[Federal Register Volume 59, Number 119 (Wednesday, June 22, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10622]
[[Page Unknown]]
[Federal Register: June 22, 1994]
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UNITED STATES 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 May 27, 1994, through June 10, 1994. The
last biweekly notice was published on June 8, 1994 (59 FR 29623).
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, 11555 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
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By July 22, 1994, 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 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, 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 20555, 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 20555, and at the local public document
room for the particular facility involved.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: April 26, 1994
Description of amendment request: The proposed amendment would
revise the Technical Specifications to change the Table 3.5-1 High
Containment Pressure ( Hi Level), Safety Injection Setting Limit from
less than or equal to 2.0 psig to less than or equal to 5.0 psig.
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 a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
It is proposed that the High Containment Pressure (Hi Level)
actuation setting of [less than or equal to] 2.0 psig be revised to
[less than or equal to] 5.0 psig. This additional operating
flexibility will decrease the frequency of Containment venting
necessary to relieve containment of
non-condensible gases which build up during normal operation.
Based upon a statistical analysis of the containment pressure
channel uncertainty for a 30 month operating cycle, a margin must be
allowed between the Technical Specification limit (plant setting)
and the Safety Analysis limit so that the Safety Analysis limit(s)
will not be exceeded under the worst circumstances. For a Technical
Specification value of [less than or equal to] 5.0 psig, the
corresponding Safety Analysis limit must be increased to 10 psig to
provide margin for the channel statistical allowance. A safety
evaluation performed pursuant to 10 CFR 50.59 is on file which
supports a change in the Safety Analysis limit from 7.3 psig
(current value) to 10.0 psig. Key conclusions of the Safety
Evaluation are that neither the probability nor the consequences of
an accident or malfunction of equipment important to safety
previously evaluated in the Safety Analysis report would be
increased.
Thus, assurance is provided that appropriate protective actions
in accordance with the Technical Specifications will be taken so
that Safety Analysis limits are not exceeded.
2. The possibility of a new or different kind of accident from
any previously analyzed has not been created.
The proposed change in the Technical Specification limit
together with the change in the Safety Analysis limit provides
adequate margin to accommodate instrument channel uncertainty over a
30 month operating cycle. Plant equipment, which would be set at the
Technical Specification limit, will therefore provide protective
functions to assure that safety analysis limits are not exceeded.
This would prevent the possibility of a new or different kind of
accident from that previously evaluated from occurring.
3. There has been no reduction in the margin of safety.
The proposed change to the Technical Specification limit would
decrease the frequency of containment purges necessary to vent the
build up of non-condensible gases during normal operation. This
would result in a decrease in the amount of radioactivity discharged
to the environment (due to decay), decrease the potential for high
Containment pressure alarms and increase the margin for an ESF trip.
The change to the Safety Analysis limits, justified by a safety
Evaluation performed in accordance with 10 CFR 50.59, assures
sufficient margin exists to accommodate channel instrument
uncertainty over the maximum operating cycle length. This margin is
necessary so that safety functions will occur and Safety Analysis
limits will be preserved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Michael L. Boyle
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: May 24, 1994
Description of amendment request: The proposed amendments would
transfer the boron concentration in Technical Specification (TS) 3.9.1
for the reactor coolant system and the refueling canal during MODE 6,
and the boron concentration in TS 3.9.12 for the spent fuel pool from
the TS to the Core Operating Limits Report (COLR). The application is
submitted in response to the guidance in Generic Letter 88-16 which
addresses the transfer of fuel cycle-specific parameter limits from the
TS to the COLR.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following analysis, performed pursuant to 10 CFR 50.91,
shows that the proposed amendment will not create a significant
hazards consideration as defined by the criteria of 10 CFR 50.92.
1. This amendment will not significantly increase the
probability or consequence of any accident previously evaluated.
No component modification, system realignment, or change in
operating procedure will occur which could affect the probability of
any accident or transient. The relocation of boron concentration
values to the COLR is an adminsitrative change which will have no
effect on the probability or consequences of any previously-analyzed
accident. The required values of boron concentration will continue
to be determined through use of approved methodologies.
2. This amendment will not create the possibility of any new or
different accidents not previously evaluated.
No component modification or system realignment will occur which
could create the possibility of a new event not previously
considered. The administrative change of relocating parameters to
the COLR, in this case boron concentration, cannot create the
probability of an accident.
3.This amendment will not involve a significant reduction in a
margin of safety.
Required boron concentrations will remain appropriate for each
cycle, and will continue to be calculated using approved
methodologies. There is no significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: David B. Matthews, Director
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: May 12, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification Sections 3.1 and 4.1 for Protective
Instrumentation, the associated bases, and tables to increase the
surveillance test intervals (STIs) and add allowable out-of-service
times (AOTs). All proposed STI and AOT changes are in accordance with
General Electric Company Licensing Topical Reports (LTRs) which have
been previously reviewed and approved by the NRC staff. Also, AOTs are
clarified in accordance with the most recently approved BWR Owners'
Group letters which were used in the development of NUREG-1433
``Standard Technical Specifications, General Electric Plants, BWR/4.''
The Technical Specification changes will permit specified Channel Tests
to be conducted quarterly rather than weekly or monthly. The amendment
will enhance operational safety by reducing 1) the potential for
inadvertent plant scrams, 2) excessive test cycles on equipment, and 3)
the diversion of plant personnel and resources on unnecessary testing.
Two additional technical changes are proposed. The first change
involves extending the Channel Calibration interval for average power
range monitor (APRM) scram instrumentation from weekly to quarterly.
GPUN has evaluated the effect of drift on the setpoint over the longer
interval for this instrumentation and found it to be acceptable. The
second change would add a quarterly Channel Calibration requirement for
High Drywell Pressure (for Core Cooling) and Turbine Trip Scram
instrumentation. This would be a new requirement not currently
incorporated in the Technical Specifications.
Nineteen editorial changes have been incorporated in
Instrumentation Sections 3.1 and 4.1 to provide clarity and
consistency. These items are editorial only and do not alter the
meaning or intent of any requirements. Examples of editorial changes
are: 1) capitalize definitions where used, 2) punctuation and
grammatical corrections, 3) ensuring consistency in STI nomenclature,
and 4) reformat of tables. A table note and its associated footnote
were deleted which involved a 1985 licensing condition which is no
longer applicable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analyses of the issue of no significant hazards
consideration, which is presented below:
NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION OF TECHNICAL
CHANGES
1. The operation of the Oyster Creek Nuclear Generating Station,
in accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The generic analysis contained in LTR NEDC-30851P-A assessed the
impact of changing RPS [reactor protection system] STIs and adding
AOTs on RPS failure frequency, scram frequency and equipment
cycling. Specifically, Section 5.7.4, ``Significant Hazards
Assessment,'' of NEDC-30851P-A states that:
Fewer challenges to the safeguards system, due to less frequent
testing of the RPS, conservatively results in a decrease of
approximately one percent in core damage frequency. This decrease is
based upon the following:
Based on the plant-specific experience presented in Appendix J,
the estimated reduction in scram frequency (0.3 scrams/yr)
represents a 1 to 2 percent decrease in core damage frequency based
on the BWR plant-specific Probabilistic Risk Assessments (PRAs)
listed in Table 5-8.
The increase in core damage frequency due to less frequent
testing is less than one percent. This increase is even lower (less
than 0.01 percent) when the changes resulting from the
implementation of the Anticipated Transients Without Scram (ATWS)
rule are considered. Therefore, this increase is more than offset by
the decrease in CDF [core damage frequency] due to fewer scrams.
The effect of reducing unnecessary cycles on RPS equipment,
although not easily quantifiable also results in a decrease in core
damage frequency.
The overall impact on core damage frequency of the changes in
allowable out-of-service times is negligible.
The BWR Owners' Group concluded that the proposed changes do not
significantly increase the probability or consequences of an
accident previously evaluated since the increase in probability of a
scram failure due to RPS unavailability is insignificant. The
overall probability of an accident is decreased as the time RPS
logic operates as designed is increased resulting in less
inadvertent scrams during testing and repair. The plant-specific
evaluation performed by GPUN and GE demonstrates that while the
Oyster Creek RPS differs from the generic model analyzed in the RPS
LTR (NEDC-30851P-A), the net effect of the differences do not alter
the generic conclusions. The AOTs proposed for RPS instrumentation
are based on improved wording developed for use in NUREG 1433,
``Standard Technical Specifications, General Electric Plants, BWR/
4,'' which ensures a loss of function does not occur. In addition,
the change to the APRM Scram Channel Calibration surveillance
interval from weekly to quarterly has been evaluated by GPUN to
determine the effect on setpoint drift. The results of the
evaluation show acceptable performance of this scram parameter
ensuring that the safety analysis remains valid. The clarification
that a Channel Calibration is not applicable to Turbine Trip Scram
instrumentation is appropriate since this trip parameter senses
turbine stop valve position via limit switches which are fixed in
position and adjusted, as necessary, during valve maintenance. This
trip parameter and its switch adjustment methods are similar to the
Main Steamline Isolation Valve [MSIV] Scram for which the Technical
Specifications require only a Channel Test.
LTR NEDC-30936P-A (Parts 1 and 2) contains an assessment of the
impact of changing STIs and AOTs for BWR ECCS Actuation
Instrumentation. Section 4.0, ``Technical Assessment of Changes,''
of NEDC-30963P-A (Part 2) states that:
The results indicate an insignificant (less than 5E-7 per year)
increase in water injection function failure frequency when STIs are
increased from 31 days to 92 days, AOTs for repair of the ECCS
actuation instrumentation are increased from one hour to 24 hours,
and AOTs for surveillance testing are increased from two to six
hours. For all four BWR models the increase represents less than 4%
increase in failure frequency. However, when other factors which
influence the overall plant safety are considered, the net result is
judged to be an improvement in plant safety.
From this generic analysis, the BWR Owners' Group concluded that
the proposed changes do not significantly increase the probability
or consequences of an accident previously evaluated since the
increase in probability of a water injection failure due to ECCS
instrumentation unavailability is insignificant and the net result
is judged to be an improvement in plant safety. The plant-specific
evaluation performed by GPUN and GE demonstrates that while the
Oyster Creek ECCS differs from the generic model analyzed in LTR
NEDC-30936P-A, the net effect of the plant-specific differences do
not alter the generic conclusions. The addition of a quarterly
Channel Calibration STI for the High Drywell Pressure ECCS
initiation parameter is consistent with the calibration interval
requirement for other similar instrumentation at Oyster Creek and
ensures the regular performance of calibrations. This is a new
requirement not currently contained in the Technical Specifications
and experience performing the High Drywell Pressure (Core Cooling)
instrument calibration at a quarterly interval has proven adequate
for instrument performance monitoring.
LTRs NEDC-30851P-A, Supplement 2 and NEDC-31677P-A contain
generic analyses assessing the impact of changing STIs and AOTs for
BWR Isolation Actuation Instrumentation which are common or not
common to RPS and ECCS instrumentation. Section 4.0, ``Summary of
Results,'' of NEDC-30851P-A, Supplement 2 states that:
The results indicated that the effects on probability of failure
to initiate isolation are very small and the effects on probability
or frequency of failure to isolate are negligible in nearly every
case. In addition, the results indicated that increasing the AOT to
24 hours for tests and repairs has a negligible effect on the
probability of failure of the isolation function. These combined
with changes to the testing intervals and allowed out-of-service
times for RPS and ECCS instrumentation provide a net improvement to
plant safety and operations.
and Section 5.6, ``Assessment of Net Effect of Changes,'' of
NEDC-31677P-A states that:
A reduction in core damage frequency (CDF) of at least as much
as estimated in the ECCS instrumentation analysis can be expected
when the isolation actuation instrumentation STIs are changed from
one month to three months. The chief contributor to this reduction
is the channel functional tests for the MSIVs. Inadvertent closure
of the MSIVs will cause an unnecessary plant scram. This reduction
in CDF more than compensates for any small incremental increase (10%
or 1.0E-07/year) in calculated isolation function failure frequency
when the STI is extended to three months.
Based on this generic analysis, the BWR Owners' Group concluded
that the proposed changes do not significantly increase the
consequences of an accident previously evaluated since the increase
in probability of an isolation failure due to isolation
instrumentation unavailability is insignificant. The proposed
wording of the AOTs is based on the clarifications used in the
development of NUREG 1433, ``Standard Technical Specifications,
General Electric Plants, BWR/4,'' which ensures a loss of function
does not occur where applied to isolation actuation instrumentation.
LTR NEDC-30851P-A, Supplement 1 contains a generic analysis
assessing the impact of changing control rod block STIs on Rod Block
failure frequency. Section 5 (Brookhaven National Laboratory
Technical Evaluation Report - Attachment 2 to the NRC SER) of NEDC-
30851P-A, Supplement 1 states that:
The BWROG proposed changes to the Technical Specifications
concerning the test requirements for BWR control rod block
instrumentation. The changes consist of increasing the surveillance
test intervals form one to three months. These test interval
extensions are consistent with the already approved changes to STIs
for the reactor protection system. The technical analysis reviewed
and verified as documented herein indicates that there will be no
significant changes in the availability of the control rod block
function if these changes are implemented. In addition, there will
be a negligible impact on the plant core melt frequency due to the
decreased testing.
Bases contained in GE Topical Report GENE-770-1-A assessed the
impact of changing STIs and AOTs on failure frequency for selected
systems. Section 2.0, ``Summary,'' of GENE-770-06-1-A states that:
Technical bases are provided for selected proposed changes to
the instrumentation STIs and AOTs that were identified in the BWROG
Improved BWR Technical Specification activity. These STI and AOT
changes are consistent with approved changes to the RPS, ECCS, and
isolation actuation instrumentation. These proposed changes do not
result in a degradation to overall plant safety.
The BWR Owners' Group concluded from the generic analysis in
NEDC-30851P-A, Supplement 1 and the bases in GENE-770-06-1-A that
the proposed changes do not significantly increase the probability
or consequences of an accident previously evaluated. GPUN's
utilization of GENE-770-06-1-A is limited to the identified AOTs for
Control Rod Block instrumentation analyzed in NEDC-30851P-A since
the Control Rod Block LTR did not explicitly address AOTs.
2. The operation of Oyster Creek Nuclear Generating Station, in
accordance with the proposed amendment, will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The addition of allowable out-of-service times (AOTs) consistent
with wording developed for use in Improved Standard Technical
Specifications to ensure no loss of function and the revision of
surveillance test intervals (STIs) does not alter the function of
RPS, ECCS, Isolation or Rod Block instrumentation nor involve any
type of plant modification. No new modes of plant operation are
involved with the changes.
Adding a quarterly Channel Calibration STI for High Drywell
Pressure instrumentation (for Core Cooling) establishes a
requirement in the Technical Specifications which is not currently
incorporated. This is an additional requirement beyond that already
in place for this instrumentation and will not alter its operation
since by their nature STIs ensure proper instrument performance. The
clarification that a Channel Calibration is not applicable to
Turbine Trip Scram instrumentation is appropriate since this trip
parameter senses turbine stop valve position via limit switches
which are fixed in position and adjusted during valve maintenance.
This trip parameter and its switch adjustment methods are similar to
the Main Steamline Isolation Valve Scram for which the Technical
Specifications require only a Channel Test. Revising the Channel
Calibration STI for APRM Scram instruments from weekly to quarterly
allows these instruments to benefit from the Channel Test STI change
provided by the generic analysis in the RPS LTR. The benefits
include a significant reduction in the number of half-scram states
the plant will undergo reducing the potential for inadvertent plant
trips. The effect of setpoint drift over the longer interval has
been evaluated and found acceptable.
The proposed changes will not alter the physical characteristics
of any plant systems or components and all safety-related systems
and components remain within their applicable design limits. Thus,
system and component performance is not adversely affected by these
changes, thereby assuring that the design capabilities of those
systems and components are not challenged in a manner not previously
assessed so as to create the possibility of a new or different kind
of accident.
3. The operation of the Oyster Creek Nuclear Generating Station,
in accordance with the proposed amendment, will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed and approved the generic studies
contained in the GE Licensing Topical Reports and has concurred with
the BWR Owners' Group that the proposed changes do not significantly
affect the availability of RPS, ECCS Actuation, Isolation Actuation
and Control Rod Block instrumentation. The proposed addition of
allowable out-of-service times for instruments addressed by the LTRs
provides reasonable times for making repairs and performing tests.
The lack of sufficient out-of-service time provided in current
Technical Specifications, increases the potential for an inadvertent
scram or equipment actuation. The proposed AOTs provide realistic
times to complete required actions without increasing overall
instrument failure frequency and ensure that no loss of function
occurs, therefore, there is no significant reduction in the margin
of safety.
The LTRs demonstrate that extending surveillance test intervals
does not result in significant changes in the probability of
instrument failure. Where Channel Calibration frequency has not
changed, assurance exists that setpoints will not be affected by
drift. In the case of the APRM Scram Channel Calibration, the
proposed change to quarterly from weekly has been evaluated and
found acceptable. Expected instrument performance over the extended
interval will assure that applicable safety analyses will continue
to be met. In addition, other instrumentation was evaluated for
drift effects of setpoints and was found acceptable. The addition of
a quarterly Channel Calibration interval for High Drywell Pressure
(for Core Cooling) is consistent with Channel Calibration STIs for
most other instrumentation at Oyster Creek and has been the interval
used to achieve an adequate level of instrument performance
monitoring. The clarification that a Channel Calibration is not
applicable to Turbine Trip Scram instrumentation ensures consistency
in the establishment of surveillance requirements. This trip
parameter senses turbine stop valve position via limit switches
which are fixed in position and adjusted during valve maintenance.
This trip parameter and its switch adjustment methods are similar to
the Main Steamline Isolation Valve Scram for which the Technical
Specifications require only a Channel Test. These proposed changes,
when coupled with the reduced probability of test-induced plant
transients and equipment failures, do not result in a reduction in
the margin of safety.
No Significant Hazards Consideration Evaluation For Editorial
Changes
The above nineteen proposed changes are editorial in nature and
are typical example I.c.2.e.i in 51FR7744. Therefore, they do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The editorial changes described above do not change the design
or operation of any structure, system or component relied upon to
prevent or mitigate the consequences of any accident evaluated.
These editorial changes also do not add new structures, systems or
components which may have an effect on existing elements of the
facility. The changes proposed correct, clarify and/or retain
existing requirements.
2. Create the possibility of a new or different kind of accident
form any accident previously evaluated.
Since neither physical changes to the facility nor changes in
its operation are involved in the proposed editorial changes to the
Technical Specifications, there is no possibility for creation of a
new or different kind of accident.
3. Involve a significant reduction in the margin of safety.
Facility configuration and operation are unaffected by the
proposed editorial changes. As a result no changes in margin of
safety occur.
The editorial changes described and evaluated above are purely
administrative to achieve consistency or correct an error in the
Technical Specifications.
The NRC staff has reviewed the licensee's analyses and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied for both the technical issues and editorial changes.
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, New Jersey
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. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: February 10, 1994
Description of amendment request: The revision proposed by
Technical Specification Change Request (TSCR) No. 230 to the Technical
Specifications would revise specification 3.7.2.c, ``Unit Electric
Power System,'' to eliminate testing of an emergency diesel generator
(EDG) when the redundant EDG is inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment serves to assure that an EDG is always
available to perform on demand and the lower number of demands for
performance reduce the probability of equipment failure. The
required action no longer requires a ``test'' be performed.
Therefore, the word ``test'' has been deleted from TS 3.7.2.c. The
change is administrative. Since the proposed amendment does not
affect the design or performance of the diesel generators or their
ability to perform their design function, the change will not result
in an increase in the consequences or probability of an accident
previously analyzed. The proposed change will increase diesel
generator reliability, thereby increasing overall plant safety.
2. Operation of the facility in accordance with the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated. Accidents
involving loss of off-site power and single failure have been
previously evaluated. The change does not introduce any new mode of
plant operation or new accident precursors, involve any physical
alterations to plant configurations, or make any changes to system
setpoints which could initiate a new or different kind of accident.
3. Operation of the facility in accordance with the proposed
amendment does not involve a significant reduction in a margin of
safety. This change does not result in a reduction in the margin of
safety since there is no margin of safety associated with the
supplemental immediate and daily testing of the operable EDG. If a
margin of safety were presumed to exist, no reduction would result
because of the proposed amendment: no physical modification to the
plant or change to procedurally prescribed operator actions resulted
from 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: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
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
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine
YankeeAtomic Power Station, Lincoln County, Maine
Date of amendment request: May 25, 1994
Description of amendment request: The proposed amendment would 1)
allow entry through an operable personnel air lock hatch to perform
surveillance testing, repair an inoperable hatch, or perform other
necessary activities inside containment, 2) update plant Technical
Specifications to reflect a previous change to the list of containment
boundary valves, 3) add a new exception to allow quarterly surveillance
testing of the excess flow check valves, 4) add a new exception to
allow periodic preventive maintenance on control room ventilation
lasting up to 30 minutes per calendar quarter without a written report
of such inoperability, and 5) make related administrative changes to
reflect and clarify items 1 through 4 above.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's analysis is
presented below:
1. The proposed amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Containment air lock hatch entry, surveillance testing of the
excess flow check valves, and preventive maintenance of control room
ventilation are of short duration and do not alter any associated
remedial action completion times, or the requirements of Technical
Specification 3.0.A. If necessary, prompt operator action to restore
containment integrity, excess flow check valve position, or control
room ventilation is assured by plant operators, or individual(s)
procedurally dedicated to perform such restoration. The subject
containment boundary valves are manual containment isolation valves,
and the current specification allows them to be repositioned under
administrative control without compensatory measures to isolate the
penetration. The boundary valves to be added remain closed during
power operation, and are opened only after the reactor is shut down
and cooldown has begun. The boundary valves to be deleted are open
only during plant heatup.
The staff therefore concludes that implementation of the proposed
change will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change to containment air lock hatch entry,
surveillance testing of the excess flow check valves, and preventive
maintenance of the control room ventilation system, will not affect
equipment reliability when such equipment is required to be
operable. The Limiting Conditions for Operation and Remedial Actions
for these items remain unchanged to govern operability of the
equipment. The containment boundary valves being added are closed
when the reactor is at power, and are opened only after the reactor
is shut down. The boundary valves being deleted are open only during
plant heatup. The subject boundary valves are manual containment
isolation valves, and the current specification allows them to be
repositioned under administrative control without compensatory
measures to isolate the penetration.
The staff therefore concludes that implementation of the
proposed change will not create any new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change would allow excess flow check valves to be
exercised through approximately 1.5 inches of valve travel on a
quarterly basis without declaring the valves inoperable or taking
compensatory measures. Such testing constitutes approximately 15
minutes per calendar quarter, during which time containment
isolation can easily be reestablished. Similarly, access through an
operable air lock hatch would allow the hatch to be open for only a
short period of time and while under control of an individual
dedicated to operating the hatch. The proposed change also permits
the control room ventilation system to be inoperable for 30 minutes
per calendar quarter, without a written report of such
inoperability. Because of the short time during which these systems
are unavailable, and because operation is easily reestablished,
there is no significant reduction in a margin of safety. The
containment boundary valves being added are closed when the reactor
is at power, and are opened only after the reactor is shut down. The
boundary valves being deleted are open only during plant heatup. The
subject boundary valves are manual containment isolation valves, and
the current specification allows them to be repositioned under
administrative control without compensatory measures to isolate the
penetration.
The staff therefore concludes that implementation of the proposed
change would not involve a significant reduction in a margin of safety.
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, Maine 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 83 Edison Drive, Augusta, Maine 04336
NRC Project Director: Walter R. Butler
Northeast Nuclear Energy Company (NNECO), Docket No. 50-
245,Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: April 29, 1994
Description of amendment request: The amendment would change the
requirement for reactor operators (RO) in Table 6.2-1 from 2 to 3 for
the RUN, STARTUP/HOT STANDBY and HOT SHUTDOWN conditions. In addition,
two typographical corrections are made to page 6-4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed changes in accordance with
10CFR50.92 and concluded that the changes do not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed changes do not involve a significant
hazards consideration because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Accident analyses for Millstone Unit No. 1 do not require a
specific number of operators. Increasing the Technical Specification
minimum to require a third RO does not decrease the effectiveness of
the shift staff in response to normal or abnormal conditions. In
fact, the third RO enhances the ability of the operating crew to
mitigate complex transients which could occur during beyond design
basis events. The shifts have trained and functioned at the higher
staffing level for several years.
The typographical corrections to page 6-4 provide a clearer
representation of the required actions, and do not affect the intent
nor implementation of the specification.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of a previously analyzed
accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The addition of a third RO to the minimum shift-crew composition
required by Technical Specification Table 6.2-1 does not affect the
operation of the unit, nor does it change any of the operating
procedures, off-normal procedures, or EOPs [emergency operating
procedures]. Staffing the control room with an additional operator
enhances the capability of the operating crew to mitigate
transients. Therefore, addition of a third RO to the minimum shift-
crew composition cannot create the possibility of a new or different
accident.
3. Involve a significant reduction in the margin of safety.
The proposed addition of a third reactor operator is to ensure
that sufficient operating staff is available to respond to complex
transients involving multiple equipment failures. Ensuring that
sufficient resources are available to cope with beyond design basis
event scenarios provides an increase 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, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: May 6, 1994
Description of amendment request: The proposed amendment would
modify the Limiting Conditions for Operation (LCO) for the Millstone
Unit 2 Technical Specifications 3.8.2.3 and 3.8.2.4 and the
surveillance requirement of TS 4.8.2.3.2.c.3. These changes relate to
the amperage requirements and the charging capability of the DC
distribution systems.
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 changes do not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
In 1993, revised battery and battery charger sizing calculations
demonstrated that a charger capacity of 400 amperes is sufficient to
provide the continuous DC loads, and is capable of recharging a
fully discharged station battery in a timely manner consistent with
the design basis discussed in Section 8.5.3.1 of the Millstone Unit
No. 2 FSAR [Final Safety Analysis Report]. The calculations
determined that the largest continuous load was 154 amperes;
therefore, 400 amperes of charging capacity could provide 246
amperes to recharge a battery.
The calculations conservatively demonstrated that this charging
capacity could recharge a battery in 10.37 hours. This recharging
time is well within the 12-hour recharging time discussed in Section
8.5.3.1 of the Millstone Unit No. 2 FSAR. Additionally, this
recharging time is more conservative than the 24-hour recharging
time stated in Section 8.3.2 of the original Safety Evaluation for
Millstone Unit No. 2. Therefore, the proposed changes do 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 proposed LCO and surveillance changes do not alter the
existing DC bus configuration, as described in Section 8.5.3.1 of
the Millstone Unit No. 2 FSAR. This bus configuration has been
previously analyzed, and was found acceptable. The proposed changes
also meet the recharging time specified in the design basis.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously analyzed.
3. Involve a significant reduction in the margin of safety.
In 1993, revised battery and battery charger sizing calculations
demonstrated that a charger capacity of 400 amperes is sufficient to
provide the continuous DC loads, and is capable of recharging a
fully discharged station battery in a timely manner consistent with
the design basis discussed in Section 8.5.3.1 of the Millstone Unit
No. 2 FSAR. The calculations determined that the largest continuous
load was 154 amperes; therefore, 400 amperes of charging capacity
could provide 246 amperes to recharge a battery. The calculations
conservatively demonstrated that this charging capacity could
recharge a battery in 10.37 hours. This recharging time is well
within the 12-hour recharging time discussed in Section 8.5.3.1 of
the Millstone Unit No. 2 FSAR. Additionally, this recharging time is
more conservative than the 24-hour recharging time stated in Section
8.3.2 of the original Safety Evaluation for Millstone Unit No. 2.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: May 6, 1994
Description of amendment request: The proposed amendment would
provide additional Technical Specification requirements regarding non-
Quality Assurance (QA) equipment utilized to achieve feedwater
isolation in response to a main steam line break (MSLB) inside
containment. Specifically the amendment would incorporate additional
sections numbered 3/4.7.1.6, titled ``Plant Systems - Main Feedwater
Isolation Components (MFICs);'' 3/4.8.2.1A, titled '' Onsite Power
Distribution Systems - A.C. Distribution - Operating;'' and 3/4.8.2.5,
titled ``Onsite Power Distribution Systems (Turbine Battery) - D.C.
Distribution - Operating.'' In addition, the proposed amendment would
modify the Index and the Bases to reflect the additional requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed changes in accordance with
10CFR50.90 and has concluded that the changes do not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed changes do not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Currently, the Millstone Unit No. 2 Technical Specifications
contain response time requirements for the feedwater isolation
valves to ensure rapid isolation of feedwater to the steam
generators and to maintain the peak containment pressure below the
containment design pressure of 54 psig. However, clear Action
Statements specifying operability requirements for the non-QA
equipment associated with feedwater isolation are not included
within the Millstone Unit No. 2 Technical Specifications. NNECO's
proposal to add sections 3/4.7.1.6, 3/4.8.2.1A, and 3/4.8.2.5 into
the Millstone Unit No. 2 Technical Specifications will incorporate
additional requirements regarding components that are credited to
provide feedwater isolation in the event of an MSLB inside
containment. These proposed changes will impose additional
limitations, restrictions, and controls not currently in place in
the Millstone Unit No. 2 Technical Specifications.
Additionally, NNECO's proposals to modify the Bases and the
Index of the Millstone Unit No. 2 Technical Specifications will: 1)
provide personnel with information concerning the additional
requirements, and 2) correct an editorial error. These proposed
changes to the Bases and the Index do not alter the manner in which
equipment is operated, nor do they affect equipment availability.
Based on the above, the proposed license amendment 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.
NNECO's proposal to add section 3/4.7.1.6, 3/4.8.2.1A, and 3/
4.8.2.5 into the Millstone Unit No. 2 Technical Specifications will
incorporate additional requirements regarding components that are
credited to provide feedwater isolation in the event of an MSLB
inside containment. These proposed changes will impose additional
limitations, restrictions, and controls not currently in place in
the Millstone Unit No. 2 Technical Specifications.
Additionally, NNECO's proposals to modify the Bases and the
Index of the Millstone Unit No. 2 Technical Specifications will: 1)
provide personnel with information concerning the additional
requirements, and 2) correct an editorial error. These proposed
changes to the Bases an the Index do not alter the manner in which
equipment is operated, nor do they affect equipment availability.
Based on the above, the proposed license amendment cannot create
the possibility of a new or different kind of accident from any
previously analyzed.
3. Involve a significant reduction in a margin of safety.
NNECO's proposal to add sections 3/4.7.1.6, 3/4.8.2.1A, and 3/
4.8.2.5 into the Millstone Unit No. 2 Technical Specifications will
incorporate additional requirements regarding components that are
credited to provide feedwater isolation in the event of an MSLB
inside containment. These proposed changes will impose additional
limitations, restrictions, and controls not currently in place in
the Millstone Unit No. 2 Technical Specifications.
Additionally, NNECO's proposals to modify the Bases and the
Index of the Millstone Unit No. 2 Technical Specifications will: 1)
provide personnel with information concerning the additional
requirements, and 2) correct an editorial error. These proposed
changes to the Bases and the Index do not alter the manner in which
equipment is operated, nor do they affect equipment availability.
Therefore, this proposed license amendment does not involve a
significant reduction in a margin of safety. In fact. The margin of
safety will be increased due to the imposition of restriction on the
non-QA equipment credited for feedwater isolation in the event of an
MSLB inside containment.
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 Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
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: May 6, 1994
Description of amendment request: The proposed amendment modifies
the monthly operational test of the reactor trip bypass breakers to
monthly staggered, such that each breaker is tested every 62 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in probability or consequences
of an accident previously evaluated.
Revising the technical specifications to require a staggered
monthly surveillance operational test of the reactor trip bypass
breakers (such that each breaker is tested every 62 days) will only
make operational testing of the reactor trip bypass breakers
consistent with operational testing of the trip breakers and the
automatic trip and interlock logic. It will also reduce cycling of
the reactor trip bypass breakers by eliminating the requirement to
test both bypass breakers during the monthly surveillance, thereby
reducing maintenance and surveillance time. The proposed changes do
not affect any of the design basis accidents nor are there any
malfunctions associated with these changes.
Additionally, this technical specification bases change only
clarifies both the meaning of a reactor trip breaker and trip
breaker train which have been included for completeness and clarity
concerning the reactor trip breaker system.
2. Create the possibility of a new or different kind of accident
previously evaluated.
Revising the technical specifications to require a staggered
monthly surveillance operational test of the reactor trip bypass
breakers (such that each breaker is tested every 62 days) will only
make operational testing of the reactor trip bypass breakers
consistent with operational testing of the reactor trip breakers and
the automatic trip and interlock logic. There are no new failure
modes associated with the proposed changes. Since the plant will
continue to operate as designed, the proposed changes will not
modify the plant response to the point where it can be considered a
new accident.
3. Involve a significant reduction in a margin of safety.
Revising the technical specifications to require a staggered
monthly surveillance operational test of the reactor trip bypass
breakers (such that each breaker is tested ever 62 days) will only
make operational testing of the reactor trip bypass breakers
consistent with operational testing of the reactor trip breakers and
the automatic trip and interlock logic. It will also reduce cycling
of the reactor trip bypass breakers by eliminating the requirement
to test both bypass breakers during the monthly surveillance,
thereby reducing maintenance and surveillance time. The proposed
changes do not have any adverse impact on the protective boundaries
nor do they affect the consequences of any accident previously
analyzed. The surveillance requirements will still ensure that the
reactor trip breakers and the reactor trip bypass breakers are
tested and within the limits. Therefore, the proposed changes will
not impact the margin of safety as designated in the bases of any
technical specification.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
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: May 18, 1994
Description of amendment request: The amendment would change
operability requirements for the Fuel Building Exhaust Filter System to
require it to be operable whenever irradiated fuel is in the spent fuel
pool, which has had less than 60 days of decay time. Surveillance
requirements for the Fuel Building Exhaust Filter System would be
changed to require that the system be tested and verified operable at
no greater than 31 days prior to its required usage.
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 SHC [significant
hazards consideration] because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed modification will revise the period of time during
which the Fuel Building Exhaust Filter System must be operable.
The propose[d] change will require that the system is operable
whenever irradiated fuel, which has decayed less than 60 days, is in
the spent fuel pool. Currently, the system is required to be
operable whenever a load is moved over the pool or fuel is being
moved in the pool.
The modification has no effect on the probability of a fuel
handling accident. The consequences of a fuel handling accident has
been evaluated at two intervals. The first time is the minimum decay
time. At this time (t=100 hours) with irradiated fuel in the pool,
the Fuel Building Exhaust Filter System is required, per the
existing and the proposed Technical Specification, to be operable.
Therefore, the consequences of an accident are identical to that
described in the FSAR [Final Safety Analyses Report]. The second
scenario evaluated is when the filters are initially isolated (t=60
days). The resultant offsite dose, assuming no filtration and lower
core inventory due to decay, are significantly lower than was
calculated at t=100 hours. Therefore, the existing accident analysis
in FSAR Section 15.7.4 is limiting and the proposed modification
will not impact the probability or consequences of an accident.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change does not impact any system or component
which could cause a fuel handling accident. The Fuel Building
Exhaust Filter System is used for accident mitigations. It's failure
cannot, in any way, create the possibility of a new or different
kind of accident.
3. Involve a significant reduction in a margin of safety.
The proposed change to the Fuel Building Exhaust Filter System
has been analyzed at the two most critical times. The first analysis
was done when the fuel is first placed in the pool, and the second
analysis was done when the filtration system is isolated. The first
event resulted in no change in assumptions in the analysis presented
in the FSAR, ergo no change in dose. The second event has been
analyzed and doses have decreased, when compared to the first event.
The system will be verified operable per the performance of
Surveillance Requirement 4.9.12a prior to fuel or load movement over
the pool. Therefore, 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 request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: May 13, 1994
Description of amendment request: This amendment would revise
Technical Specifications Surveillance Requirement 4.8.1.1.2e.8, which
requires that an emergency diesel generator be retested within 5
minutes after completing a 24-hour endurance run.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS change would revise the Emergency Diesel
Generator (EDG) surveillance criteria to allow the hot restart test
to be performed independent of the Engineered Safety Features (ESF)
load sequencing test and the 24 hour endurance run. The proposed
surveillance requirements would continue to demonstrate that the
objectives of each of these tests are met. Specifically, the EDG's
are shown to be capable of starting the ESF loads in the required
sequence, operating at full load for an extended period of time, and
restarting from a full load temperature condition. Therefore, the
proposed changes would not adversely affect the EDG's ability to
support mitigation of the consequences of any previously evaluated
accident. The proposed changes to the surveillance requirements do
not affect the initiation or progression of any accident sequence.
Therefore, the proposed change does not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This proposed TS change does not require physical changes to the
plant or equipment, and does not impact any design or functional
requirements of the Emergency Diesel Generators (EDGs). The proposed
change affects surveillance test criteria such that increased
scheduling flexibility is allowed while the test objectives
associated with demonstrating EDG operability continue to be met.
The proposed changes do not allow any plant configurations that are
presently prohibited by the Technical Specifications.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed TS change does not involve a change to the physical
design or functional requirements of the Emergency Diesel Generators
(EDGs). Surveillance testing in accordance with the proposed
Technical Specification will continue to demonstrate the ability of
the EDG's to perform their intended function of providing electrical
power to ESF systems needed to mitigate design basis transients,
consistent with the plant safety analyses. The margin of safety
demonstrated by the plant safety analyses is therefore not affected
by the proposed change.
Therefore, the proposed TS change does not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Charles L. Miller
Philadelphia Electric Company, Public Service Electric and Gas
Company,Delmarva Power and Light Company, and Atlantic City
Electric Company,Docket No. 50-277, Peach Bottom Atomic Power
Station, Unit No. 2,York County, Pennsylvania
Date of application for amendment: May 13, 1994
Description of amendment request: The proposed amendment would
extend the Type A test (i.e., Containment Integrated Leak Rate Test)
interval on a one-time basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The accidents which are potentially negatively impacted by the
proposed change are any Loss of Coolant Accident (LOCA) inside
primary containment as described in the PBAPS [Peach Bottom Atomic
Power Station], Units 2 and 3 UFSAR [Updated Final Safety Analysis
Report].
The proposed change increases the surveillance interval of the
10 CFR [Part] 50, Appendix J Type A test (i.e., Containment
Integrated Leakage Rate Test (CILRT)) from 42 months to 66 months.
This test is performed to determine that the total leakage from
containment does not exceed the maximum allowable primary
containment leakage rate (i.e., designated La) at a calculated peak
containment internal pressure (Pa), as defined in 10 CFR [Part] 50,
Appendix J. The primary containment limits the leakage of
radioactive material during and following design bases accidents in
order to comply with the offsite dose limits specified in 10 CFR
[Part] 100. Accordingly, the primary containment is not an accident
initiator, it is an accident mitigator. No physical or operational
changes to the containment structure, plant systems, or components
would be made as a result of the proposed change. Therefore, the
probability of occurrence of an accident previously evaluated is not
increased.
The failure effects that are potentially created by the proposed
one-time TS change have been considered. The relevant components
important to safety which are potentially affected are the
containment structure, plant systems, and containment penetrations.
There are no physical or operational changes to any plant equipment
associated with the proposed TS change. Therefore, the probability
or consequences of a malfunction of equipment important to safety is
not increased.
The proposed change introduces the possibility that primary
containment leakage in excess of the allowable value (i.e., La)
would remain undetected during the proposed 24 month extension of
the interval between the second and third Type A test. The types of
mechanisms which could cause degradation of the primary containment
can be categorized into two types. These are: 1) degradation due to
work which is performed as part of a modification or maintenance
activity on a component or system (i.e., activity-based), or; 2)
degradation resulting from a time-based failure mechanism.
A review of activity-based failure mechanisms has determined
that the potential from degradation due to activity based mechanisms
is minimal.
Regarding the potential for primary containment degradation due
to a time-based mechanism, we have concluded that the PBAPS Local
Leak Rate Test (LLRT) program would identify most types of
penetration leakage. The LLRT program involves measurement of
leakage from Type B and Type C primary containment penetrations as
defined in 10 CFR [Part] 50, Appendix J.
The 10 CFR [Part] 50, Appendix J, Type B tests are intended to
detect local leaks and to measure leakage across pressure containing
or leakage-limiting boundaries other than valves, such as
containment penetrations incorporating resilient seals, gaskets,
expansion bellows, flexible seal assemblies, door operating
mechanism penetrations that are part of the containment system,
doors, and hatches. 10 CFR [Part] 50, Appendix J, Type C testing is
intended to measure reactor system primary containment isolation
valve leakage rates. The frequency of the Type B and Type C testing
is not being altered by the proposed TS change. [However, in an
April 18, 1994 letter, the licensee has requested a 60-day extension
of the Type B and Type C testing.] The acceptance criterion for Type
B and Type C leakage is 0.6 La (i.e., 0.3 % wt/day) which, when
compared to the Type A test acceptance criterion of 0.75 La (i.e.,
0.375 % wt/day), is a significant portion of the Type A test
allowable leakage.
The proposed TS change only extends the interval between two
consecutive Type A tests. The Type B and Type C tests will be
performed as required. The Type B and Type C tests will continue to
be used to confirm that the containment isolation valves and
penetrations have not degraded. Containment system components that
would not be tested are the containment structure itself and small
diameter instrumentation lines. Time-based degradation of any of the
instrumentation lines would most likely be identified by faulty
instrument indication or during instrument calibrations that will be
performed during the PBAPS, Unit 2 refueling outage 10. In examining
the potential for a time-based failure mechanism that could cause
significant degradation of the containment structure, we concluded
that the risk, if any, of such a mechanism is small since the design
requirements and fabrication specifications established for the
containment structure are in themselves adequate to ensure
containment leak tight integrity.
Based on the above evaluation, we have concluded that the
proposed TS change will have a negligible impact on the consequences
of any accident previously evaluated. To support this conclusion, a
review of the PBAPS, Unit 2 CILRT history was performed. This review
identified that the only failure mechanism that has been detected
during the past CILRTs is an activity based component failure, and
that there is no indication of any time-based degradation that would
not be identified during performance of Type B and Type C tests.
Although this review concluded that the risk of undetected
primary containment degradation is not increased, the Individual
Plant Examination (IPE) for PBAPS, Units 2 and 3, was also reviewed
in order to assess the impact of exceeding the primary containment
allowable leakage rate, if a non-mechanistic activity type (i.e.,
time-based) failure were to occur. The IPE included an evaluation of
the effect of various containment leakage sizes under different
scenarios. The IPE results showed that a containment leakage rate of
35% wt/day would represent less than a 5% increase in risk to the
public being exposed to radiation. This evaluation was based on a
study performed by Oak Ridge National Laboratory for light water
reactors that evaluated the impact of leakage rates on public risk.
As stated earlier, the current value of La for PBAPS, Unit 2, is
0.5% wt/day, which is significantly less than the 35% wt/day
discussed in the IPE evaluation.
Therefore, the proposed TS change involving a one-time extension
of the Type A test interval and performing the third Type A test
after the second Appendix J 10-year service period will not involve
an increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change is an increase of a surveillance test
interval and does not make any physical or operational changes to
existing plant systems or components. Primary containment acts as an
accident mitigator not initiator. Therefore, the possibility of a
different type of accident than any previously evaluated or the
possibility of a different type of equipment malfunction is not
introduced.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The total primary containment leakage rate ensures that the
total containment leakage volume will not exceed the value assumed
in the safety analyses at the peak accident pressure. As an added
conservatism, the measured overall leakage rate is further limited
to less than or equal to 0.75 La during performance of periodic
tests to account for possible degradation of the containment leakage
barriers between leakage tests. There is the potential that
containment degradation could remain undetected during the proposed
24 month surveillance interval extension and result in the
containment leakage exceeding the allowable value assumed in safety
analysis. A review of activity-based failure mechanisms has
determined that the potential from degradation due to activity based
mechanisms is minimal.
Regarding the potential for primary containment degradation due
to a time-based mechanism, we have concluded that the PBAPS Local
Leak Rate Test (LLRT) program would identify most types of
penetration leakage. The LLRT program involves measurement of
leakage from Type B and Type C primary containment penetrations as
defined in 10 CFR [Part] 50, Appendix J.
The 10 CFR [Part] 50, Appendix J, Type B tests are intended to
detect local leaks and to measure leakage across pressure containing
or leakage-limiting boundaries other than valves, such as
containment penetrations incorporating resilient seals, gaskets,
expansion bellows, flexible seal assemblies, door operating
mechanism penetrations that are part of the containment system,
doors, and hatches. 10 CFR [Part] 50, Appendix J, Type C testing is
intended to measure reactor system primary containment isolation
valve leakage rates. The frequency of the Type B and Type C testing
is not being altered by the proposed TS change.
Finally, a review of the results of previous PBAPS, Unit 2 CILRT
results concluded that the only failure mechanism which has been
detected during the past CILRTs is activity-based and that there is
no indication of time-based failures that would not be identified
during performance of Type B and Type C tests. Therefore, we have
concluded that the proposed extended test interval would not result
in a non-detectable PBAPS, Unit 2 primary containment leakage rate
in excess of the allowable value (i.e., 0.5% wt/day) established by
the TS and 10 CFR [Part] 50, Appendix J.
Therefore, the proposed TS change does not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: 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, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Charles L. Miller
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: March 11, 1994
Description of amendment request: The proposed amendment would
reduce the allowed outage time for the residual heat removal (RHR)
suction relief valves (SRVs) in accordance with the guidance of Generic
Letter (GL) 90-06.
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 probability or consequences of an accident previously
evaluated is not significantly increased.
This change decreases the allowed outage time of a Low
Temperature Overpressure Protection (LTOP) system. There is no
hardware, software, or operating methodology change, so there is no
increase in probability or consequences. Since the time allowed for
one train of this equipment to be inoperable is shorter, the
probability of an overpressure event not being mitigated has also
been reduced. The consequences will not change unless the system or
operation of the system changes.
2. [The proposed change will not] [c]reate the possibility of a
new or different kind of accident from any previously analyzed.
As this proposed change will not involve any changes to
hardware, software, or operating practices, it cannot create any
possibility of new or different kinds of accidents from those
previously analyzed. The RHR SRVs are intended to provide protection
against a rupture of a pressure boundary from an over-pressure
condition which has the potential to result in core uncovery. The
original design basis of the plant complies with the requirements of
10 CFR 50 Appendix G and uses the RHR SRVs to meet the fracture
toughness requirements of 10 CFR 50 Appendix G. This change only
increases the availability of this protection and does not create
any new or different kinds of accidents.
3. [The proposed amendment does not] [i]nvolve a significant
reduction in a margin of safety.
SCE&G already has administrative controls in place to minimize
the possibility of an overpressure event occurring as well as to
assure that there are two trains of LTOP equipment operable during
the modes when the potential exists for this event. There are
controls to preclude the inadvertent start-up of a Reactor Coolant
Pump or Charging Pump and controls to ensure that both RHR Suction
Isolation Valves for each train are open and remain open except for
testing and maintenance. This alignment is maintained until the RHR
System is realigned for its ECCS function. These controls are
proceduralized in plant operating procedures.
This change does not involve a significant reduction in a margin
of safety as nothing is changed which affects the margin in a
negative direction. The decrease in AOT actually increases the
margin since the allowed time for one train to be inoperable has
been 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: Fairfield County Library,
Garden and Washington Streets, Winnsboro, South Carolina 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: William H. Bateman
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 16, 1994 (TS 94-03)
Description of amendment request: The proposed change would remove
Table 3.3-2, ``Reactor Trip System Instrumentation Response Times,''
and Table 3.3-5, ``Engineered Safety Features Response Times,'' from
the technical specifications and incorporate the limits into the
Updated Final Safety Analysis Report. In addition, references to these
tables in Specifications 3.3.1.1, 3.3.2.1, and 4.3.1.1.3 (for Unit 1)
and 3.3.1, 3.3.2, and 4.3.1.1.3 (for Unit 2) would be removed. A
footnote would be added to Specification 4.3.1.1.3 indicating that
neutron detectors are exempt from response time testing. These changes
have been proposed in accordance with Generic Letter 93-08. A change to
the Bases would indicate that the response time limits would be
maintained in the Updated Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not alter the response time limit
requirements for the reactor trip or engineered safety feature
actuation systems or surveillance testing and frequency. Placing
these limits in the Updated Final Safety Analysis Report (UFSAR)
will ensure the plant design basis is maintained in accordance with
10 CFR 50.59. Since no actual changes to response time limits or
surveillance requirements are involved, the probability or
consequences of an accident are not increased.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes does not affect any plant equipment,
functions, or setpoint by relocating response time limits to the
UFSAR. Therefore, the possibility of a new or different kind of
accident is not created.
3. Involve a significant reduction in a margin of safety.
The proposed change will continue to require SQN to maintain the
plant functions at the required setpoints necessary for the design
basis and to support the accident analysis. The margin of safety is
not reduced because there is no change to plant functions and the 10
CFR 50.59 process will continue to ensure the plant design basis is
appropriately maintained.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 18, 1994 (TS 94-05)
Description of amendment request: The proposed change would add a
note to the action statement for Limiting Condition for Operation
3.7.7, ``Control Room Emergency Ventilation System,'' indicating that
the provisions of TS 3.0.3 are not applicable while performing actions
associated with a tornado warning.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The control room emergency ventilation system (CREVS) was
designed to ensure control room habitability during accident
conditions. The design basis of SQN does not include an accident
creating a contaminated air condition concurrent with a tornado. The
ability of the CREVS to perform its design function has not been
affected by this change. The proposed change will not increase the
possibility or consequences of an accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
An accident involving a contaminated air condition and a tornado
have been analyzed as part of the SQN design basis. Both accidents
are assumed to occur independently. This change does not create a
new or different accident not previously analyzed.
3. Involve a significant reduction in a margin of safety.
The design basis of the CREVS is not impacted by this TS change.
There is no change in any assumptions made in the Final Safety
Analysis Report. Therefore, there is no reduction in the margin of
safety as a result of this change.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 19, 1994; superseded May 16, 1994
(TS 93-04)
Description of amendment request: The proposed change would clarify
and consolidate the technical specifications (TS) regarding the dual
function of the containment vacuum relief system (i.e., the vacuum
relief and containment isolation functions). The proposed changes would
revise TS 3/4.6.6, ``Vacuum Relief Valves,'' to indicate the actions
that would be required should one or more vacuum relief (VR) lines be
incapable of performing its containment isolation function or incapable
of performing its VR function. In addition, the testing requirements
would be revised to add specific requirements and reflect the inservice
test (IST) program by relocating the testing requirements from TS
4.6.3.2.d and Table 3.6-2 to the new TS 4.6.6 (and to Sequoyah's IST
program). Other proposed changes affect Bases 3/4.6.6 section and TS
index pages to reflect the proposed changes indicated above. This
proposed change was originally noticed on May 12, 1993 (58 FR 28060),
which is superseded by this notice.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
TVA's proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change does not increase the probability of an accident
since the proposed change does not affect any plant systems,
equipment, or components. The dual design functions of SQN's
containment vacuum relief (VR) system (i.e., provide containment VR
and containment isolation) are not affected. The consequences of an
event are not significantly increased by the proposed increase in
allowed outage time from 4 hours to 72 hours for returning an
inoperable VR system to operable status. The probability of an event
during the relatively short duration of the TS completion times, in
conjunction with the redundancy provided in the design of the
system, provide sufficient assurance that the VR lines are available
for mitigating an accident or abnormal event.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
No physical modification is being made to any plant hardware or
plant operating setpoints, limits, or operating procedures as a
result of this change. TVA's proposed change provides a TS
improvement that clarifies the TS requirements associated with the
dual design function of SQN's VR system. The proposed change removes
the potential for creating a conflict between Specification 3/4.6.3,
``Containment Isolation Valves,'' and Specification 3/4.6.6,
``Vacuum Relief Valves.''
The proposed change does not alter any accident analysis or any
assumptions used to support the accident analyses. The containment
leakage assumptions used to determine offsite dose limits for
compliance with 10 CFR 100 are not affected. The analysis that
supports the containment VR system also remains unchanged. The
proposed 72-hour and 1-hour completion times for returning an
inoperable VR line to operable status are consistent with the NUREG-
1431 and NUREG-1433. Consequently, the proposed change does not
create the possibility of a new or different kind of accident from
any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The margin of safety provided by the design of SQN's containment
VR system remains unchanged. TVA's proposed change does not affect
the VR function or the containment isolation function that currently
exists in SQN TSs. The proposed change eliminates the potential for
conflicting requirements within SQN TSs and ensures that the proper
action is taken to preserve these dual design functions while the
plant is in Modes 1, 2, 3, or 4. TVA's proposed change provides a TS
improvement that combines these functional requirements into a
single specification. Both VR and containment isolation requirements
will continue to be provided. Accordingly, the proposed change does
not involve a reduction in the margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: February 14, 1993
Brief description of amendments: The proposed amendments would
revise the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2,
technical specifications (TS) by (1) changing the allowable value for
Unit 2 overtemperature N-16 and pressurizer pressure-low setpoints, (2)
deleting Equation 2.2-1 from TS 2.2.1, and (3) administrative 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 a previously evaluated
accident.
Overtemperature N-16, Unit 2
Incorporation of the increased temperature uncertainties
reported by Rosemount will change the Allowable Value of the
Overtemperature N-16 trip function. The change does not affect the
Safety Analysis Limits assumed in the accident analysis. Because the
change only impacts the Allowable Value for a setpoint and does not
affect any system designs or operations, the change does not
increase the probability of an accident. Although the Allowable
Value is changed in the conservative direction, the change assures
that, considering the newly identified transmitter uncertainty, the
trip actuates prior to the conditions assumed in the accident
analyses. As such, there is no impact on the consequences of any
accidents previously evaluated.
Pressurizer Pressure - Low, Unit 2
The added uncertainties change the Allowable Value of the Unit 2
Pressurizer Pressure-Low Reactor Trip function. The change does not
affect the Safety Analysis Limits assumed in the accident analysis.
Because the change only impacts the Allowable Value for a setpoint
and does not affect the system design or operations, the change does
not increase the probability of an accident. Although the Allowable
Value is changed in the conservative direction, the change assures
that, considering the newly identified transmitter uncertainty, the
trip actuates prior to the conditions assumed for the accident
analyses. As such, there is no impact on the consequences of any
accidents previously evaluated.
Equation 2.2-1
The changes to Specifications 2.2.1 and 3.3.2, to Tables 2.2-1
and 3.3-3, and to the bases sections will require recalibration of
the channel and removal of any accumulated errors in any function
whose ``as found'' setpoint is found to be less conservative than
its allowable value. These changes delete a potentially less
conservative option and will result in actual channel operation
closer to the nominal setpoint and within the allowable value band.
These changes will in effect validate one of the assumptions made in
the accident analysis and will not increase the probability or
consequences of any accident evaluated in the Safety Analysis
Report.
Administrative Changes
The changes to combine the Unit 1 and Unit 2 line items into a
dual Unit line if the Trip Setpoint and Allowable Value values are
the same is administrative and meant as a human factors improvement
for operator convenience. The change does not affect the operation
of any equipment, the operating point of any equipment, nor any
equipment hardware and thus does not increase the probability or
consequences of any accident evaluated in the Safety Analysis
Report.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously analyzed accident.
Overtemperature N-16 and Pressurizer Pressure - Low, Unit 2
As the proposed amendment changes only the Unit 2 Allowable
Values of the Overtemperature N-16 reactor trip and the Pressurizer
Pressure-Low reactor trip and does not have any physical effect on
the transmitter or circuitry, there are no new or different types of
accident introduced.
Equation 2.2-1
Deletion of this equation and its associated action statements,
definitions and values does not introduce any physical changes to
any systems, structures, or components. The change merely assures
that setpoints which are less conservative than their Allowable
Value are recalibrated prior to being declared operable. These
changes do not introduce any new credible failure modes which may
create the possibility of a new or different accident.
Administrative Changes
Combining line items for Unit 1 and Unit 2 into a dual Unit
entry for administrative purposes does not introduce any new
credible failure modes which may create the possibility of a new or
different accident.
3. The proposed change does not involve a significant reduction
in the margin of safety.
Overtemperature N-16 and Pressurizer Pressure - Low, Unit 2
Incorporation of the added temperature uncertainties of the
Rosemount transmitters assures that the safety analysis limits
assumed in the accident analyses for Overtemperature N-16 and
Pressurizer Pressure-Low reactor trip functions for Unit 2 are met.
There is no change in the acceptance criteria or the results of
these analyses due to this change. Thus there is no effect on the
margin of safety.
Equation 2.2-1
Deletion of Equation 2.2-1, related actions and associated
definitions and values, merely eliminates one option to assure that
the safety analysis assumptions are met. This option is not
presently in use and the accident analyses assumptions have been and
will continue to be met using the other option (to re-calibrate
channels prior to restoring operability). Thus the margin of safety
is unaffected.
Administrative Changes
Combining the Unit 1 and Unit 2 line items of Table 2.2-1 for
RTS [Reactor Trip Systems] functions and of Table 3.3-3 for ESFAS
[Engineered Safety Features Actuation System] functions into dual
unit entries does not change the Trip Setpoint or the Allowable
Value for the functions. The margin of safety is unaffected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
NRC Project Director: William D. Beckner
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: April 19, 1994
Description of amendment request: The proposed amendment revises
Technical Specification 6.2.2.g to reflect a title designation change
within the Wolf Creek Nuclear Operating Corporation (WCNOC)
organization. The title of Supervisor Operations is being changed to
Assistant Manager Operations. The title change does not represent any
change in reporting relationships, job responsibilities, or overall
organizational commitments.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
This change involves an administrative change to the WCNOC
organization and to the position title and as such has no effect on
plant equipment or the technical qualification of plant personnel.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated. This
change is administrative in nature and does not involve any change
to installed plant systems or the overall operating philosophy of
Wolf Creek Generating Station.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change does not involve a significant reduction in
a margin of safety. This change does not involve any changes in
overall organizational commitments. A position title change alone
does not reduce 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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: Theodore R. Quay
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: April 19, 1994
Description of amendment request: The proposed amendment revises
Technical Specification Table 3.6-1, ``Containment Isolation Valves,''
by deleting reference to two (2) valves. The Technical Specification
change reflects a planned modification which removes the essential
service water (ESW) containment air cooler return line isolation valve
bypass valves and associated piping.
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.
After the design modification is completed the ESW Containment
Penetrations will be provided with stainless steel isolation valves,
which will be provided with automatic SIS [safety injection signal]
actuation signals to open automatically to provide required cooling
water flow to the Containment Air Coolers following a LOCA [loss-of-
coolant accident] or MSLB [main steamline break]. Replacement of the
current carbon steel isolation valves with stainless steel valves
and removing the unnecessary bypass lines and bypass isolation
valves will reduce the amount of seat leakage currently experienced
with these valves.
The probability of occurrence of a previously evaluated accident
is not increased because this modification does not introduce any
new potential accident initiating conditions. The consequences of an
accident previously evaluated is not increased because the ability
of containment to restrict the release of any fission product
radioactivity to the environment will not be degraded by this
modification.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed modification will reduce the number of containment
isolation valves and replace several carbon steel isolation valves
with stainless steel valves, which will be less susceptible to
erosion and corrosion. Thus, potential system leakage will be
reduced by this modification, while valve reliability will be
enhanced. The new valves are designed to the original ESW System
requirements, and removal of the bypass lines and bypass isolation
valves will not result in a malfunction of any other plant
equipment. Therefore, this proposed modification will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The removal of the bypass lines and bypass isolation valves will
not adversely affect containment isolation capability for credible
accident scenarios. Due to a previous design change, the bypass
lines are no longer required to ensure adequate cooling flow to the
Containment Air Coolers. In addition, the operability and
reliability of the remaining isolation valves will be enhanced by
replacing the current carbon steel valves with stainless steel
valves.
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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: Theodore R. Quay
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: May 11, 1994
Brief description of amendment request: The amendment would allow
reduced power operation as a function of reactor coolant system (RCS)
total flow rate for flow rate reductions of up to 5 percent below the
currently specified flow rate. Operation will be allowed at total flow
rates slightly lower than (293,540 gpm X (1.0 plus C1)) if rated
thermal power (RTP) is reduced by 1.5 percent for each one percent that
RCS total flow is less than this rate. This change would provide for
needed operational margin and flexibility without the unnecessary
penalty of a large power reduction.
Date of publication of individual notice in Federal Register: May
25, 1994 (59 FR 27079)
Expiration date of individual notice: June 24, 1994
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
NuclearPower Station, Unit 1, New London County, Connecticut
Date of amendment request: May 27, 1994
Description of amendment request: The amendment would add a new
section to Technical Specification Section 6.17 and would require that
procedures be in place to provide for monitoring and sampling of
emergency service water (ESW) discharge flow during accident conditions
when a positive differential pressure cannot be maintained between ESW
and low pressure coolant injection (LPCI) in the LPCI heat exchangers.
Date of publication of individual notice in Federal Register: June
7, 1994 (59 FR 29448)
Expiration date of individual notice: July 7, 1994
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
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 20555, 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: February 18, 1994, as
supplemented by letter dated May 16, 1994
Brief description of amendments: These amendments modify Technical
Specification (TS) Figure 3.2-1, ``REACTOR COOLANT COLD LEG vs CORE
POWER LEVEL,'' of TS 3/4.2.6, ``REACTOR COOLANT COLD LEG TEMPERATURE,''
for Units 1 and 3 to include the cold leg temperature between 552 deg.F
and 562 deg.F at core power levels between 90 percent and 100 percent
within the AREA OF ACCEPTABLE OPERATION. Also, the proposed amendments
modify TS 3/4.1.1.4, ``MINIMUM TEMPERATURE FOR CRITICALITY,'' and BASES
3/4.1.1.4, ``MINIMUM TEMPERATURE FOR CRITICALITY,'' to allow the
minimum temperature for criticality to be established at 545 deg.F,
rather than the current value of 552 deg.F, to establish the
surveillance temperature at 552 deg.F, rather than the current
557 deg.F, and to clarify the BASES for this TS.
Date of issuance: June 7, 1994
Effective date: NPF-41 and NPF-51, prior to startup from the next
refueling outage; NPF-74, no later than 45 days from the date of
issuance.
Amendment Nos.: 77, 63, and 49
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14886) The additional information contained in the May 16, 1994, letter
was clarifying in nature, was within the scope of the initial notice,
and did not affect the NRC staff's proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated June 7, 1994.No
significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County,North Carolina
Date of application for amendments: April 14, 1994, as supplemented
on May 16, 1994.
Brief description of amendments: The amendments change the
Technical Specifications (TS) to relocate the Instrument Response Time
Tables to the Updated Final Safety Analysis Report in accordance with
NRC Generic Letter 93-08.
Date of issuance: May 31, 1994
Effective date: May 31, 1994
Amendment Nos.: 171 and 202
Facility Operating License Nos. DPR-71 and DPR-62. Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: April 26, 1994 (59 FR
21785) The May 16, 1994, letter provided clarifying information that
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 May 31, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
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: August 20, 1993, as
supplemented by letters dated December 27, 1993, March 22, 1994, and
May 31, 1994.
Brief description of amendments: The amendments delete Technical
Specification Section 3/4.6.1.5, ``Primary Containment Structural
Integrity'' which includes Surveillance Requirements for the Primary
Containment Tendons and adds a Technical Specification requirement to
establish, implement, and maintain a comprehensive containment tendon
program. The containment tendon program is based on Regulatory Guide
1.35, Rev. 3, and is titled ``Inservice Inspection Program for Post
Tensioning Tendons.'' The new program will allow the Unit 1 and 2
containments to be tested as twin containments.
Date of issuance: June 3, 1994
Effective date: June 3, 1994
Amendment Nos.: 100 and 84
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59746) The supplemental information submitted December 27, 1993,
March 22, 1994, and May 31, 1994, contained clarifying information
related to the original request, and did not change the no significant
hazards finding. The Commission's related evaluation of the amendments
is contained in a Safety Evaluation dated June 3, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: December 20, 1993
Brief description of amendments: The amendments increase the
minimum critical power ratio (MCPR) from 1.06 to 1.07 for Quad Cities,
Units 1 and 2, as a result of the planned implementation of GE 8x8NB-3
fuel for Cycle 14 of each unit.
Date of issuance: June 10, 1994
Effective date: June 10, 1994
Amendment Nos.: 146 and 142
Facility Operating License Nos. DPR-29 and DPR-30. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 17, 1994 (59
FR 10003) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 10, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of application for amendment: February 25, 1994
Brief description of amendment: The amendment adds a new Technical
Specification 3/4.7.12, ``Ultimate Heat Sink'' and its associated Bases
Section 3/4.7.12.
Date of Issuance: May 31, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 172
Facility Operating License No. DPR-61. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17596) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated May 31, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
PointNuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: December 6, 1993
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) to provide several temporary one-time changes that
are necessary to support the fuel out, chemical decontamination program
that is currently scheduled for the upcoming 1995 refueling outage.
Specifically, the amendment revises the definition of the cold shutdown
condition in TS 1.2.1 by changing the upper limit of Tavg for the
cold shutdown condition from 200 deg.F to 250 deg.F. The amendment also
revises the definition of the hot shutdown condition in TS 1.2.2 by
changing the lower limit of Tavg for the hot shutdown condition
from greater than 200 deg.F to greater than 250 deg.F.
Date of issuance: June 9, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 170
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 16, 1994 (59
FR 7687) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 9, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: January 27, 1994
Brief description of amendments: The amendments would eliminate the
humidity control functions of the containment purge (VP) system
humidistats by deleting the surveillance requirement (SR) for periodic
verification of automatic isolation of the VP system on a high relative
humidity (RH) test signal and heater failure from the existing SR for
Catawba Units 1 and 2.
Date of issuance: May 25, 1994
Effective date: May 25, 1994
Amendment Nos.: 118 and 112
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10005) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 25, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: April 29, 1993, as supplemented
May 16, 1994
Brief description of amendments: The amendments delete License
Condition 2.C.(20) from Facility Operating License NPF-35 for Unit 1,
and License Condition 2.C.(11) from Facility Operating License NPF-52
for Unit 2. These conditions address engine teardown and inspection
required following the crankshaft failure of an Enterprise emergency
diesel generator at the Shoreham Nuclear Plant.
Date of issuance: June 2, 1994
Effective date: June 2, 1994
Amendment Nos.: 119/113
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 26, 1993 (58 FR
30192) The May 16, 1994, letter provided additional information that
did not change the scope of the April 29, 1993, application and
proposed initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 2, 1994.No significant hazards
consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: January 13, 1993, as
supplemented January 28, February 17, and April 26, 1993.
Brief description of amendments: The amendments revise Technical
Specification Table 2.2.1, Sections 3/4.1.2.5, 3/4.1.2.6, 3/4.5.1.1, 3/
4.5.5, and their associated Bases, and Technical Specification 6.9.1.9,
to relocate the values of certain cycle-dependent limits from the
Technical Specifications to the Core Operating Limits Report.
Date of issuance: May 31, 1994
Effective date: May 31, 1994
Amendment Nos.: 143 and 125
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41503) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 31, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
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 application of amendments: March 23, 1994, as supplemented
April 14, May 11, and May 17 (two letters) 1994.
Brief description of amendments: The amendments relating to the
March 23, 1994, application revise Technical Specification (TS) 6.9.2,
``Core Operating Limits Report,'' (COLR) to include a reference to a
Duke Power Company Topical Report describing an analytical method for
determining the core operating limits. Specifically, the amendments
add: ``(4) DPC-NE-1004A, Nuclear Design Methodology Using CASMO-3/
SIMULATE-3P,'' to TS 6.9.2.
The May 11, 1994, letter added a statement to TS 6.9.2 that the
approved methods used to determine the core operating limits given in
TS 6.9.1 are specified in the COLR. The May 11 and 17, 1994, letters
provided information regarding Duke Power's transition from the EPRI-
NODE-P based methodology to the simulate methodology. Revision 1 to the
COLR for Oconee 1 Cycle 16 was submitted by letter dated May 17, 1994.
The April 14, 1994, letter revised the TS Table of Contents to
delete reference to Table 4.4-1. This table was removed from the TS by
an amendment issued on September 16, 1993.
Date of Issuance: June 8, 1994
Effective date: June 8, 1994
Amendment Nos.: 206, 206, and 203
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22007) The April 14, May 11, and May 17 (two letters), 1994, letters
provided additional information that did not change the scope of the
March 23, 1994, application and the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
June 8, 1994. No significant hazards consideration comments received:
No
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 23, 1993
Brief description of amendment: The amendment revised the Technical
Specifications in accordance with Generic Letter 93-05, ``Line Item
Technical Specification Improvements To Reduce Surveillance
Requirements For Testing During Power Operation'' for radiation
monitors, pressurizer heaters, reactor coolant isolation valves, and
auxiliary feedwater pumps.
Date of issuance: June 6, 1994
Effective date: June 6, 1994
Amendment No.: 96
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 16, 1994 (59
FR 7689) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 6, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: December 30, 1993
Brief description of amendments: The proposed change would allow a
one time extension of the allowable outage time for each residual heat
removal (RHR) pump from 3 to 7 days to allow modifications to the RHR
system while the plant is in Mode 1.
Date of issuance: May 31, 1994
Effective date: May 31, 1994
Amendment Nos.: 72 and 51
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10007) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 31, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: November 19, 1993
Brief description of amendments: The amendments modify Technical
Specification Table 3.3-2, Engineered Safety Features Actuation System
Instrumentation, modifying the Mode for which Item 6.e, ``Trip of All
Main Feedwater Pumps, Start Motor-Driven Pumps,'' is required to be
operable.
Date of issuance: June 1, 1994
Effective date: June 1, 1994
Amendment Nos.: 73 and 52
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67847) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June, 1, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: March 1, 1994
Brief description of amendments: The amendments modify Technical
Specification (TS) 3.2.4, ``Quadrant Power Tilt Ratio,'' by adding an
exception to the requirements of TS 3.0.4.
Date of issuance: June 1, 1994
Effective date: June 1, 1994
Amendment Nos.: 74 and 53
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17599) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 1, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: February 7, 1994
Brief description of amendment: The amendment revises the plant
Technical Specifications (TS) to require the Three Mile Island, Unit 1
(TMI-1) annual radioactive effluent release report for the previous
calendar year be submitted by May 1 of each year. The current TS
requires the TMI-1 report be submitted within 60 days after January 1
of each year. Changing the TMI-1 due date to May 1 enables the licensee
to combine the reports for TMI-1 and TMI-2 into a single report with a
common due date.
Date of Issuance: June 10, 1994
Effective date: As of its date of issuance to be implemented within
30 days.
Amendment No.: 185
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17600) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 10, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
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 27, 1993, as supplemented by letter
dated April 18, 1994.
Brief description of amendments: The amendments upgrade the fuel
used in the South Texas Project reactors to Westinghouse VANTAGE 5
Hybrid (V5H) design and implement several analytical and operational
upgrades into the South Texas Project Updated Final Safety Analysis
Report. The amendments modify related setpoints, limiting conditions
for operation, surveillance requirements, design features information,
and associated bases in the following specifications: TS Table 2.2-1,
``Reactor Trip System Instrumentation Trip Setpoints,'' TS Figure 3.1-
1, ``Required Shutdown Margin for Modes 1 and 2,'' TS Figure 3.1-2,
``Required Shutdown Margin for Mode 5,'' TS Figure 3.1-2a, ``MTC versus
Power Level,'' TS 3/4.2.5, ``Power Distribution Limits - DNB
Parameter,'' TS Table 3.3-4, ``Engineered Safety Features Actuation
System Instrumentation Trip Setpoints,'' TS 3/4.6.1.1, ``Primary
Containment - Containment Integrity,'' TS 3/4.6.1.2, ``Containment
Systems - Containment Leakage,'' TS 3/4.6.1.3, ``Containment Systems -
Containment Air Locks,'' TS 3/4.6.1.5, ``Containment Systems - Air
Temperature,'' TS 3/4.7.1.2, ``Plant Systems - Auxiliary Feedwater
System,'' TS 5.2.1, ``Containment - Configuration,'' TS 5.3.1,
``Reactor Core - Fuel Assemblies,'' TS 5.6.1, ``Fuel Storage -
Criticality,'' and adds TS Figure 5.6-7, ``Minimum IFBA Content for In-
Containment Rack Fuel Storage.''
Date of issuance: May 27, 1994
Effective date: May 27, 1994, to be implemented prior to completion
of Unit 1 REO5
Amendment Nos.: Unit 1 - Amendment No. 61; Unit 2 - Amendment No.
50
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36436) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 27, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook,
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of application for amendment: March 9, 1994, as supplemented
April 13, 1994.
Brief description of amendment: The amendment revises the Technical
Specifications to allow a one-time extension for Type B and C leak rate
tests. The Commission had previously granted a one-time schedular
exemption from the requirements in 10 CFR Part 50, Appendix J,
paragraphs III.D.2.(a) and III.D.3. The exemption extends the maximum
allowable time between tests by 150 days.
Date of issuance: June 1, 1994
Effective date: June 1, 1994
Amendment No.: 162
Facility Operating License No. DPR-74. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22009). The April 13, 1994, supplemental letter provided clarifying
information that was within the scope of the April 28, 1994, notice.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated June 1, 1994.No significant hazards
consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns
Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama
Date of application for amendment: January 14, 1992 (TS 300)
Brief description of amendments: The amendments add requirements to
the Browns Ferry Units 1 and 3 Technical Specifications to ensure
thermal-hydraulic stability, consistent with guidance provided by NRC
Bulletin 88-07 ``Power Oscillations in Boiling Water Reactors,'' and
Supplement 1 to that Bulletin.
Date of issuance: May 31, 1994
Effective date: May 31, 1994
Amendment Nos.: 206 and 179
Facility Operating License Nos. DPR-33 and DPR-68:
Date of initial notice in Federal Register: April 15, 1992 (57 FR
13138) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 31, 1994.No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: February 8, 1994 (TS 93-14)
Brief description of amendments: The amendments increase the
pressure setpoint for the motor driven auxiliary feedwater pumps
switchover from the condensate storage tank to the essential raw
cooling water supply.
Date of issuance: May 27,1994
Effective date: May 27, 1994
Amendment Nos.: 183 and 175
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12368) The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated May 27, 1994.No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: July 14, 1993
Brief description of amendment: This amendment revises Sections 3.6
and 4.6 of the Technical Specifications to incorporate reactor coolant
system leakage detection requirements to address Generic Letter 88-01
``NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in
BWR Austenitic Stainless Steel Piping.''
Date of issuance: June 1, 1994
Effective date: June 1, 1994
Amendment No.: 139
Facility Operating License No. DPR-28. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12370) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 1, 1994 No significant
hazards consideration comments received: No
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301.
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 20555,
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 July 22, 1994, 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 20555 and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. 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 20555, 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).
Commonwealth Edison Company, Docket No. STN 50-456, Braidwood
Station, Unit No. 1, Will County, Illinois
Date of application for amendments: April 25, 1994, as supplemented
April 28, 1994, April 30, 1994, May 2, 1994, May 4, 1994, and May 6,
1994.
Brief description of amendments: The amendment revises Braidwood,
Unit 1, technical specifications (TSs) in Appendix A to the operating
license by adding additional surveillance and operating requirements to
Section 4.4.5.2, ``Steam Generator Tube Sample Selection and
Inspection; Section 4.4.5.4, ``Acceptance
Criteria; Section 4.4.5.5, ``Reports; and Section
3.4.6.2. This amendment is applicable only for 100 calendar days from
the date of issuance, not counting any time when the Thot
temperature is below 500 deg.F. These changes revise the existing steam
generator tube repair criteria to allow usage of the voltage-based
criteria identified by the staff in draft NUREG-1477 as the interim
plugging criteria (IPC). Additionally, a footnote is added to TS 3.4.8
to limit the dose equivalent iodine-131 concentration to 0.35
microcuries per gram of coolant for the limited time period cited
above. The Unit 1 Bases are revised to be consistent with the changes
cited above.
Date of issuance: May 7, 1994
Effective date: May 7, 1994
Amendment No.: 50
Facility Operating License No. NPF-72. The amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes. The NRC published a public
notice of the proposed amendment, issued a proposed finding of no
significant hazards consideration and requested that any comments on
the proposed no significant hazards consideration be provided to the
staff by the close of business on May 5, 1994. The notice was published
in the Herald News and the Morris Daily Herald on May 3, 1994. The
Commission's related evaluation of the amendment, finding of emergency
circumstances, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated May 7, 1994.
Attorney for the licensee: Michael I. Miller, Esquire, Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
Local Public Document Room location: Wilmington Township Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
NRC Project Director: James E. Dyer
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: May 19, 1994
Brief description of amendment: The amendment revises the
surveillance requirements in TS 3.3.9.3 and 3.3.10.3, to change the
neutron power limits i.e., 105 neutron counts per second (cps) and
1E-6 amperes (amps) indications on the source and intermediate range
instruments, respectively, for verifying overlap between them.
Date of issuance: May 27, 1994
Effective date: May 27, 1994
Amendment No.: 150
Facility Operating License No. DPR-72. Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No. The Commission's related
evaluation of the amendment and the final determination of no
significant hazards consideration comments are contained in a Safety
Evaluation dated May 27, 1994.
Attorney for the Licensee: Harold F. Reis, Esquire, Newman and
Holtzer, P.C., 1615 L Street, NW., Washington DC 20036
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
NRC Project Director: Herbert N. Berkow
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
MillstoneNuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: May 27, 1994, as supplemented
June 1, 1994.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) by adding a footnote to Tables 3.3-3, 3.3-4 and
3.3-5 of the Millstone Unit No. 2 TS denoting that the operability of
the automatic initiation logic for the auxiliary feedwater system will
rely on operator action for the remainder of Cycle 12.
Date of issuance: June 7, 1994
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 176
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: No. The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated June 7, 1994.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz
Dated at Rockville, Maryland, this 15th day of June 1994.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/IIOffice of Nuclear Reactor
Regulation
[Doc. 94-15025 Filed 6-21-94 8:45 am]
BILLING CODE 7590-01F
