of
events. Each system, structure, and component that communicates with
the reactor vessel has been verified to be within its design and
operational margin, and no unanticipated plant transients will occur
as a result of the safety lift function setpoint tolerance change.
The negative tolerance of the safety function lift setpoint
remains above the highest setpoint of the S/RV relief mode, and
therefore normal vessel pressure. This margin provides reasonable
assurance that inadvertent opening of an S/RV will not occur during
power operations.
This proposed change does not add any new systems, structures or
supports, nor does it introduce new S/RV operating modes.
Therefore, this change would not create the possibility of a new
or different kind of accident from any previously analyzed.
c. This change would not involve a significant reduction in the
margin of safety.
The increase in the S/RV safety function lift tolerance has been
analyzed for bounding limiting events and accident conditions. [The
safety function lift setpoint tolerance used in the analysis bounds
the proposed plus/minus 3% setpoint tolerance.] No condition exists
that reduces the margin of safety on the reactor coolant pressure
boundary or any system, structure or component that is required to
operate during vessel overpressurization events. Fuel operating
limits are based on the results of these analyses; therefore,
adequate fuel thermal margin is maintained.
[The negative tolerance of the safety function lift setpoint
remains above the highest setpoint of the S/RV relief mode, and
therefore normal vessel pressure. This margin provides reasonable
assurance that inadvertent opening of an S/RV will not occur during
power operations.]
Therefore, this change would not involve a significant reduction
in the margin of safety.
Based on the above evaluation, Entergy Operations, Inc. has
concluded that operation in accordance with the proposed amendment
involves no significant hazards considerations.
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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: February 22, 1996
Description of amendment request: The amendment proposes to delete
a specification which requires a thorough inspection of the Emergency
Diesel Generator (EDG) every 24 months during shutdown. In addition
this Technical Specification proposes to delete the phrase ``in any
thirty day period'' from a specification concerning Allowed Outage time
(AOT).
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 determined that this [technical specification
change request] TSCR poses no significant hazard as defined by the
NRC in 10 CFR 50.92.
1. State the basis for the determination that the proposed
activity will or will not increase the probability of occurrence of
the consequences of an accident.
The proposed activity deletes the requirement to inspect EDGs
during shut down from the Technical Specifications. It further
modifies the operability of a single EDG for a limited and defined
period of time. These changes do not affect the design or
performance of the EDGs or their ability to perform their design
function. Analysis using PRA techniques indicates the changes do not
significantly increase the probability or consequences of an
accident.
2. State the basis for the determination that the activity does
or does not create a possibility of an accident or malfunction of a
different type than any previously identified in the SAR.
The EDGs are not the source of any accident described in the
SAR. These changes do not modify the design or performance of the
EDGs and do not affect plant functions or actions. Therefore, the
proposed change does not create the possibility of an accident or
malfunction of a different type than those previously identified.
3. State the basis for the determination that the margin of
safety is not reduced. The proposed changes are designed to improve
EDG reliability and availability during shutdown periods by
providing flexibility in the scheduling and performance of
maintenance. The surveillance intervals are unchanged and
operability requirements are only modified to an acceptable degree.
The proposed activity does not alter the basis of
[[Page 13526]]
any technical specification that is related to the establishment or
maintenance of a nuclear safety margin. Therefore, the margin of
safety is not significantly reduced by this action.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: February 23, 1996
Description of amendment request: The proposed change to the
Technical Specifications would allow the implementation of 10 CFR 50,
Appendix J, Option B.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
GPU Nuclear has determined that this TSCR [technical
specification change request] involves no significant hazards
considerations as defined by NRC in 10 CFR 50.92.
The major changes from the existing Oyster Creek Technical
Specifications requested in accordance with the Option B
requirements:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident or
malfunction of equipment important to safety as previously evaluated
in the Safety Analysis Report.
The proposed change implements Option B of 10 CFR 50, Appendix J
on performance based containment leakage testing. The proposed
change does not involve a change to the plant design or operation.
Therefore, the proposed change does not affect any of the parameters
or conditions that contribute to initiation of any of the analyzed
accidents or malfunctions. The proposed change does request an
allowable extension of containment testing. Therefore, a
hypothetical leak could remain undetected for a greater period of
time. This slight increase in risk has been determined to be
insignificant as:
Type A Testing
NUREG 1493 determined that the effect of containment leakage on
overall accident risk is small as risk is dominated by accident
sequences that result in the failure or bypass of the containment.
Industry wide PCILRTs have demonstrated that only a small fraction
of the leaks discovered during testing exceeded acceptance criteria,
and that the leak rate has been only marginally above the acceptable
limit. Only 3% of all leaks can be detected only by PCILRT,
therefore, only 3% of the theoretical leaks are affected by the
extension to the Type A test interval. Experience at Oyster Creek
agrees with the industry wide data in that the majority of the
detected leakage from the primary containment is found through Type
B and C testing.
NUREG 1493 found that these observations, together with the
insensitivity of reactor accident risk to the containment leakage
rate, demonstrates that increasing the Type A leakage test intervals
would have a minimal impact on public risk.
Type B and C Testing
Penetrations are designed to ensure reliability of the
containment isolation function. Type B penetrations use a double
passive seal (e.g. o-ring, gasket) and Type C penetrations use a
double isolation valve design to ensure reliability of the isolation
function. Because valves perform the isolation function actively,
they are more likely to fail on demand (e.g. failure to completely
close on demand). To address this failure mode, Type C valves are
subjected to increased design constraints and testing to ensure both
acceptable leak rates and stroke times. The proposed change does not
alter the installation, operation, operating environment, or testing
method of these valves. Therefore, the proposed change does not
introduce any new component failure modes, nor does it affect the
probability of occurrence of any existing evaluated failure mode.
The failure of any single penetration barrier (isolation valve
or passive seal) does not cause penetration failure. Therefore, a
double failure would have to occur to cause a failure of the
penetration and affect containment. Additionally, the proposed
change does not change the acceptance criteria for acceptable
leakage testing.
The proposed change does not alter plant design or operation,
nor does it alter the allowable maximum leakage rate limit. Thus,
the proposed change does not affect the probability of occurrence
nor the consequences of any evaluated accident or malfunction of
equipment important to safety.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of an accident or
malfunction different from any accident or malfunction previously
evaluated.
The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
initiation of any accidents. This change only involves the reduction
in Type A, B and C test frequencies, and the Type A test pressure.
Type A Testing
The only changes proposed to the Type A testing are to frequency
and test pressure. As the proposed test pressure is grater than the
existing test pressure, no new type of accident or malfunction is
created, and the increase in pressure provides an additional margin
of safety. The increase in pressure provides an additional margin of
safety. The increase in surveillance interval cannot introduce any
new type of accident or malfunction.
The PCILRT is presently performed at 20 psig. Performance of the
PCILRT at Pa (35 PSIG) will provide a more direct leak rate for
analysis.Pa is the design pressure of the torus (the drywell
design pressure is 44 psig, but the torus is non isolable form the
drywell. Therefore, Pa will not create the possibility of the
failure of the torus due to overpressurization. No new accident
modes can be created by extending the test intervals. No safety
related functions or components are altered as a result of this
change. Therefore, no new accident or malfunction different form
those evaluated in the Safety Analysis Report can result due to the
increase in test pressure or increase in surveillance interval.
Type B and C Testing
The proposed change only deals with the frequency of performing
Type B and C testing. It does not change what components are tested
or the method of testing. There is no proposed change to the design
or operation of the plant. Therefore, no new accident or malfunction
different form those evaluated in the Safety Analysis Report can
result due to the increase in test pressure or increase in
surveillance interval.
3. Operation of the facility in accordance with the proposed
amendment would not decrease the margin of safety as defined in the
bases of the Technical Specifications.
Type A Testing
Except for the method of defining the test frequency and
pressure at which the PCILRT is performed, the methods for
performing the actual test are not changed. However, the proposed
change can increase the probability that an increase in leakage
could go undetected for an extended period of time. NUREG 1493 has
determined that under several different accident scenarios, the
increased risk of radioactivity release from containment is
negligible with the implementation of these proposed changes.
Type B and C Testing
The proposed change only affects the frequency of Type B and C
testing. The methods for performing the actual test are not changed.
The design or operation of Type B and C components are not changed.
The proposed change will result in a longer interval between tests
of good performing Type B and C components.
The margin of safety that has the potential of being impacted by
the proposed change involves the offsite dose consequences of
postulated accidents which are directly related to containment
leakage rate. The containment isolation system is designed to limit
leakage to La, which is defined by the Oyster Creek Technical
Specifications to be 1.0 percent by weight of the containment air at
35 psig per 24 hours. The limitation on
[[Page 13527]]
containment leakage rate is designed to ensure the total leakage
volume will not exceed the value assumed in the accident analyses at
the peak accident pressure (Pa). The margin of safety for the
offsite dose consequences of postulated accidents directly related
to the containment leakage rate is maintained by meeting the 1.0
La acceptance criteria. The La value is not being modified
by this proposed Technical Specification change request.
Therefore, the margin of safety as defined in the bases for the
Technical Specification will not be reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: February 22, 1996 (AEP:NRC:0659AA)
Description of amendment requests: The proposed amendments would
revise the technical specifications to remove the requirement that the
Operations Superintendent must hold or have held a Senior Operator
License at Cook Nuclear Plant, or a similar reactor. In addition, a
mid-level operations manager will only be required to hold a Senior
Operator License if the Operations Superintendent does not hold one.
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:
Per 10 CFR 50.92, this proposed change does not involve a
significant hazards consideration because the change does not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated,
2. create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. involve a significant reduction in a margin of safety.
Criterion 1
The amendment request does not involve a significant increase in
the probability or consequences of [an] accident previously
evaluated because the proposed change to the Technical Specification
does not affect the assumptions, parameters, or results of any UFSAR
[updated final safety analysis report] accident analysis. The
proposed amendment does not modify any existing equipment. It is
concluded that the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2
The proposed change does not involve physical changes to the
plant or changes in plant operating configuration. The proposed
change updates the requirements for the Operations Superintendent.
Thus, it is concluded that the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3
The proposed change updates the requirements for Operations
Superintendent. There is no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of amendment request: March 12, 1996 (AEP:NRC:1248)
Description of amendment request: The proposed amendment would
remove the technical specifications related to shutdown and control rod
position indication while in modes 3, 4, and 5. The change would make
the Unit 2 technical specifications consistent with the Unit 1
technical specifications and the Standard Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, this proposed change does not involve a
significant hazards consideration because the change does not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated,
2. create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. involve a significant reduction in a margin of safety.
Criterion 1
The boron concentration in the reactor coolant system will be
high enough to assure adequate SDM in modes 3, 4, and 5. The
calculation to obtain the required boron concentration takes into
account the position of the rods. Shutdown margin is assumed as an
initial condition in the safety analysis. The safety analysis
establishes a SDM that ensures specified acceptable fuel design
limits are not exceeded. As long as the SDM is satisfied, no change
in the probability or consequences of an accident previously
evaluated will result from the proposed deletion of the ``position
indicator - shutdown'' specification. It is noted that this change
is consistent with the new ISTS approved by the NRC as NUREG-1431,
Rev. 1.
Criterion 2
The ability to insert the control and shutdown rods provided by
the rod control system is not affected by the OPERABILITY status of
the ARPI system. As mentioned previously, the reactor coolant system
boron concentration will be high enough to assure adequate SDM is
maintained. Therefore, it is concluded that the proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Criterion 3
The margin of safety requirements are not affected by the
removal of this T/S. The required SDM which is an initial condition
in the safety analysis, is unaffected since the reactor coolant
system boron concentration is increased to address the potential
``all rods out'' configuration. Based on these considerations, it is
concluded that 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: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: November 29, 1995
Description of amendment request: The proposed amendment would
[[Page 13528]]
modify the Technical Specifications to remove the requirement for
additional pressure relief by a residual heat removal (RHR) spring
relief valve during low temperature overpressure protection (LTOP)
conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change to delete Technical Specification 3.4.D.3b
has been evaluated against the standards of 10 CFR 50.92 and has
been determined not to involve a significant hazards consideration.
This proposed change does not:
1. Involve a significant increase in the probability or
consequence of an accident previously analyzed. The Power Operative
Relief Valves (PORVs) remain operable to mitigate any LTOP event.
Thus, this change does not result in an increase in the probability
or consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously evaluated. Removing the RHR spring relief valve as
an additional relief requirement does not create the possibility of a
new or different kind of accident since the proposal involves neither a
hardware modification nor the creation of a unique operating condition.
3. Involve a significant reduction in a margin of safety.
Removing the RHR spring relief valve as an additional requirement
does not change the results of any of the FSAR Chapter 14 events.
The PORVs remain operable to maintain the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director:
John Zwolinski
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: November 29, 1995
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3.14 to decrease the maximum steam
generator (SG) primary-to-secondary leakage rate from 0.15 gpm to 0.10
gpm and would modify TS 4.10 by revising the requirements for
unscheduled SG tube inspections that are performed on each SG following
a primary-to-secondary tube leak.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
A steam generator leakage assumption greater than the proposed 0.10
gpm/SG limit has been used in the FSAR [Final Safety Analysis
Report] Chapter 14 safety analyses. Thus, the FSAR Chapter 14 safety
analyses remain bounding. Assuring that an adequate leakage limit
exists that initiates corrective actions in a timely manner is
important to ensuring a steam generator tube rupture event does not
take place. This change modifies the steam generator post-leakage
testing requirements to focus inspections on leaking tubes and areas
likely to produce similar leakage, in lieu of an expanded test
campaign of all three steam generators. Without this change,
Technical Specifications require inspection of 3% of the tubes in
each steam generator. By inspecting the critical areas of the
affected steam generator and possibly expanding inspections to the
critical areas of the remaining steam generators, the probability
and/or consequences of previously evaluated accidents (e.g., steam
generator tube rupture) 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 proposed changes will not involve a modification to
existing hardware at the plant. The decrease in the maximum
allowable steam generator primary leakage rate tends to provide
additional time for operator action to take place which, if timely
enough, would avoid the consequences of a tube rupture event. The
proposed inspection campaign requires inspection of the critical
area and may be expanded to the other steam generators to ensure
that additional tubes will not fail due to similar causes. This
modified inspection campaign does not introduce the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety. The FSAR Chapter 14 safety analyses assume a
higher steam generator leakage rate and therefore remain
conservative. The proposed reduction in the allowable leakage
provides a greater margin of safety since it is more conservative
than the present value. This change modifies inspection requirements
of Technical Specifications and does not impact the plant design or
equipment. The modified inspection requirements following a plant
shutdown due to tube leakage concentrate steam generator tube
inspections in those areas believed to be most susceptible to flaws.
For these reasons, we believe the proposed changes increase the
margin of safety by inspecting the critical areas of the steam
generator(s) in lieu of additional random inspections.
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: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director:
John Zwolinski
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of amendment request: November 8, 1995
Description of amendment request: The amendment request would
revise the Technical Specifications (TS) for the jet pumps to be
consistent with the limiting conditions for operation and surveillance
requirements in the Standard Technical Specifications for General
Electric Plants (NUREG-1433).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
...The proposed change does not involve an [significant hazards
consideration] SHC because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The new LCO [Limiting Condition for Operation] does not diminish
the existing requirement that all jet pumps must be operable, nor
does it affect the time available to achieve cold shutdown should a
pump become inoperable. The new LCO does eliminate the ability to
continue to operate with the indication (but not the function) of a
single jet pump inoperable. This does not increase the possibility
of an unnecessary plant shutdown due to inoperable instrumentation
since sufficient flexibility exists in the surveillance requirement
so that operability of the jet pumps can be verified. This change
eliminates the LCO that allowed continued operation with conditions
that could potentially mask an inoperable pump. The new LCO is more
limiting in ensuring that the plant is operated in a condition for
which accidents were analyzed.
The new surveillance requirement provides a more accurate method
of ensuring
[[Page 13529]]
the jet pumps remain operable. The new surveillance criteria are
more sensitive to jet pump failures and the degradation of the jet
pumps prior to failure.
Based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The new LCO and surveillance does not change the manner in which
the plant is operated, nor does it reduce the operability
requirements of any jet pump, Therefore, no new or different kind of
accident can be created by the new specification. The surveillances
that will be performed do not require any new hardware or plant
evolutions. Therefore, the proposed change to the LCO and
surveillance cannot create the possibility of a new or different
kind of accident.
3. Involve a significant reduction in the margin of safety.
The margin of safety that currently exists is not diminished by
this change. The requirement to place the reactor in cold shutdown
within 24 hours should a jet pump become inoperable is maintained.
The LCO which allowed continued operation with indication for one
pump inoperable has been eliminated.
The new surveillance requirement continues to demonstrate the
operability of the jet pumps and during operation, continues to be
performed at the same interval as in the current technical
specifications. The note (which allows the surveillance to be
deferred until four hours after the associated recirculation loop is
in operation and 24 hours after exceeding 25% of rated thermal
power) does not significantly affect the margin of safety. The time
that the unit would be operating in these conditions would be small,
and the stress placed on the pump at less than 25% power is lower.
Based on the above, this change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London,
Connecticut
Date of amendment request: November 3, 1995
Description of amendment request: The proposed amendment will
extend the allowed outage time from 48 hours to 7 days for an emergency
core cooling system train that is declared inoperable as a result of an
inoperable low pressure safety injection subsystem.
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:
Pursuant to 10CFR50.92, Northeast Nuclear Energy Company (NNECO)
has reviewed the proposed change to extend the allowed outage time
(AOT) for an inoperable low pressure safety injection (LPSI)
subsystem from the existing limit of 48 hours to 7 days. In
addition, the change to modify the completion time for the Action
Statement and the criteria for the Surveillance Requirements were
also reviewed. NNECO concludes that these changes do not involve a
significant hazards consideration (SHC) since the proposed change
satisfies the criteria in 10CFR50.92(c). That is, the proposed
change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed amendments for Millstone Unit No. 2 will extend the
action completion AOT for a single inoperable LPSI train from 48
hours to 7 days. A LPSI subsystem is designed as a part of each
emergency core cooling system (ECCS) train to supplement safety
injection tank inventory during the early stages of mitigating a
design basis accident (DBA). As such, components of the LPSI
subsystem are not accident initiators, and an extended AOT to
restore operability of an inoperable LPSI subsystem would not
increase the probability of occurrence of accidents previously
analyzed.
The safety analyses for Millstone Unit No. 2 demonstrates that
ECCS performance acceptance criteria are satisfied with only one of
the two redundant ECCS trains operating during the postulated DBA.
The proposed technical specification revisions involve the AOT for a
single inoperable LPSI subsystem, and do not change the conditions
assumed for the minimum amount of operating equipment needed for
accident mitigation. Therefore, the consequences of an accident
previously evaluated will not be significantly increased.
In addition, CE NPSD-995 recognizes that when an ECCS train is
inoperable due to a LPSI subsystem being unavailable, due either to
being declared inoperable (by failing a surveillance requirement) or
is intentionally taken out-of-service (for corrective or preventive
maintenance), the core damage frequency (CDF) during power operation
increases. The results of the PRA presented in CE NPSD-995 show that
the proposed increase in the ECCS AOT (due to LPSI unavailability)
from 48 hours to 7 days does not cause a significant increase in the
overall CDF of Millstone Unit No. 2.
The analyses indicate that continued plant operation with a
single LPSI subsystem out-of-service may result in a small increase
in ``at power risk;'' however, that risk increase will be negligibly
small and controlled effectively via the Maintenance Rule and the
risk monitor program that minimizes the outage time and prevents
entering into an unacceptable risk configuration. In addition, the
proposed AOT extension for the LPSI subsystem is evaluated as having
negligible impact on the large early radiological release
probability for Combustion Engineering pressurized water reactors in
the event of a design basis accident.
Therefore, operation in accordance with the proposed amendment
would not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed amendment will not change the physical plant or the
modes of plant operation defined in the technical specifications.
The changes do not involve the addition or modification of equipment
nor do they alter the design of plant systems. Therefore, operation
of Millstone Unit No. 2 in accordance with its proposed amendment
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in the margin of safety.
The margin of safety associated with the ECCS train is
established by acceptance criteria for system performance defined in
10CFR50.46. The proposed amendment will not change this acceptance
criteria nor the operability requirements for equipment that is used
to achieve such performance as demonstrated in the Millstone Unit
No. 2 safety analyses. Moreover, an integrated assessment of the
risk impact of extending the AOT for a single inoperable LPSI train
has concluded that the risk contribution is small. Therefore,
operation of Millstone Unit No. 2 in accordance with its proposed
amendment would not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
[[Page 13530]]
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: September 12, 1995
Description of amendment request: The amendment would revise and
reformat Technical Specification (TS) 6.3.1 to add the requirement that
the Assistant Operations Manager shall hold a senior reactor operator
(SRO) license if the Operations Manager does not hold an SRO license
for Millstone Unit 3. Also the footnote would be deleted from TS 6.3.1
that previously granted a one-time three year exception to the
qualification requirements for the Operations Manager and an exception
for the Assistant Operations Manager to hold a license instead of the
Operations Manager.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
...The proposed change does not involve an [significant hazards
consideration] SHC because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed change affects an administrative control, which was
based on the guidance of ANSI N18.1-1971. ANSI N18.1-1971
recommended that the Operations Manager hold an SRO license. The
current guidance in Section 4.2.2 of ANSI/ANS 3.1-1987 recommends,
as one option, that the Operations Manager have held a license for a
similar unit and the Operations Middle Manager hold an SRO license.
While the Operations Middle Manager position does not exist at
Millstone Unit No. 3, [Northeast Nuclear Energy Company] NNECO has
created the position of Assistant Operations Manager. The individual
in this position would meet the requirements for, and would have
responsibilities as recommended in, ANSI/ANS 3.1-1987 for the
Operations Middle Manager position.
Therefore, the proposed change requests an exception to ANSI
N18.1-1971 to allow use of ANSI/ANS 3.1-1987 in a limited
circumstance. Specifically, the proposed revision to Technical
Specification 6.3.1 would require the Operations Manager to either
hold an SRO license at Millstone Unit No. 3 or have held an SRO at a
[pressurized water reactor] PWR.
If the Operations Manager does not hold an SRO license at
Millstone Unit No. 3, the specification will require the Assistant
Operations Manager to hold, and continue to hold, an SRO license.
The proposed change includes the requirement for the Operations
Manager to have held a license for a similar unit (a PWR) in
accordance with Section 4.2.2 of ANSI/ANS 3.1-1987. For those areas
of knowledge that require an SRO license, the Assistant Operations
Manager will provide the technical guidance normally provided by the
Operations Manager.
The proposed change does not alter the design of any system,
structure, or component, nor does it change the way plant systems
are operated. It does not reduce the knowledge, qualifications, or
skills of licensed operators, and does not affect the way the
Operations Department is managed by the Operations Manager. The
Operations Manager will continue to maintain the effective
performance of his personnel and ensure the plant is operated safely
and in accordance with the requirements of the operating license.
Additionally, the Control Room Operators will continue to be
supervised by the licensed Shift Supervisors.
The proposed change does not detract from the Operations
Manager's ability to perform his primary responsibilities. In this
case, by having previously held an SRO license, the Operations
Manager has achieved the necessary training, skills, and experience
to fully understand the operation of plant equipment and the watch
requirements for operators. In summary, the proposed change does not
affect the ability of the Operations Manager to provide the plant
oversight required of his position. Thus, it does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change to Technical Specification 6.3.1 does not
affect the design or function of any plant system, structure, or
component, nor does it change the way plant systems are operated. It
does not affect the performance of NRC licensed operators. Operation
of the plant in conformance with technical specifications and other
license requirements will continue to be supervised by personnel who
hold an NRC SRO license. The proposed change to Technical
Specification 6.3.1 ensures that the Operations Manager will be a
knowledgeable and qualified individual to have held an SRO license
at a PWR. Based on the above, the proposed change does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed change involves an administrative control that is
not related to the margin of safety. The proposed change does not
reduce the level of knowledge or experience required of an
individual who fills the Operations Manager position, nor does it
affect the conservative manner in which the plant is operated. The
Control Room Operators will continue to be supervised by personnel
who hold an SRO license. Thus, 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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: November 21, 1995
Description of amendment request: The licensee proposes to change
Technical Specification Section 1.33 and Bases Sections 3/4.3.3.9 and
3/4.3.3.10, and 3/4.11.2.1. The changes clarify the definition of
source check to include a source check from a light emitting diode
(LED), as well as from ionizing radiation.
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 concludes that these changes do not involve a
significant hazards consideration since the proposed changes satisfy
the criteria in 10CFR50.92(c). That is, the proposed changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed changes to the definition of source check clarifies
the source check for the liquid and gaseous effluent radiation
monitors. These monitors do not provide a safety function and only
serve to provide radiological information to plant operators,
therefore, the changes will not increase the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes to the definition of source check have no
effect on the ability of the monitors to perform their designed
function. The clarification to the surveillance do not involve any
physical modifications to any equipment, structures, or components.
The monitors already have the internal LEDs which were originally
used to perform the source check. The proposed changes have no
impact on design basis accidents, and the changes will not modify
plant response or create a new or unanalyzed event.
3. Involve a significant reduction in the margin of safety.
[[Page 13531]]
The proposed changes to the definition of source check do not
have any impact on the protective boundaries and, therefore, have no
impact on the safety limits for these boundaries. The
instrumentation associated with these changes do not provide a
safety function and only serve to provide radiological information
to plant operators. The instrumentation has no affect on the
operation of any safety-related equipment. As such, these changes
have no impact on the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: February 15, 1996
Description of amendment request: The amendment changes the
Technical Specifications to implement 10 CFR Part 50, Appendix J,
Option B, by creating Technical Specification Section 5.5.12, ``Primary
Containment Leakage Rate Testing Program,'' which refers to Regulatory
Guide 1.163, ``Performance-Based Containment Leakage-Test Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) The proposed changes do not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
The adoption of 10 CFR 50, Appendix J Option B will not involve
a significant increase in the probability or consequences of any
accident previously evaluated. The proposed changes to the TS
[Technical Specifications] reflect the use of the performance-based
containment leakage-testing program. The USNRC has approved the use
of a performance-based option for containment leakage testing
programs when it amended 10 CFR 50, Appendix J (60 FR 49495). For
adoption of the revised regulation, licensees are required to
incorporate into their TS, by general reference, the USNRC
regulatory guide or other plant-specific implementing document used
to develop their performance-based leakage testing program. A new
Administrative Control subsection (5.5.12, ``Primary Containment
Leakage Rate Testing Program'') has been added that requires the
establishment and maintenance of a Primary Containment Leakage Rate
Testing Program. The TS will still require the performance of a
periodic general visual inspection of the containment to ensure
early detection of any structural deterioration of the containment
that may occur.
As concluded in NUREG-1493, given the insensitivity of risk to
containment leakage rate and the small fraction of leakage paths
detected solely by ILRT [Integrated Leak Rate Test] testing,
increasing the interval between ILRTs is possible with minimal
impact on public risk. Additionally, performance-based alternatives
to current LLRT [Local Leak Rate Test] requirements are feasible
without significant risk impacts. Additionally, these changes will
not alter any safety limits which ensure the integrity of fuel
barriers, and will not result in a significant increase to onsite or
offsite dose.
No physical changes are being made to the plant, nor are there
any changes being made in the operation of the plant as a result of
these changes which could involve a significant increase in the
probability or consequences of any accident previously evaluated.
Additionally, these changes will not alter the operation of
equipment assumed to be available for the mitigation of accidents or
transients.
2) The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The adoption of 10 CFR 50, Appendix J Option B will not create
the possibility of a new or different type of accident from any
previously evaluated. These changes to the PBAPS, Units 2 and 3 TS
will not involve any changes to plant systems, structures or
components (SCCs) which could act as new accident initiators. These
changes will not impact the manner in which SSCs are tested such
that a new or different type of accident from any previously
evaluated could be created.
3) The proposed changes do not result in a significant reduction
in the margin of safety.
No margins of safety are reduced as a result of the proposed
adoption of 10 CFR 50, Appendix J Option B. As stated previously,
the USNRC has approved the use of this performance-based option for
containment leakage testing programs when it amended 10 CFR 50,
Appendix J (60 FR 49495). These changes will not impact core limits
or any other parameters that are used in the mitigation of a UFSAR
[Updated Final Safety Analysis Report] design-basis accident or
transient. Additionally, these changes do not introduce any hardware
changes, and will not alter the intended operation of plant
structures, systems or components utilized in the mitigation of
UFSAR design-basis accidents or transients. These changes will not
introduce any new failure modes of plant equipment not previously
evaluated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101
NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket No. 50-387,
Susquehanna Steam Electric Station, Unit 1, Luzerne County,
Pennsylvania
Date of amendment request: January 26, 1996
Description of amendment request: The proposed amendment removes
three pressure relief valves from Technical Specification Table 3.6.3-
1, ``Primary Containment Isolation Valves,'' since these valves are no
longer needed to support the steam condensing mode of the residual heat
removal (RHR) system and are being removed from the plant.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
With the prior deletion of the steam condensing mode of RHR and
the isolation of the high and low pressure interfaces, the three
pressure relief valves that are being removed from the plant have no
active function. Their passive function of maintaining system or
containment integrity will be fulfilled by blind flanges. Also, the
RHR and RCIC [reactor core isolation cooling] piping are provided
with overpressure protection from other pressure relief valves.
Therefore, the removal of these pressure relief valves does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The pressure relief valves that are being removed had two
primary functions. First,
[[Page 13532]]
they provided overpressure protection for the RHR and RCIC piping
during the steam condensing mode of RHR. Since the steam condensing
mode has been deleted from the plant, these valves no longer have
that function. Also, overpressure protection of the RHR and RCIC
piping is provided by other existing pressure relief valves. Second,
these valves maintained system or containment integrity. When the
pressure relief valves are removed from the plant, they will be
replaced with blind flanges or equivalent that will maintain system
or containment integrity. Therefore, the removal of the three
pressure relief valves 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.
Since the steam condensing mode of RHR has been eliminated, the
three pressure relief valves have no active function. Their passive
function of maintaining system or containment integrity will be
fulfilled by blind flanges or equivalent. Also, overpressure
protection of RHR and RCIC piping is provided by other existing
pressure relief valves. Therefore, the removal of the three pressure
relief valves does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: January 25, 1996
Description of amendment request: The amendment proposes to revise
the allowed out-of-service times for single inoperable Emergency Diesel
Generators (EDGs) to accommodate on-line maintenance of the EDGs. In
addition, two line item changes are proposed: (1) to improve safety by
reducing EDG testing at power; and (2) to revise the ac power
requirements during cold shutdown or refueling modes to make the James
A. FitzPatrick (JAF) Technical Specifications consistent with the
Standard Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
a. EMERGENCY DIESEL GENERATOR LCO [Limiting Conditions for
Operation] AT POWER
The proposed changes to the Technical Specifications will allow
longer Allowed Out of Service Times [AOTs] to perform necessary
repair and maintenance on individual Emergency Diesel Generators
while at power. This extended AOT will enhance scheduling of
preventive maintenance of individual EDGs without significantly
increasing the probability or consequences of an accident previously
evaluated. The risk evaluations contained in the JAF quantitative
analyses of the EDGs determined that the probability of an accident
by increasing the AOT for an individual EDG from 7 days to 14 days
is non-risk-significant. The primary reason for this low relative
risk is due to the designed redundancy and capability to respond to
an accident when a single diesel generator is out of service. LOCA
[loss-of-coolant accident] Analyses that assume the worst case line
break while an EDG is out of service indicate the plant can be
safely shut down with the remaining EDGs. Even if another EDG should
fail during the AOT, at least one Core Spray and one Residual Heat
Removal (RHR) Low Pressure Coolant Injection pump can provide the
required flow to bring the plant to safe shut down. Furthermore,
long term suppression pool and reactor shutdown cooling is provided
by any one of the three remaining RHR pumps for a single EDG out of
service or by two remaining RHR pumps assuming an additional EDG
failure during the AOT.
Increasing the EDG AOT does not involve physical alteration of
any plant equipment and does not affect analysis assumptions
regarding functioning of required equipment designed to mitigate the
consequences of accidents. Further, the severity of postulated
accidents and resulting radiological effluent releases will not be
affected by the increased AOT for a single EDG.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN
Changing the number of EDGs required during plant shutdown does
not involve physical alteration of any plant equipment and does not
affect analysis assumptions regarding functioning of required
equipment designed to mitigate the consequences of accidents.
Further, the severity of postulated accidents and resulting
radiological effluent releases will not be affected by the change in
the LCO during shutdown.
c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION
The proposed change to the Technical Specification will reduce
the required number of tests to be performed when an EDG or EDG
System is inoperable. This proposed change to TS requirements
addresses the concern of excessive testing that could result in EDG
wear which is counter-productive to safety in terms of equipment
degradation and availability. This change is consistent with Generic
Letter 93-05 guidance for implementing such recommendations. The
proposed Technical Specifications will not result in a change to the
design or operation of the facility, therefore, this change will not
result in a significant increase in the probability or consequences
of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
a. EMERGENCY DIESEL GENERATOR LCO AT POWER
Extending the AOT for an individual EDG does not necessitate
physical alteration of the plant or changes in parameters governing
normal plant operation. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated for JAF plant.
b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN
Changing the number of EDGs required during shutdown does not
necessitate physical alteration of the plant or changes in
parameters governing normal plant operation. Thus, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated for JAF plant.
c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION
The proposed change does not change design, operation or the
testing process. The nature of this change precludes the possibility
of a new or different kind of accident. The proposed change to
complete the required action does not involve any hardware changes,
nor changes to the operation of the equipment nor does it change the
ability of the equipment to perform its intended function.
Performing the testing on an extended time cannot initiate any type
of accident.
3. Involve a significant reduction in the margin of safety.
a. EMERGENCY DIESEL GENERATOR LCO AT POWER
As discussed above, the JAF quantitative evaluation determined
that the change in risk associated with extending the AOT for a
single EDG is non-risk-significant. In addition, the design provides
adequate redundancy for safe shut down during the AOT for a single
EDG out of service. This is supported by the LOCA analyses including
analyses for long term suppression pool and reactor shutdown
cooling.
b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN
The margin of safety is not affected by changing the number of
EDGs required during shutdown. One offsite power source or one EDG
ensure the availability of the
[[Page 13533]]
required power to recover from postulated accident events during
shutdown. When the required number of operable systems is not met,
all work that could potentially initiate a postulated accident event
during shutdown is suspended.
c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION
The proposed change to Technical Specifications reduces testing
at reactor power. The overall effect is a net gain in plant safety
by avoiding the potential for unnecessary wear that could degrade
the EDGs at power. Implementation of these changes is consistent
with the guidance provided by the NRC in Generic Letter 93-05. The
proposed change to the EDG testing requirements does not reduce the
ability of the equipment to perform its intended safety function.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: January 30, 1996
Description of amendment request: The proposed Technical
Specifications change will delete the requirement that oxygen
concentrations for both normal and transient conditions not exceed
saturation when the reactor coolant is below 250 degrees F. The
Technical Specifications change will also eliminate the surveillance
requirement for reactor coolant chemistry sampling of chloride,
fluoride, and oxygen concentration during maintenance activities when
fuel is removed from the reactor vessel and the Reactor Coolant System
(RCS) is drained below the reactor vessel flange regardless of whether
the upper internal and/or vessel heat are in place or not.
Administrative result of the changes being made, capitalize Technical
Specifications defined terms to maintain consistency within the
Technical Specifications, and the word ``degrees'' is spelled-out when
referring to the Fahrenheit temperature, rather than using the symbol.
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 Surry Power Station in accordance
with the proposed changes will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Since the RCS and the RHR [Residual Heat Removal] System are
drained when the RCS inventory is reduced below the reactor vessel
flange for maintenance or refueling activities, the concentrations
of chlorides and fluorides will not change. During these maintenance
or refueling activities, only controlled makeup to the RCS is
planned, and any planned or unplanned makeup to the RCS would be
detected by available level indication. Sampling for chloride and
fluoride concentrations in the RCS will be performed prior to
draining the system. Sampling of the reactor coolant for chloride
and fluoride concentrations will resume when the RCS is filled. The
chloride and fluoride concentrations will be known and will be
maintained consistent with the Technical Specification Limiting
Condition for Operation and Action Statements. Also, when the RCS
inventory is drained below the reactor vessel flange, the RCS is
vented and open to the containment building atmosphere with the
reactor coolant liquid considered oxygen saturated. Technical
Specification 3.1.F.4 allows normal and off-normal ``saturated''
oxygen concentrations when reactor coolant temperature is below 250
degrees F. Consequently, sampling the reactor coolant for oxygen
concentration under these conditions is not required and the
Technical Specification Table 4.1-2B specified sampling frequency of
five (5) times per week is not necessary since the oxygen
concentration continues to remain in compliance with the Technical
Specification limit, measures are available and action can be taken
to correct the condition prior to any deleterious effect.
Surry Technical Specifications 3.1.F.1 prohibits reactor coolant
temperature from exceeding 250 degrees F unless chloride, fluoride,
and oxygen concentrations are within specified limits. Therefore a
significant increase in the probability or consequences of an
accident previously evaluated does not exist.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The materials that are exposed to reactor coolant are corrosion
resistant. They were chosen for specific applications within the
system and for their compatibility with the reactor coolant. The
chemical composition of the reactor coolant will be maintained
within the specifications given within Technical Specification
3.1.F, Updated Final Safety Analysis Report Table 4.2-2, and
Technical Specification Table 4.1-2B. Because of the time dependent
nature of any adverse affects from chloride, fluoride, and oxygen
concentrations in excess of the Technical Specifications limits,
measures are available and can be taken to correct the condition
while the reactor is in a safe shutdown condition, prior to any
deleterious effect. No hardware modifications are involved. System
configuration and plant operations are not being changed. Therefore,
the possibility of a new or different kind of accident from any
accident previously evaluated has not been created.
(3) Involve a significant reduction in the margin of safety.
This change does not involve a significant reduction in the
margin of safety since the chloride and fluoride concentrations are
maintained within their specified values prior to RCS drain down and
following refill. The time period during which the RCS inventory is
reduced below the reactor vessel flange and fuel is removed from the
vessel, is short and insignificant in terms of the parameters
necessary to initiate a corrosion concern. Existing Technical
Specifications Action Statements and Allowed Technical Specification
values for normal and off-normal concentrations of chlorides and
fluorides are not being changed. No hardware modifications are
involved. System configuration and plant operations are not being
changed. Surry Technical Specification 3.1.F.1 remains unaffected by
this change and continues to prohibit reactor coolant temperature
from exceeding 250 degrees F unless chloride, fluoride, and oxygen
concentrations are within specified limits.
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: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: Eugene V. Imbro
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
[[Page 13534]]
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.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County,
Texas
Date of amendment request: February 29, 1996
Description of amendment request: The proposed amendment would
include the addition of Technical Specification 3.10.8 which would
allow a one-time only extension of the standby diesel generator (SDG)
allowed outage time for a cumulative 21 days on ``A'' train SDG. In
addition, it would also allow a one-time only extension of the allowed
outage time on ``A'' train essential cooling water loop for a
cumulative 7 days. This one-time only change would become effective on
April 10, 1996, and expire on May 15, 1996.Date of individual notice in
the Federal Register: March 8, 1996 (61 FR 9502)
Expiration date of individual notice: April 8, 1996
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 1, 1995, as supplemented by letters
dated June 22, August 28, November 22, and December 19, 1995, and
January 4, January 8 (two letters), and January 23, 1996
Description of amendment request: The proposed amendment would
provide a special test exception that would allow an extension of the
standby diesel generator (SDG) allowed outage time for a cumulative 21
days on each SDG once per fuel cycle, and it would also allow an
extension of the essential cooling water (ECW) loop allowed outage time
for a cumulative 7 days on each ECW loop once per fuel cycle. These
extended allowed outage times will be used to perform required
inspections and maintenance on the SDGs and the ECW system during power
operation.
Date of individual notice in the Federal Register: February 8, 1996
(61 FR 4805)
Expiration date of individual notice: March 11, 1996
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: March 1, 1996 (supersedes December 11,
1995, application)
Description of amendment request: The proposed amendment would
revise Technical Specification Section 4.7, ``Surveillance Requirements
for Primary Containment Automatic Isolation Valves.'' Specifically, the
proposed amendment would revise the replacement frequency of the seat
seals for the drywell and suppression chamber purge and vent valves
from every 5 years to every six operating cycles.
Date of individual notice in the Federal Register: March 8, 1996
(61 FR 9504)
Expiration date of individual notice: April 8, 1996
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: November 7, 1995, as
supplemented by letter dated January 17, 1996.
Brief description of amendments: These amendments adopt the
improved Standard Technical Specifications (NUREG-1432) format and
content of Section 5.0, ``Design Features,'' as modified by approved
changes to the improved Standard Technical Specifications.
Date of issuance: March 6, 1996
Effective date: March 6, 1996, to be implemented within 45 days of
the date of issuance.
Amendment Nos.: Unit 1 - Amendment No. 104; Unit 2 - Amendment No.
93; Unit 3 - Amendment No. 76
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65673) The January 17, 1996, supplemental letter provided clarifying
information and did not change the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated March 6, 1996.No
significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
[[Page 13535]]
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 16, 1996
Brief description of amendment: The amendment allows a one-time
extension for the performance of the trip actuating device operational
test for one of the safety injection manual initiation switches listed
in Technical Specification Table 4.3-2, Item 1a.Date of issuance: March
11, 1996
Effective date: March 11, 1996
Amendment No. 63
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: Yes (61 FR 7125). That 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 March 27, 1996, but indicated that if the Commission makes a
final no significant hazards consideration determination any such
hearing would take place after issuance of the amendment. The
Commission's related evaluation of the amendment, finding of exigent
circumstances, and final determination of no significant hazards
consideration is contained in a Safety Evaluation dated March 11, 1996
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, Docket
Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and
2, Will County, Illinois
Date of application for amendments: January 11, 1996
Brief description of amendments: The amendments revise the action
statements and allowed outage time for inoperability of one channel and
both channels of source range neutron flux instrumentation in Shutdown
Modes 3, 4, and 5.
Date of issuance: March 15, 1996
Effective date: March 15, 1996
Amendment Nos.: 80, 80, 72, and 72
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 31, 1996 (61 FR
3509) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 15, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: November 14, 1995, as
supplemented January 4, 1996 and February 29, 1996.
Brief description of amendments: The amendments revise the
Technical Specifications to incorporate 10 CFR Part 50, Appendix J,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' Option B.
Date of issuance: March 11, 1996 Effective date: Immediately, to be
implemented no later than June 30, 1996.
Amendment Nos.: 110 and 95
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 7, 1995 (60 FR
62896) The January 4, 1996, submittal provided additional clarifying
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
March 11, 1996. No significant hazards consideration comments received:
No
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: September 20, 1995, as
supplemented December 18 and December 22, 1995.
Brief description of amendment: The amendment allows a one-time
surveillance interval extension for certain 18-month surveillances
listed in new Technical Specification Tables 4.0.2-1 and 4.0.2-2. Date
of issuance: March 1, 1996
Effective date:
March 1, 1996, with full implementation within 90 days.
Amendment No.: 106
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58400). The December 18, 1995, letter corrected a typographical
error on one of the proposed TS pages and provided a corrected Table of
Contents page to reflect the addition of the new Tables. The December
22, 1995, letter provided additional information on the licensee's
review of historical plant drift data. This information was within the
scope of the original application and did not change the staff's
initial no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated March 1, 1996.No significant hazards
consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: November 10, 1995
Brief description of amendments: The amendments revise the
Technical Specifications for containment systems to reflect the
adoption of the requirements of 10 CFR Part 50, Appendix J, Option B,
and the implementation of a performance-based containment leak-rate
testing program at the Edwin I. Hatch Nuclear Plant, Units 1 and 2.
Date of issuance: March 6, 1996
Effective date: As of the date of issuance to be implemented within
90 days
Amendment Nos.: Unit 1 - 200 - Unit 2 - 141
Facility Operating License Nos. DPR-57 and NPF-5. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65679) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 6, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
[[Page 13536]]
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: December 5, 1995
Brief description of amendment: The amendment revises the submittal
date for the Annual Exposure Data Report bringing Oyster Creek into
conference with 10 CFR 20.2206 and relaxes an overly restrictive
administrative requirement.
Date of Issuance: March 4, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 183
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1629). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated March 4, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: December 14, 1995
Brief description of amendment: The amendment modifies Technical
Specification 3.4.2, ``Flow Control Valves (FCVs),'' by deleting
Surveillance Requirement (SR) 3.4.2.2, which required periodic
verification that the average rate of movement of each reactor
recirculation system FCV was limited to less than or equal to 11% per
second in the opening and closing directions. Due to a plant
modification, the requirement is not applicable.
Date of issuance: March 11, 1996
Effective date: March 11, 1996
Amendment No.: 103
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1630) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 11, 1996. No significant hazards
consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: November 10, 1995
(AEP:NRC:0896X). This application superseded a request dated June 15,
1995 (AEP:NRC:0896V).
Brief description of amendments: The amendments change the 18-month
emergency diesel generator surveillance test from a 24-hour run to an
8-hour run and add voltage and frequency measurement and power factor
monitoring.
Date of issuance: March 11, 1996
Effective date: March 11, 1996, with full implementation within 45
days
Amendment Nos.: Unit 1 - 207, Unit 2 - 191
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65682) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 11, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: June 20, 1995, as supplemented
December 19, 1995.
Brief description of amendments: The amendments relocate the fire
protection program elements from the Technical Specifications and
incorporate, by reference, the NRC-approved Fire Protection Program and
major commitments, including the fire hazards analysis, into the
Updated Final Safety Analysis Report. In addition, the amendments
revise the operating licenses to include the NRC's standard fire
protection license condition.
Date of issuance: March 11, 1996
Effective date: March 11, 1996, with full implementation within 180
days
Amendment Nos.: Unit 1 - 208, Unit 2 - 192
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications and the operating licenses.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47620). The December 19, 1995, supplement clarified the license
conditions by providing specific approval dates for previous fire
protection safety evaluations. This information was within the scope of
the original application and did not change the staff's initial
proposed no significant hazards consideration determination.The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated March 11, 1996.No significant hazards
consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
MillstoneNuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: June 29, 1995
Brief description of amendment: The amendment revises the Technical
Specifications to extend the surveillance schedule from 18 months to
each refueling interval (nominally 24 months) for specifications
4.6.4.2, 4.7.1.2.1.c, 4.7.3.b, 4.7.4.b,and 4.7.10.e. It also deletes
specification 4.6.4.2.a and the phrase ``during shutdown'' from these
specifications.Date of issuance: March 4, 1996
Effective date: As of the date of issuance, to be implemented
within 90 days.
Amendment No.: 127
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58402) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 4, 1996. 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, CT 06360.
Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California
Date of application for amendment: January 18, 1996
Brief description of amendment: The amendment revises the combined
Technical Specifications (TS) for the Diablo Canyon Nuclear Power
Plant, Unit No. 1. TS 3.8.1.1, ``Electrical Power Systems - A.C.
Sources - Operating,'' is revised to allow operation of Unit 1 in Mode
3 (Hot Standby) during installation of a replacement non-vital
auxiliary transformer 11, for a one time
[[Page 13537]]
extension of up to 48 hours beyond the 72 hours allowed by TS 3.8.1.1,
Action Statement (a).
Date of issuance: March 8, 1996
Effective date: March 8, 1996
Amendment No.: Unit 1 - Amendment No. 111
Facility Operating License No. DPR-80: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 1, 1996 (61 FR
3737) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 8, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: December 27, 1995
Brief description of amendments: The amendments revised the
combined Technical Specifications (TS) 3/4.6.1.1, Containment
Integrity; 3/4.6.1.2, Containment Leakage; 3/4.6.1.3, Containment Air
Locks; 3/4.6.1.6, Containment Structural Integrity; 3/4.6.3,
Containment Isolation Valves; their associated Bases; and adds
Specification 6.8.4 j., Containment Leakage Rate Testing Program to
implement the performance based leakage rate testing program as
permitted by 10 CFR Part 50, Appendix J, rather than paraphrasing the
requirements of the regulation. These changes will support the
implementation of the performance based testing of Option B to Appendix
J, for Type A, B, and C containment leakage rate testing and the
appropriate rescheduling of testing.
Date of issuance: March 1, 1996 Effective date: March 1, 1996
Amendment Nos.: Unit 1 - Amendment No. 110; Unit 2 - Amendment No.
109
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 31, 1996 (61 FR
3502) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 1, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: April 13, 1994, as supplemented
December 6, 1995
Brief description of amendment: The proposed changes revise the
Quality Assurance audit frequencies in the Hope Creek Technical
Specifications. These revisions will permit an audit frequency based on
performance and transfer subsequent control over the audit program to
the Updated Final Safety Analysis Report.
Date of issuance: March 11, 1996
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 95
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29633) The December 6, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination nor the original Federal Register
notice.The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 11, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: April 13, 1994, as supplemented
December 6, 1995.
Brief description of amendments: The proposed changes revise the
Quality Assurance audit frequencies in the Salem Unit Nos. 1 and 2
Technical Specifications. These revisions will permit an audit
frequency based on performance and transfer subsequent control over the
audit program to the Updated Final Safety Analysis Report.
Date of issuance: March 11, 1996
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment Nos. 181 and 162
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29633) The December 6, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination nor the original Federal Register
notice.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 11, 1996No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: December 8, 1995 (TS 93-09)
Brief description of amendments: The amendments revise the
setpoints and time delays for the auxiliary feedwater loss-of-power and
the 6.9-kilovolt shutdown board loss-of-voltage and degraded voltage
instruments.
Date of issuance: March 1, 1996
Effective date: March 1, 1996
Amendment Nos.: 219 and 209
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
181) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 1, 1996.No significant hazards
consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: January 4, 1996 (TS 95-22)
Brief description of amendments: The amendments change the
surveillance test frequency specified for the functional tests of the
containment, fuel storage pool, and control room radiation monitors
from monthly to quarterly.
Date of issuance: March 4, 1996
Effective date: March 4, 1996
Amendment Nos.: 220 and 210
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: January 31, 1996 (61 FR
3503)
[[Page 13538]]
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated Macrh 4, 1996.No significant hazards
consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: January 16, 1996, and supplement
dated March 1, 1996
Brief description of amendment: This amendment approves that part
of the request that defers the drywell bypass leakage test during the
current refueling outage. The remainder of the licensee's request is
still under NRC staff review.
Date of issuance: March 8, 1996
Effective date: March 8, 1996
Amendment No. 82
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1996 (61 FR
3951) The March 1, 1996, supplemental letter was clarifying in nature
and did not affect the initital no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 8, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: December 9, 1994, as
supplemented by letters dated September 13, 1995, and February 9, 1996.
Brief description of amendment: The amendment revises Technical
Specifications (TS) 4.3.2.2, TS 4.7.1.2.1, and the Bases for TS 3/4
7.1.2 to decrease the frequency of auxiliary feedwater pump testing,
remove inconsistencies in testing requirements for the turbine-driven
auxiliary feedwater pump, and clarify performance parameters in the TS
Bases.
Date of issuance: March 11, 1996
Effective date: March 11, 1996, to be implemented within 30 days
from the date of issuance.
Amendment No.: 108
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6314). The September 13, 1995, and February 9, 1996, supplemental
letters provided additional clarifying information and did not change
the original no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated March 11, 1996.No significant hazards
consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: September 19, 1995
Brief description of amendments: The amendments revised the maximum
allowable power range neutron flux high setpoints for operation with
inoperable main steam safety valves.
Date of issuance: March 6, 1996
Effective date: March 6, 1996
Amendment Nos.: 199 and 180
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: October 25, 1995 (60 FR
54724) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 6, 1996No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: November 22, 1995
Brief description of amendment: The amendment replaces the
Technical Specification (TS) requirements associated with the boron
dilution mitigation system (BDMS) with alarms, indicators, procedures
and controls to allow proper resolution of potential boron dilution
events.
Date of issuance: March 1, 1996
Effective date: March 1, 1996, to be implemented prior to the
startup from the eighth refueling outage.
Amendment No.: 96
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 31, 1996 (61 FR
3503) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 1, 1996.No significant hazards
consideration comments received: No. Local Public Document Room
locations: Emporia State University, William Allen White Library, 1200
Commercial Street, Emporia, Kansas 66801 and Washburn University School
of Law Library, Topeka, Kansas 66621
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: December 20, 1995, as supplemented by
letter dated February 8, 1996.
Brief description of amendment: The amendment revises the Technical
Specifications to reflect the approval of the use of 10 CFR Part 50,
Appendix J, Option B for the Wolf Creek Generating Station containment
leakage rate test program.
Date of issuance: March 1, 1996
Effective date: March 1, 1996, to be implemented prior to startup
from the eighth refueling outage.
Amendment No.: 97
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 31, 1996 (61 FR
3504) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 1, 1996.No significant hazards
consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: December 13, 1995
Brief description of amendment: The amendment revises the minimum
and maximum flow requirements for the centrifugal charging pumps (CCPs)
and safety injection pumps (SIPs) specified in Technical Specification
(TS) Surveillance Requirement 4.5.2.h. Specifically, the amendment (1)
decreases the minimum limits on the sum of the injection line flow
rates,
[[Page 13539]]
excluding the highest flow rate, from 346 gallons per minute (gpm) to
330 gpm for the CCPs and from 459 gpm to 450 gpm for the SIPs, and (2)
revises the maximum pump flow rate for the SIPs from 665 to 670 gpm,
but retains the CCPs maximum pump flow rate at its current value of 556
gpm.Date of issuance: March 5, 1996
Effective date: March 5, 1996, to be implemented prior to startup
from the eighth refueling outage.
Amendment No.: 98
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1639) The February 5, 1996, supplemental letter provided additional
clarifying information and did not change the original no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
March 5, 1996.No significant hazards consideration comments received:
No. Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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 room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By April 26, 1996, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
[[Page 13540]]
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. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
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, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
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: March 6, 1996
Brief description of amendment: This amendment revises TS 3/4 5.2,
ECCS SUBSYSTEMS - T avg greater than or equal to 280 deg.F by
modifying Surveillance Requirement 4.5.2.b to defer venting of the
Emergency Core Cooling System flow path which does not have manual
venting capability until the tenth refueling outage.
Date of issuance: March 7, 1996
Effective date: March 7, 1996
Amendment No: 208
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No. The Commission's related
evaluation of the amendments, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated March 7, 1996.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Dated at Rockville, Maryland, this 20th day of March 1996.
For the Nuclear Regulatory Commission
Steven A. Varga, Director,
Division of Reactor Projects - I/II,Office of Nuclear Reactor
Regulation
[Doc. 96-7259 Filed 3-26-96; 8:45 am]
BILLING CODE 7590-01-F
