[Federal Register Volume 62, Number 107 (Wednesday, June 4, 1997)]
[Notices]
[Pages 30629-30652]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10604]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 12, 1997, through May 22, 1997. The last
biweekly notice was published on May 21, 1997 (62 FR 27792).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By July 7, 1997, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first
[[Page 30630]]
prehearing conference scheduled in the proceeding, but such an amended
petition must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of amendment request: April 22, 1997 (supersedes October 15,
1996, request)
Description of amendment request: The proposed amendment would
revise the Big Rock Point Technical Specifications to correct several
administrative and editorial inconsistencies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes are clarifications within the Technical
Specifications, and do not alter the technical content of the
technical specifications. Plant operation or configuration is not
affected. The postulated doses received by the general public and
plant personnel as a direct result of accidents previously
described, are not affected. Plant operation or configuration is not
affected. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes are either clarifications to correct
inconsistencies within the Technical Specifications, or corrections
of typographical errors. The proposed changes do not alter the
facility in any way, therefore the proposed changes do not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change[s] [do] not affect any margin of safety as
defined by the Plant Technical Specifications.
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: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: John N. Hannon
Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of amendment request: April 30, 1997
Description of amendment request: The proposed amendment would
alter the company name in the Facility Operating License DPR-6 and
Technical Specifications for the Big Rock Point Plant. Specifically,
the proposed amendment would revise the company name from ``Consumers
Power Company'' to ``Consumers Energy Company.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
[[Page 30631]]
The proposed changes alter the company name in the Facility
Operating License and Technical Specifications to reflect the change
from ``Consumers Power Company'' to ``Consumers Energy Company''.
The company will continue to own all of the same assets, will
continue to serve the same customers, and will continue to honor all
existing obligations and commitments.
Since the proposed changes do not alter the technical content of
any Facility Operating License or Technical Specifications
requirements, they do not alter the design, function, or operation
of any plant structure, system, or component.
Therefore, the changes will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes alter the company name in the Facility
Operating License and Technical Specifications to reflect the change
from ``Consumers Power Company'' to ``Consumers Energy Company''.
The company will continue to own all of the same assets, will
continue to serve the same customers, and will continue to honor all
existing obligations and commitments.
Since the proposed changes do not alter the technical content of
any Facility Operating License or Technical Specifications
requirements, they do not alter the design, function, or operation
of any plant structure, system, or component.
Therefore, the changes will not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
Since the proposed changes do not alter the technical content of
any Facility Operating License or Technical Specifications
requirements, they do not alter the design, function, or operation
of any plant structure, system, or component.
Therefore, the changes will not involve a reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: John N. Hannon
Duke Power Company, Docket No. 50-413, Catawba Nuclear Station,
Unit 1, York County, South Carolina
Date of amendment request: May 8, 1997
Description of amendment request: The proposed amendment would add
a phrase to the footnote to Section 3.4.1.2 of the Technical
Specifications that would permit all reactor coolant pumps (RCPs) to be
deenergized for up to 4 hours during Mode 3 on a one-time basis.
Currently, the RCPs are permitted to be deenergized for up to 1 hour
during Mode 3. The proposed change would allow the licensee to perform
a natural circulation test using the new steam generators (installed in
late 1996).
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 activity does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed natural circulation test would be performed in Mode
3 with the reactor subcritical. This transient is bounded by the
transient analyzed in UFSAR [Updated Final Safety Analysis Report]
Section 15.2.6, Loss of Non-Emergency AC Power to the Station
Auxiliaries. For this ANS [American Nuclear Society] Condition II
event, the reactor is assumed to be operating at 102% power, the
turbine driven auxiliary feedwater pump is assumed unavailable and
each steam generator is assumed to have 18% of the steam generator
tubes plugged. By contrast, the planned natural circulation test
would be performed with the reactor subcritical, less than 0.1% of
the tubes plugged in each steam generator, and all support systems
such as auxiliary feedwater, operable for the test. Therefore, the
proposed natural circulation test would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2) The activity does not create the possibility of a new or
different type of accident from any accident previously evaluated.
The proposed change does not involve a physical alteration of
the unit (i.e., no new or different equipment will be installed),
nor will the function of equipment be changed. The change will allow
for a one time performance of a natural circulation test in Mode 3
which will provide useful data on the natural circulation
capabilities of the new Babcock and Wilcox International (BWI) steam
generators that were recently installed at Catawba Unit 1. The test
data will be utilized to validate analysis and simulator models.
Plant operators will also receive valuable experience from
performance of the test. The test will be conducted using written
and approved procedures. An Emergency procedure (EP/1/A/5000/ECA-
0.1) is also available to the Operators for this test. This test is
bounded by the Loss of Non-Emergency AC Power to the Station
Auxiliaries event in Section 15.2.6 of the Catawba UFSAR. For these
reasons, the planned natural circulation test will not create the
possibility of a new or different type of accident from any
previously evaluated.
3) The activity does not involve a significant reduction in the
margin of safety.
Margin of safety is associated with confidence in the ability of
the fission product barriers (the fuel and fuel cladding, the
Reactor Coolant System pressure boundary, and the containment) to
limit the level of radiation doses to the public. As demonstrated by
the bounding UFSAR analysis in Section 15.2.6, none of the fission
product barriers are adversely impacted by the proposed one-time
change. The proposed change does not alter the manner in which
safety limits, limiting safety system setpoints, or limiting
conditions for operation are determined. For these reasons, the
activity does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
proposed amendments involve no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: February 24, 1997, as supplemented on
April 24, 1997.
Description of amendment request: The licensee proposed changes to
Technical Specification Section 6.9.1.7, Core Operating Limits Report,
to reflect use of the Westinghouse Best Estimate Large Break Loss-of-
Coolant Accident (LOCA) methodology for large break LOCA analysis,
including supporting documents.
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:
Question 1 Does the proposed license amendment involve a
significant increase in the probability or consequences of an
accident previously evaluated?
The plant conditions assumed in the analysis are bounded by the
design conditions for all equipment in the plant. Therefore, there
will be no increase in the probability of a Loss of Coolant Accident
[[Page 30632]]
(LOCA). The consequences of a LOCA are not being increased. That is,
it is shown that the emergency core cooling system is designed so
that its calculated cooling performance conforms to the criteria
contained in 10 CFR 50.46 paragraph (b). No other accident is
potentially affected by this change. Therefore, neither the BiWeekly
probability nor the consequences of an accident previously evaluated
is increased due to the proposed change.
Question 2 Does the proposed license amendment create the
possibility of a new or different kind of accident from any accident
previously evaluated?
No new modes of plant operation are being introduced. The
parameters assumed in the analysis are within the design limits of
existing plant equipment. All plant systems will perform as designed
in response to a potential accident. Therefore, the proposed license
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
Question 3 Does the proposed amendment involve a significant
reduction in the margin of safety?
The analysis in support of the proposed license amendment
realistically models the expected response of the Turkey Point Units
3 & 4 nuclear core during a postulated LOCA. Uncertainties have been
accounted for as required by 10 CFR 50.46. A sufficient number of
loss of coolant accidents with different break sizes, different
break locations and other variations in properties have been
calculated to provide assurance that the most severe postulated loss
of coolant accidents were analyzed. It has been shown by the
analysis that there is a high level of probability the criteria
contained in 10 CFR 50.46 paragraph (b) would not be exceeded.
Therefore, the proposed amendment does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420
NRC Project Director: Frederick J. Hebdon
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: February 17, 1997 as revised May 1,
1997.
Description of amendment request: The proposed amendment would
change the Crystal River Unit 3 (CR-3) Technical Specifications (TS) to
implement 10 CFR Part 50, Appendix J, ``Primary Reactor Containment
Leakage Testing for Water-Cooled Reactors,'' Option B. This option
allows to change from prescriptive testing requirements to performance-
based testing requirements based on the leakage rate testing history of
the containment and components. The proposed TS changes include
revision to TS 3.6.1, 3.6.3, and addition of ``Containment Leakage Rate
Testing Program'' to TS 5.0. The licensee did not propose any
deviations from methods approved by the Commission and endorsed in the
applicable regulatory guide. This notice supersedes the previous notice
dated February 28, 1997 (62 FR 9214)
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:
Criterion 1
Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The TS amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes to the TS are to implement Option B of 10
CFR 50, Appendix J, at CR-3. The proposed changes will result in
increased intervals between containment leakage tests based on the
leakage rate testing history. The proposed changes do not involve a
change to the plant design or operation and does not change the
testing methodology.
NUREG-1493, ``Performance-Based Containment Leak-Test Program,''
provides the technical basis of 10 CFR 50, Appendix J, Option B.
NUREG-1493 contains a detailed evaluation of the expected leakage
from containment and the associated consequences. The increased risk
due to increasing the intervals between containment leakage tests
was also evaluated. The NUREG used a statistical approach to
determine that the increase in the expected dose to the public due
to decreasing the testing frequency is extremely low. NUREG-1493
also concluded that a small increase is justifiable in comparison to
the benefits from decreasing the testing frequency. The primary
benefit is in the reduction in occupational radiation exposure.
Criterion 2
Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The TS amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed TS amendment incorporates the performance-based
testing approach authorized by 10 CFR 50 Appendix, J, Option B.
Decreasing the testing frequency allowed by this change does not
involve a change to plant design or operation. Safety related
equipment and safety functions are not altered as a result of this
change. Decreasing the testing frequency does not affect testing
methodology. As a result, the proposed change does not affect any of
the parameters or conditions that could contribute to the initiation
of any accidents.
Criterion 3
Does not involve a significant reduction in the margin of
safety.
This TS amendment does not involve a significant reduction in
the margin of safety.
The proposed TS amendment does not change the methodology of the
containment leakage rate testing program or program acceptance
criteria. The proposed TS change does affect the frequency of
containment leakage rate testing. With an increased interval between
tests, a small possibility exists that an increase in leakage could
go undetected for a longer period of time. Based on the operational
experience at CR-3, it has been demonstrated that the leak-tightness
of the containment building has consistently been significantly
below the allowable leakage limit. Adequate controls are in place to
ensure that required maintenance and modifications are performed.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Attorney for licensee: R. Alexander Glenn, Corporate Counsel,
Florida Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg,
Florida 33733-4042
NRC Project Director: Frederick J. Hebdon
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: March 27, 1997, as supplemented April 3,
and May 1, 1997.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) for the Crystal River Nuclear
Plant Unit 3 (CR3) relating to the Once Through Steam Generator's
(OTSG's) tube inspection acceptance criteria. Specifically, the
licensee proposed to:
(1) revise TS 3.4.12 (d) to specify 150 gallons per day limit on
primary-to-secondary leakage through either OTSG;
(2) add TS 5.6.2.10.2 e. to define inspection requirements and
disposition criteria for applicable tubes in the ``B'' OTSG first span;
[[Page 30633]]
(3) revise TS 5.6.2.10.4.a.7 to define ``pit-like Intergranular
attack indications
(4) revise TS 5.6.2.10 and 5.7.2 to delete requirements that were
specific to the interim tube plugging criteria applicable until Refuel
11.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
FPC Response:
No. The CR-3 components addressed by this proposed change are
the Once Through Steam Generators (OTSGs), identified by plant
tagging procedures as RCSG-1A and RCSG-1B. The OTSGs are straight
tube, straight shell heat exchangers which allow for heat removal
and the subsequent production of steam as a result of heat transfer
from the primary side reactor coolant to the secondary side
feedwater. Proposed changes are; retaining reduced primary-to-
secondary leak rates approved previously for one cycle only,
returning inspection result reporting requirements to those
previously implemented, and establishing new inspection requirements
for the ``B'' OTSG. Sunset clauses are being removed from pages
containing requirements effective for one refueling outage and
subsequent operating cycle only.
Based on review of Chapter 14 of the CR-3 Final Safety Analysis
Report (FSAR), FPC performed analyses to assess the consequences of
a steam generator tube rupture event, including the complete
severing of a steam generator tube. The analyses concluded that CR-3
was sufficiently designed to ensure that, in the event of a steam
generator tube rupture, the radiological doses would not exceed the
allowable limits prescribed by 10 CFR 100, and would not result in
additional tube failures and further degradation of the reactor
coolant pressure boundary.
Retaining the present primary-to-secondary leakage limit (LCO
3.4.12, RCS Operational Leakage) that was previously approved for
the current operating cycle will continue to provide assurance that
should a significant leak occur, it would be detected and the plant
will be shut down in a timely manner to reduce the likelihood of a
potential tube rupture. This value of primary-to-secondary leakage
applicable to both OTSGs is conservative relative to existing safety
analyses and would result in lower doses than currently calculated
and found acceptable. Removing reporting requirements specific to
use of alternate flaw sizing criteria approved for Refueling Outage
10 only, and returning to previous reporting requirements applicable
to both OTSGs, has no effect on operating plant safety. These
requirements are administrative only and do not affect steam
generator inspection or disposition of inspection results.
The proposed change to the ``B'' OTSG inspection criteria
establishes that future inspections will include 100% inspection of
the first span of specific tubes which are known to have indications
of degradation. The degradation of these tubes is attributed to a
common non-random mechanism.
The results of inspections of these tubes will be dispositioned
using the same criteria as all other OTSG tubes for determination of
the need for plugging or sleeving. Therefore, the proposed change
will not increase the probability or consequence of an accident
previously evaluated as all tubes degraded beyond acceptable limits
will be subject to consistent corrective actions.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
FPC Response:
No. The purpose of OTSG tube inspection is to identify tubes
that may have a higher potential for failure due to degradation that
results in a reduced ability to withstand operating conditions.
Neither the type of inspection of OTSG tubes nor the process for
performing inspections will be changed by this amendment. Consistent
criteria will be applied to disposition inspection results and
consistent corrective actions will be taken for tubes that exceed
this criteria. Retaining the lower leakage limit is conservative
relative to existing analyses. Changes to revise requirements for
reporting inspection results, and remove ``sunset'' clauses
addressing the applicability of License Amendment 154 until
Refueling Outage 11 only, do not alter the design or operation of
the OTSGs. Therefore, no new or different kind of accident will be
created as a result of these changes.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in margin of safety?
FPC Response:
No. The analyses that have been performed on the effects of OTSG
tube failures, as reported in the CR-3 FSAR, have demonstrated that
internal and offsite consequences are within allowable limits. This
change will not alter the acceptance criteria for inspection
results. Since this change will assure that a group of tubes with
existing first span pit-like inter-granular attack indications are
inspected each inspection period, the likelihood of detecting active
degradation, as well as the probability of repairing degraded tubes
prior to the degradation resulting in a through-wall opening or tube
rupture, is increased. Retaining the currently accepted primary-to-
secondary leakage limit continues to provide assurance that should a
significant leak occur, it would be detected and the plant will be
shut down in a timely manner to reduce the likelihood of a potential
tube rupture, thereby maintaining or improving the existing margin
of safety. Changes to revise requirements for reporting inspection
results, and remove ``sunset'' clauses addressing the applicability
of License Amendment 154 until Refueling Outage 11 only, do not
alter the design or operation of the OTSGs. Therefore, these changes
will not involve a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Attorney for licensee: R. Alexander Glenn, Corporate Counsel,
Florida Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg,
Florida 33733-4042
NRC Project Director: Frederick J. Hebdon
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: May 8, 1997
Description of amendment request: The proposed amendment
incorporates additional analytical methods, GPU Nuclear Topical
Reports, TR-078, TR-087, TR-091, and TR-092P, previously approved by
the NRC, to Technical Specifications (TS) Section 6.9.5.2. These
Topical Reports will be utilized by GPU Nuclear to perform core reload
design analysis for the Three Mile Island, Unit 1 (TMI-1) Facility. TS
6.9.5.2 is also being editorially revised to relocate the existing note
that the current revision level shall be specified in the Core
Operating Limits Report (COLR) such that it applies to the additional
Topical Reports, as well as BAW-10179 P-A.
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 (SHC), which is presented below:
GPU Nuclear has determined that this Technical Specification
Change Request poses no significant hazards as defined by 10 CFR
50.92.
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed change to reference the
analytical methodologies specified in GPU Nuclear Topical Reports
TR-078, TR-087, TR-091,and TR-092 use[d] in TMI-1 core reload design
analysis is considered administrative since these Topical Reports
[[Page 30634]]
have been reviewed and approved by the NRC for use at TMI-1.
Therefore, the proposed change does not involve a significant
increase in the probability of occurrence or the consequences of an
accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. The proposed change
to reference NRC-approved GPU Nuclear Topical Reports TR-078, TR-
087, TR-091, and TR-092P will continue to ensure that approved
methods and criteria are used to establish core operating limits.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed change to reference NRC-approved GPU Nuclear
Topical Reports TR-078, TR-087, TR-091, and TR-092P maintains
existing margins of safety since approved methods and criteria are
still used to establish core operating limits.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Patrick D. Milano, Acting
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 8, 1997
Description of amendment request: The proposed amendment would
modify the minimum accuracy stated in Technical Specification (TS)
Table 3.3-8, Meteorological Monitoring Instrumentation,''
for the instruments used to measure wind speed and air temperature -
delta T. TS Bases Section 3/4.3.3.4 would also be modified to reflect
the proposed changes to TS Table 3.3-8.
Regulatory Guide 1.23 (Safety Guide 23), ``Onsite Meteorological
Programs,'' dated March 17, 1972, provides recommended instrument
accuracies for meteorological instrumentation. The proposed minimum
instrument accuracies for the air temperature - delta T and the wind
speed (only when the wind speed is greater than 5 miles per hour) do
not meet the recommended accuracies of Regulatory Guide 1.23. However,
margin is included to account for uncertainties.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes modify the accuracy requirements for the
instruments which are used to measure wind speed and air temperature
- delta T. The data obtained from the meteorological instrumentation
would be used to: a) estimate the public dose following routine or
accidental releases of airborne radioactivity, b) make decisions
regarding actions to protect the public in the event of an accident
involving a release of airborne radioactivity, and c) establish
radiological dispersion parameters to determine radiological doses
in design basis accident calculations.
The proposed minimum instrument accuracy requirements are more
than sufficient to meet the purposes denoted above. The
meteorological parameters measurement uncertainties insignificantly
affect the results when compared to the accuracies of the source
term estimates, meteorological dispersion models, dose models, and
meteorological forecasting. Therefore, there is no impact on the
consequences (offsite doses) associated with previously evaluated
accidents.
The proposed changes do not alter the way any structure, system,
or component functions, do not alter the manner in which the plant
is operated, and do not have any impact on the protective boundaries
and safety limits for the protective boundaries. Therefore, the
proposed changes do not impact the probability of any previously
evaluated accidents.
Thus, the license amendment request does not impact the
probability of an accident previously evaluated nor does it involve
a significant increase in the consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes modify the accuracy requirements for the
instruments which are used to measure wind speed and air temperature
- delta T. The data provided by these instruments assist in
responding to a design basis accident which may involve a release of
airborne radioactivity. The instruments are used for post accident
monitoring and serve a passive role; they cannot initiate or
mitigate any accident.
The proposed changes do not alter the way any structure, system,
or component functions and do not alter the manner in which the
plant is operated. They do not introduce any new failure modes.
Thus, the license amendment request 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.
As discussed above, the proposed changes modify the accuracy
requirements for the instruments which are used to measure wind
speed and air temperature - delta T which could impact the
radiological dispersion coefficient used to determine radiological
doses in design basis accident calculations. However, the
differences in the instrument accuracies and the Regulatory Guide
1.23 requirements have been determined not to significantly affect
the dispersion coefficients. Thus, there is no significant impact on
offsite doses associated with previously analyzed accidents.
Therefore, there is no significant reduction in the margin of safety
for the design basis accident analysis.
Thus, the license amendment request does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: April 14, 1997
Description of amendment request: Technical Specification 3.4.9.3.a
requires two relief valves be operable to protect the reactor coolant
system from overpressurization when any reactor coolant system cold leg
is less than 350F. The proposed amendment revises the setpoint of the
residual heat removal suction relief valves.
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
[[Page 30635]]
consideration, which is presented below:
NNECO has reviewed the proposed change in accordance with 10CFR
50.92 and has concluded that the change does not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10CFR 50.92(c) are not
satisfied. The proposed change does not involve a SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to Technical Specification 3.4.9.3 to
decrease the setpoint of the residual heat removal suction relief
valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus]
3% ([greater than or equal to] 426.8 psig and [less than or equal
to] 453.2 psig) is consistent with the design capabilities and
system requirements of the relief valves and the relief valves are
not credited in previously evaluated accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change to Technical Specification 3.4.9.3 to
decrease the setpoint of the residual heat removal suction relief
valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus]
3% ([greater than or equal to] 426.8 psig and [less than or equal
to] 453.2 psig) does not change the operation of the residual heat
removal system, reactor coolant system or any system component
during normal or accident evaluations. The proposed change to the
setpoint of the residual heat removal suction relief valves from 450
psig [plus or minus] 3% to 440 psig [plus or minus] 3% ([greater
than or equal to] 426.8 psig and [less than or equal to] 453.2 psig)
also ensures protection of the reactor coolant system against cold
overpressurization transients in accordance with the requirements of
10CFR50, Appendix G.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to Technical Specification 3.4.9.3 to
decrease the setpoint of the residual heat removal suction relief
valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus]
3% ([greater than or equal to] 426.8 psig and [less than or equal
to] 453.2 psig) provides an acceptable allowance between the maximum
relief valve setpoint ([less than or equal to] 453.2 psig) and
10CFR50, Appendix G requirements. The proposed change to the
setpoint provides sufficient allowance between the minimum relief
valve setpoint ([greater than or equal to] 426.8 psig) and reactor
coolant system pressure when residual heat removal system is
unisolated from the reactor coolant system to minimize the
probability of an inadvertent residual heat removal system relief
valve opening.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed change does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: April 28, 1997
Description of amendment request: Technical Specification
Surveillances 4.1.2.3.1, 4.1.2.4.1, 4.5.2.f, and 4.5.2.h require the
charging and safety injection pumps to be tested on a periodic basis
and after modifications that alter subsystem flow characteristics. The
proposed amendment would increase the required differential pressure at
recirculation flow for the safety injection and centrifugal charging
pumps; decrease the required individual safety injection and
centrifugal charging pump injection line flow rate; increase the
allowed individual safety injection pump total flow rate; and make
editorial 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:
NNECO has reviewed the proposed changes in accordance with
10CFR50.92 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
satisfied. The proposed changes do not involve [an] SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to Technical Specification Surveillances
4.1.2.3.1, 4.1.2.4.1, and 4.5.2.f to increase the required discharge
pressure for the centrifugal charging pumps on recirculation flow
during surveillance testing from [greater than or equal to] 2411
psid to [greater than or equal to] 5676 ft (2464 psid) are
consistent with centrifugal charging pump design requirements. The
change in the referenced units from differential pressure measured
in psid to total head measured in feet for the centrifugal charging
pumps and safety injection pumps during surveillance testing is an
administrative change.
The proposed changes to Technical Specification Surveillance
4.5.2.f to increase the required discharge pressure for the safety
injection pumps on recirculation flow during surveillance testing
from [greater than or equal to] 1348 psid to [greater than or equal
to] 3240 ft (1406 psid) are consistent with safety injection pump
design requirements.
The proposed changes to Surveillance 4.5.2.h: to decrease the
required individual centrifugal charging pump injection line flow
rate sum from [greater than or equal to] 339 gpm to [greater than or
equal to] 310.5 gpm, decrease the required individual safety
injection pump injection line flow rate sum from [greater than or
equal to] 442.5 gpm to [greater than or equal to] 423.4 gpm,
increase the required individual safety injection Pump A total flow
rate from [less than or equal to] 670 gpm to [less than or equal to]
675 gpm, and increase the required individual safety injection Pump
B total flow rate from [less than or equal to]
650 gpm to [less than or equal to] 675 gpm are consistent with
centrifugal charging pump and safety injection pump design
requirements.
The proposed changes are consistent with equipment design
requirements and performing surveillance testing does not involve a
significant increase in the probability of an accident previously
evaluated.
The proposed changes to the surveillance testing of the
centrifugal charging pumps and safety injection pumps provide the
necessary assurance that the pumps will function consistent with the
flows used in the accident analyses and does not involve a
significant increase in the consequence of an accident previously
evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to the surveillance testing of the
centrifugal charging pumps and safety injection pumps do not change
the operation of the centrifugal charging or safety injection
systems or any of its components during normal or accident
evaluations. The increase in the allowed maximum safety injection
pump flow does not impact the cold overpressure accident analysis.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
[[Page 30636]]
3. Involve a significant reduction in a margin of safety.
The proposed changes to Technical Specification Surveillances
4.1.2.3.1, 4.1.2.4.1 and 4.5.2.f to increase the required discharge
pressure for the centrifugal charging pumps on recirculation flow
during surveillance testing from [greater than or equal to] 2411
psid to [greater than or equal to] 5676 ft (2464 psid) provides an
acceptable margin between the required surveillance and design pump
performance to provide assurance that the pumps will operate
consistent with the assumptions of the accident analysis.
The proposed changes to Technical Specification Surveillance
4.5.2.f to increase the required discharge pressure for the safety
injection pumps on recirculation flow during surveillance testing
from [greater than or equal to] 1348 psid to [greater than or equal
to] 3240 ft (1406 psid) provides an acceptable margin between the
required surveillance and design pump performance to provide
assurance that the safety injection pumps will operate consistent
with the assumptions of the accident analysis.
The proposed changes to Surveillance 4.5.2.h to decrease the
required individual centrifugal charging pump injection line flow
rate sum from [greater than or equal to] 339 gpm to [greater than or
equal to] 310.5 gpm, decrease the required individual safety
injection pump injection line flow rate sum from [greater than or
equal to] 442.5 gpm to [greater than or equal to] 423.4 gpm,
increase the required individual safety injection Pump A total flow
rate from [less than or equal to] 670 gpm to [less than or equal to]
675 gpm and increase the required individual safety injection Pump B
total flow rate from [less than or equal to] 650 gpm to [less than
or equal to] 675 gpm are consistent with the assumptions of the
accident analysis. The maximum allowed safety injection flow is
consistent with the vendor recommendation for maximum continuous
runout flow. Also, the safety injection
pumps are disabled during specific normal operating modes,
consistent with the assumptions of the accident analysis, to ensure
that they can not be an injection source when the cold overpressure
system is required to be operable and thus the increase in maximum
safety injection pump flow does not affect the cold overpressure
accident analysis.
The change in the referenced units in Technical Specification
Surveillances 4.1.2.3.1, 4.1.2.4.1 and 4.5.2.f from differential
pressure measured in psid to total head measured in feet for the
centrifugal charging pumps and safety injection pumps during
surveillance testing is an administrative change.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed changes do not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: April 28, 1997
Description of amendment request: Technical Specification 3.7.6
requires that flood protection be provided for the service water pump
cubicles and components when the water level exceeds a specific value.
The proposed amendment (1) adds the closing of the service water pump
cubicle sump drain valves, (2) revises the wording of the action
statement to be consistent with the limiting condition for operation,
and (3) revises the associated Bases section.
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 change in accordance with
10CFR50.92 and has concluded that the change does not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed change does not involve [an] SHC because the
change would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to Technical Specification 3.7.6 identify
additional manual actions to be performed to provide external lood
protection for the service water pump cubicles in the event of high
water level (13 ft MSL) [mean sea level]. The cubicle sump drain
valves which are to be closed are part of a modification which
installed a drain line from the sump of each cubicle to the intake
bay in order to provide a passive means of removing internal leakage
from the cubicle. The cubicle sump drain valves are normally
maintained in the open position.
The drain valves meet the intent of RG [Regulatory Guide] 1.59
for ``hardened protection'' and RG 1.102 for ``incorporated
barriers'' in a manner similar to that of the cubicle watertight
doors. RG 1.59 states that hardened protection ``must be passive and
in place, as it is to be used for flood protection, during normal
plant operation''. RG 1.102 states that ``the plant should be
designed and operated to keep doors necessary for flood protection
closed during normal operation''. The Response to FSAR [Final Safety
Analysis Report] Question No. 240.9 established the acceptability of
the practice of maintaining one service water pump cubicle
watertight door open and the other door closed during normal
operations.
The proposed change in the action statement to initiate action
when water level is exceeding 13 feet MSL rather than at 13 feet MSL
is a clarification only which provides consistency between the
limiting condition for operation and the action statements.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to Technical Specification 3.7.6 identify
additional, simple to perform manual actions to provide external
flood protection for the service water pump cubicles.
The proposed change in the action statement to initiate action
when water level is exceeding 13 feet MSL rather than at 13 feet MSL
and the proposed changes to the bases are considered clarifications.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to Technical Specification 3.7.6 identify
additional, simple to perform manual actions to provide external
flood protection for the service water pump cubicles in the
event of high water level (13 ft MSL). The plant modification which
made these additional actions necessary was made to provide for
improved internal flood protection.
The proposed change in the action statement to initiate action
when water level is exceeding 13 feet MSL rather than at 13 feet MSL
and the proposed changes to the bases are considered clarifications.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed change does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
[[Page 30637]]
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: May 1, 1997
Description of amendment request: Technical Specifications 3/
4.8.2.2 and 3/4.8.3.2 specify which electrical power systems are
required to be operable in Modes 5 and 6. The proposed amendment would
clarify the requirements by identifying the specific equipment required
and their alignments in Modes 5 and 6.
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 10CFR
50.92 and has concluded that the change does not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10CFR 50.92(c) are not
satisfied. The proposed changes do not involve [an] SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed change to Technical Specification 3/4.8.2.2 to
replace the wording ``As a minimum, one 125 volt battery bank and
its associated full capacity charger'' to ``As a minimum, one Train
(A or B) of batteries and their associated full capacity
chargers'' will increase the required battery banks operable from
one to two.[]
This change is being proposed to resolve an inconsistency with
Technical Specification 3/4.8.3.2 which currently requires two
battery banks energized in modes 5 and 6.
The proposed change to...Technical Specifications 3/4.8.2.2 and
3/4.8.3.2 to identify the specific equipment required and its
alignment during modes 5 and 6 is being proposed to reduce the
vagueness in the present Technical Specifications. This proposed
change will specify the equipment required operable for the
electrical distribution systems during modes 5 and 6.
These proposed changes are considered administrative and do not
alter the manner in which any system or component is operated or
expected to respond during an accident. Therefore, the proposed
changes do not involve a significant increase in the probability or
consequence of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to Technical Specification 3/4.8.2.2 to
increase the required battery banks operable from one to two and to
reword Technical Specifications 3/4.8.2.2 and 3/4.8.3.2 to identify
the specific equipment required operable during modes 5 and 6 do not
alter the manner in which any system or component is operated or
expected to respond during normal or accident conditions.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to Technical Specification 3/4.8.2.2 to
increase the required battery banks operable from one to two is
being proposed to resolve an inconsistency with Technical
Specification 3/4.8.3.2 which currently requires two battery banks
energized in modes 5 and 6. This is considered an administrative
change.
The proposed changes to...Technical Specifications 3/4.8.2.2 and
3/4.8.3.2 are being proposed to reduce the vagueness in the present
technical specifications by identifying the specific equipment
required operable during modes 5 and 6. The change will provide a
greater level of assurance that the electrical distribution systems
will be correctly aligned and surveilled. This is also considered an
administrative change.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed changes to not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: May 5, 1997
Description of amendment request: Technical Specification
Surveillance 4.8.4.1 requires periodic testing of lower voltage circuit
breakers for all containment penetration conductor overcurrent
protective devices. The proposed amendment would modify the
requirements for determining the operability of lower voltage circuit
breakers by using the manufacturer's curve of current versus time to
test delay trip elements, clarify the use of two pole in series
testing, and expand the Bases description of the testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed change to Technical Specification Surveillance
4.8.4.1 to modify the requirements for determining the operability
of lower voltage circuit breakers by using the manufacture's curve
of current versus time to test long time and short-time delay trip
elements will not change the requirement that periodic testing be
performed to determine breaker operability. The circuit breaker
testing is consistent with the design of the components and
performing surveillance testing does not involve a significant
increase in the probability of an accident previously evaluated. The
proposed change will provide assurance that the breakers will
perform consistent with accident analyses and does not involve a
significant increase in the consequence of an accident previously
evaluated.
The proposed change to the surveillance to modify the wording
associated with the use of two pole in series testing to determine
Molded Case Circuit Breaker (MCCB) operability following the failure
of [an] MCCB to pass a single pole test was previously approved in
License Amendment No. 13. The modified wording clarifies the testing
by specifically stating in the surveillance that the two pole in
series test determines MCCB operability. This is considered an
administrative change.
The proposed change to expand the description of the long-time
and short-time delay trip elements testing in the Bases Section is
also considered an administrative change.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
[[Page 30638]]
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change to use a curve of current versus time
instead of the description in Technical Specification Surveillance
4.8.4.1 of the [] long-time and short-time delay trip element
testing does not alter the design, operation, or maintenance of the
lower voltage circuit breakers.
The proposed change to the surveillance to modify the wording
associated with the use of two pole in series testing to determine
MCCB operability and the expanded description of the long-time and
short-time delay elements testing in the Bases Section are
considered administrative changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The current wording of Technical Specification Surveillance
4.8.4.1 requires testing of long-time delay trip elements with a
current value of exactly 300% of the pickup setting and short-time
delay trip elements with a current value of exactly 150% of the
pickup setting. The testing [cannot] be performed at exact values.
Circuit breaker manufactures develop a curve of current versus time
for each breaker type that specifies the allowable time to trip for
a specified current. Using the curve for a given breaker type, the
operability of a circuit breaker can be verified by inserting a
given current and verifying that the breaker trips within the
allowable time delay band width for that current. Testing by the
industry is typically performed at approximately 300% of the pickup
setting for long-time delay trip elements and approximately 150% of
the pickup setting for short-time delay trip elements. The proposed
change to the surveillance to modify the requirements for
determining the operability of circuit breakers by using the
manufacturer's curve of current versus time to test delay trip
elements will continue to provide assurance that lower voltage
circuit breakers for all containment penetration conductor
overcurrent protective devices will operate consistent with the
assumptions of the accident analysis.
The proposed change to the surveillance to modify the wording
associated with the use of two pole in series testing to determine
MCCB operability and the expanded description of the long-time and
short-time delay trip elements testing in the Bases Section are
considered administrative changes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed changes do not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut
Date of amendment request: May 5, 1997
Description of amendment request: Technical Specification
Surveillance 4.5.2.b.1 requires that the emergency core cooling system
(ECCS) piping be verified full of water at least once per 31 days. The
proposed amendment would revise the surveillance to exempt the
operating charging pump(s) and associated piping from the requirement
to be verified full of water and move the description of the
verification method from the surveillance to the Bases section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed change to Technical Specification Surveillance
4.5.2.b.1 to exempt the operating centrifugal charging pump(s) and
associated piping from the requirement to be vented will not effect
the requirement the ECCS piping be full of water. An operating
centrifugal charging pump and the associated piping is self venting
and cannot develop voids and pockets of entrained gases. This change
is consistent with the design of the charging system and ensuring
that ECCS piping is full of water does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
The proposed change Technical Specification Surveillance
4.5.2.b.1 to move and expand the description of the venting method
from the surveillance to the Bases Section are considered
administrative changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change to exempt the operating centrifugal charging
pump(s) and associated piping from the requirement to be
periodically vented by crediting its self venting capabilities does
not change the operation of the charging system or any of its
components during normal or accident evaluations.
The proposed changes to move and expand the description of the
venting method from the surveillance to the Bases Section are
considered administrative changes.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to Technical Specification Surveillance
4.5.2.b.1 to exempt the operating centrifugal charging pump(s) and
associated piping from the requirement to be manually vented by
crediting its self venting capabilities, is consistent with the
design of the charging system. This proposed change continues to
ensure that ECCS piping is full of water and thus, does not involve
a significant reduction in a margin of safety.
The proposed change to Technical Specification Surveillance
4.5.2.b.1 to move the description of the venting method from
thesurveillance to the Bases Section is considered an administrative
change. Currently the surveillance identifies that ECCS piping is to
be verified full of water by venting ECCS pump casings and
accessible discharge piping high points except for the RSS
[recirculation spray system] pump, RSS heat exchanger and associated
RSS piping that are not maintained filled with water during plant
operation. The venting description will be expanded when moved to
the bases to include an exclusion for the above described operating
centrifugal charging pump(s) and associated piping and the venting
method used for nonoperating centrifugal charging pumps. The
centrifugal charging pumps have top mounted suction and discharge
nozzles and do not have casing vents. The pump manufacturer has
indicated that venting the pump suction pipe will assure that the
pump is full of water. This venting of the nonoperating centrifugal
charging pumps is accomplished by using a pump suction line test
connection.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed change does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
[[Page 30639]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: April 17, 1997
Description of amendment request: This license amendment request
revises Technical Specification (TS) 2.12, ``Control Room System,'' to
delete the Limiting Conditions of Operation (LCO) and associated
surveillance for the control room temperature and replace it with an
LCO and surveillance on the control room air conditioning (A/C) system.
The remainder of TS 2.12 is being rewritten consistent with the
requirements of the Combustion Engineering Standard TS (NUREG-1432,
Rev. 1). In reviewing requirements for refueling and shutdown
operations, additional TS improvement were identified. Therefore, the
definition section, TS 2.1 ``Reactor Coolant System,'' 2.6
``Containment System,'' 2.8 ``Refueling Operations,'' and associated
surveillance requirements are proposed for revision to incorporate the
design basis requirements for refueling operations and to correspond to
NUREG-1432.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change will incorporate new requirements for the
control room air conditioning system, control room filtration
system, and refueling operations. In addition, the proposed change
will ensure that the Limiting Condition for Operations and
surveillance requirements are consistent with the design basis of a
fuel handling accident as documented in the FCS Updated Safety
Analysis Report (USAR).
CONTROL ROOM SYSTEMS
The control room air conditioning (A/C) system consists of two
redundant A/C units, VA-46A and VA-46B. Each unit has sufficient
capacity to meet the cooling requirements for personnel and
equipment inside the control room envelope. Each A/C unit is
equipped with an air-cooled condenser located inside a protective
enclosure outdoors on the roof of the Auxiliary Building. Each A/C
unit's refrigerant compressor, air cooling coils, fans, and dampers
are located inside of the control room envelope. Each unit has a
waterside economizer coil that allows air cooling with Component
Cooling Water (CCW). When cooling water temperature is sufficiently
low, a temperature-activated valve at each A/C unit allows cooling
water flow through the waterside economizer. This valve also diverts
flow away from the waterside economizer if cooling water temperature
is too high. The air-operated CCW isolation valves to the A/C units
fail closed and are automatically closed on a Ventilation Isolation
Actuation Signal (VIAS) to prevent CCW flow through the waterside
economizers in a post-accident situation.
Technical Specification (TS) 2.12(1) requires that the
temperature within the control room and control cabinets be
maintained below 120 deg.F does not meet any of the four criteria
contained in 10 CFR 50.36 for inclusion in TS. However, the
equipment required to maintain this temperature, the control room
air conditioning system, meets Criterion 3 of 10 CFR 50.36 in that
the system functions to mitigate a design basis accident by
maintaining the control room in a habitable environment.
Therefore, it is proposed that this TS be revised to delete the
control room temperature as a LCO and require that two control room
air conditioning trains be operable when the reactor coolant
temperature is above 210 deg.F. The design temperature limits of
instrumentation and controls inside of the control room will be
maintained in the Basis Section of TS 2.12.
The allowed outage time for one train of control room air
conditioning is proposed as 30 days. This is consistent with
Combustion Engineering Standard TS 3.7.12 (NUREG-1432 Rev. 1). In
addition, the FCS Probabilistic Risk Assessment model was reviewed
and validated a 30 day outage time as being non-risk significant.
The impact on Core Damage Frequency (CDF) from a 30 day LCO was
based on the assumption that one cooling unit was always inoperable
and thus under the LCO for an entire year. This allows the analysis
to consider unlimited entries into the LCO and a full LCO duration
for each entry. Using this assumption, the baseline (annually) CDF
of 1.53E-5 would increase by 21.6% to a frequency of 1.86E-5. In
accordance with EPRI's ``PSA Applications Guide,'' this small
increase in CDF can be classified as ``non-risk significant.''
Specification 2.12(2)
Specification 2.12(2) requires that a thermometer be in the
control room at all times. This instrumentation does not meet any of
the four criteria contained in 10 CFR 50.36 for inclusion in the FCS
TS. Therefore, the requirement is proposed for relocation to the FCS
USAR.
Specification 2.12(3)
Specification 2.12(3) requires that all areas of the plant
containing safety related instrumentation be observed during hot
functional testing to determine local temperatures and monitored
during operation if normal plant ventilation is not available. It is
proposed to delete this TS. The requirement to monitor and determine
local temperatures during hot functional testing was met during the
initial startup phase of FCS and is no longer applicable. The
requirement to monitor temperatures within the plant during normal
operation does not meet any of the four criteria contained in 10 CFR
50.36 for inclusion in TS and therefore is being deleted.
The requirement to control temperatures for safety related
instrumentation and controls, and initiate supplementary cooling if
required, is currently described in USAR Section 9.10. These USAR
requirements are controlled by plant procedures. Any changes to
these requirements would require an evaluation be conducted in
accordance with 10 CFR 50.59.
Specification 2.12(4)
Specification 2.12(4) allows one control room air filtration
system to be inoperable for 7 days or a plant shutdown be commenced.
This specification does not state which modes of operation it
applies to.
Therefore, it is proposed to revise this specification to
require two trains of control room air filtration systems to be
operable when the reactor coolant temperature is above 210 deg.F.
The allowed outage time will be maintained at 7 days and a total of
42 hours will be allowed to take the plant to cold shutdown. The 42
hour time period is consistent with TS 2.0.1 which addresses
equipment outages in excess of what is specifically allowed by
individual specifications.
The proposed changes for the control room systems consist of
providing additional restrictions on operation of the control room
air filtration systems and control room air conditioning system.
These changes ensure that equipment required to mitigate the
consequences of an accident are operable. Therefore, the proposed
changes do not increase the probability or consequences of an
accident previously evaluated.
REFUELING OPERATIONS
The design bases of the fuel handling accident and refueling
operations were reviewed and several inadequacies were identified
related to refueling operations. Therefore, revisions are proposed
for the TS Definition section, TS 2.6 on containment integrity, and
TS 2.8 on refueling operations to reflect NUREG-1432.
Definitions
Cold Shutdown Condition & Refueling Shutdown Condition
The changes proposed for the definitions of Cold Shutdown
Condition, and Refueling Shutdown Condition clarify these
definitions. The plant is in Cold Shutdown when Tcold is
less than 210 deg.F, and the reactor coolant is at least Shutdown
Boron Concentration but less than Refueling Boron Concentration.
Similarly, the definition for Refueling Shutdown is clarified to
apply when Tcold is less than 210 deg.F and the reactor
[[Page 30640]]
coolant is at least Refueling Boron Concentration. This change does
not propose any new operating modes but merely clarifies when the
definitions are applicable.
Core Alterations
The definition for Core Alterations is being revised to reflect
the requirements of NUREG-1432. This revision deletes ``any
component'' from the definition and clarifies that the components
considered by this definition are those that could affect
reactivity. In addition, the revision adds nuclear fuel to the
definition such that movement of fuel within the reactor vessel will
be defined as a core alteration and not a refueling operation.
Refueling Operations
The definition of Refueling Operations is being revised to
delete control element assemblies (CEA) or startup sources from the
definition since these are items that are included in the definition
of Core Alterations. Additionally, it is being revised to specify
that the definition is limited to movement of irradiated fuel
outside of the reactor pressure vessel since fuel movement inside
the reactor vessel is included in the definition of Core Alteration.
Finally, a clarification is being added to state that suspension of
refueling operations shall not preclude completion of movement of
irradiated fuel to a safe, conservative position.
In Operation
The definition of In Operation is being revised to include the
definition of operable. This is a more conservative interpretation
than currently exists.
Specification 2.1 ``Reactor Coolant System''
It is proposed to revise TS 2.1.1(3) to include shutdown cooling
requirements when the reactor coolant system (RCS) temperature is
below 210 deg.F with fuel in the reactor and the reactor vessel head
fully tensioned. The definitions of Cold Shutdown (Mode 4) and
Refueling Shutdown (Mode 5) contained in the TS make no distinction
as to the status of the reactor vessel head or RCS temperature. The
only difference between the two defined modes is boron
concentration. Higher or lower boron concentration affects shutdown
margin but does not affect decay heat load, which is the basis for
this specification.
Technical Specification 2.1.1(4) was intended to address
shutdown cooling requirements during refueling operations. However,
this is already addressed in TS 2.8. Therefore, it is proposed to
delete TS 2.1.1(4) and the exception since new specifications
addressing shutdown cooling loop requirements during Mode 5 with
fuel in the reactor and with one or more reactor vessel head closure
bolts less than fully tensioned are proposed for inclusion in TS 2.8
(Refueling Operations).
The associated statements supporting these items in the Basis
section are also proposed for deletion. Prior to any reactor vessel
head closure bolts being loosened, TS 2.1.1 will be applicable which
will require two shutdown cooling loops. As soon as a closure bolt
is loosened, the new proposed TS 2.8 would be applicable which also
requires two shutdown cooling loops whenever there is less than 23
feet of water above the core. The requirements of TS 2.1.1(3) are
similar to NUREG-1432, Specifications 3.4.7 and 3.4.8.
Specification 2.6 ``Containment System''
Currently, TS 2.6(1)c states that containment integrity shall
not be violated when the reactor vessel head is removed if the boron
concentration is less than refueling concentration. However,
Specification 2.6(1)c has no required actions and therefore, TS
2.0.1 must be entered when the LCO is not met. In this situation,
(reactor vessel head removed), TS 2.0.1 is ineffective because the
plant would already be in Refueling Shutdown. Thus, TS 2.6(1)c is
proposed for deletion.
Currently, Specification 2.6(1)d requires that except for
testing one control element drive mechanism at a time, positive
reactivity changes shall not be made by CEA motion or boron dilution
unless containment integrity is intact. Specification 2.6(1)d is
proposed for deletion as it is unnecessarily restrictive.
Specification 2.8.1(1) as proposed eliminates the need for
containment integrity when the reactor is in Refueling Shutdown.
Specification 2.8.1(1) requires sufficient shutdown margin to
preclude a criticality event and also prescribes actions to restore
the shutdown margin if necessary. Small positive reactivity
increases whether by CEA motion or boron dilution will not cause a
criticality event due to the need to maintain at least a 5% shutdown
margin. Therefore, the requirement to maintain containment integrity
is unnecessarily restrictive since a criticality event cannot occur
when a shutdown margin of at least 5% exists. Specification 2.8.1(1)
is consistent with the requirements of NUREG-1432, Specification
3.9.1.
A new specification (TS 2.8.2(1)) is proposed that provides
requirements for containment closure during core alterations and
refueling operations inside of containment. The design basis of the
Fort Calhoun Station does not require full containment integrity
during a fuel handling accident. As stated in USAR Section 14.18,
the fuel handling accident does not take credit for containment
isolation. Therefore, requiring full containment integrity is
inappropriate and requirements for containment closure are proposed
for addition to TS 2.8 consistent with NUREG-1432 Specification
3.9.2.
Specification 2.10.2 governs operation of CEAs and monitoring of
selected core parameters. Specification 2.10.2 ensures (1) adequate
shutdown margin following a reactor trip, (2) that the moderator
temperature coefficient (MTC) is within the limits of the safety
analysis, and (3) CEA operation is within the limits of the setpoint
and safety analysis. Specification 2.10.2 ensures that the reactor
will be maintained sufficiently subcritical to preclude inadvertent
criticality and provides actions (i.e., boration) to be taken to
ensure that the required shutdown margin is available. Thus, TS
2.10.2 precludes the need for containment integrity when the plant
is in cold shutdown.
Specification 2.8 ``Refueling Operations''
It is proposed that TS 2.8 be rewritten to reflect NUREG-1432.
Currently, this specification applies to any refueling operation.
However, no distinction is made between refueling operations within
containment and refueling operations within the spent fuel pool. In
addition, several initial assumptions of a fuel handling accident
are not addressed by the current TS 2.8.
Specification 2.8(1)
The current TS 2.8(1) is inadequate. This specification requires
that the equipment hatch and one door in the Personnel Air Lock be
properly closed, and all automatic containment isolation valves be
operable or at least one valve closed. The specification does not
define what is meant by a properly closed equipment hatch; that
information is currently contained in the Basis of TS 2.1.1. In
addition, inclusion of all automatic containment isolation valves
instead of those providing direct access to the outside atmosphere
is incorrect.
The containment isolation system is defined in USAR Section
5.9.5 as those devices actuated by a Containment Isolation Actuation
Signal (CIAS) or a Steam Generator Isolation Signal (SGIS). This
includes many valves that have no design basis function during a
fuel handling accident. A CIAS is initiated by a Containment
Pressure High Signal or a Pressurizer Pressure Low Signal. Neither
of these signals are required to be operable during refueling
operations as these signals would/could not respond to a fuel
handling accident.
The correct requirements are specified in TS 2.8(2) which only
requires that closure be initiated by the Ventilation Isolation
Actuation Signal (VIAS) for the containment pressure relief, air
sample, and purge system valves. Due to these inadequacies, it is
proposed to delete TS 2.8(1) and replace it with a new Specification
2.8.2(1) which is consistent with NUREG-1432 Specification 3.9.3.
Specification 2.8(2)
It is proposed that TS 2.8(2) be deleted and replaced by new
Specifications 2.8.2(3) and 2.8.3(5). The requirement to maintain an
operable Ventilation Isolation Actuation Signal with input from the
containment atmosphere gaseous and auxiliary building exhaust stack
gaseous radiation monitors is consistent with current requirements
and required actions are consistent with NUREG-1432, Specification
3.3.8. Radiation Monitor RM-052 functions as a ``swing'' monitor,
i.e., it can be aligned to monitor either containment or the
auxiliary building exhaust ventilation stack. Radiation Monitor RM-
052 is powered by either MCC-3B1/AI-40C (like RM-051) or MCC-4C2/AI-
40D (like RM-062).
Technical Specification 2.7, Electrical System is not required
to be applied when the RCS is below 300 deg.F. Above 300 deg.F, TS
2.7 requires both 4160-VAC buses to be operable. Thus, above
300 deg.F the required radiation monitors must be powered from
independent 480-VAC buses supplied by independent 4160-VAC buses.
During refueling outages, bus alignments other than those used
during power operation are used to permit electrical system
maintenance and modifications.
In the loss of offsite power event, the radiation monitor sample
pumps and control room HVAC units stop and will not restart
[[Page 30641]]
until the emergency diesel generators (EDGs) reenergize the system.
The fuel handling equipment also stops and does not restart when the
EDGs reenergize the system, thus minimizing the potential of a fuel
handling accident. When the EDGs reenergize the buses, VIAS will
operate as designed. Therefore, when the RCS is below 300 deg.F, the
required monitors need only be powered from independent 480-VAC
buses supplied by a single 4160-VAC bus.
There is no need to assume that a fuel handling accident occurs
immediately followed by a loss of offsite power. However, in the
unlikely event that this should occur, there would be no effect on
the site boundary dose since VIAS is not credited in USAR Section
14.18 (Fuel Handling Accident). In this situation, when the EDGs
reenergize the buses, the control room HVAC units will restart in
the filtered air makeup mode and the stack radiation monitor sample
pump will restart. However, the containment radiation monitor sample
lines remain isolated preventing the restart of the monitor sample
pump after receipt of a VIAS.
Specification 2.8(3)
It is proposed that TS 2.8(3) be deleted. This requirement does
not meet any of the four criteria contained in 10 CFR 50.36 for
inclusion in the TS. The requirement that radiation levels in
containment and the spent fuel pool shall be monitored during
refueling operations will be incorporated into the FCS USAR.
Specification 2.8(6)
It is proposed that TS 2.8(6) be deleted. This requirement does
not meet any of the four criteria contained in 10 CFR 50.36 for
inclusion in the TS. The requirements that direct communication
between personnel in the control room and at the refueling machine
shall be available whenever core alterations are taking place will
be incorporated into the FCS USAR.
Specification 2.8(7)
It is proposed that TS 2.8(7) be deleted and replaced with a new
Specification 2.8.3(4). The requirement to place the spent fuel pool
ventilation system in operation prior to refueling operations is
consistent with the current TS. It is being clarified that this
specification only applies to refueling operations in the spent fuel
pool, and not when conducting refueling operations inside of
containment. Additionally, it is being clarified that TS 2.0.1 is
not applicable to this activity, as reactor operation is independent
of fuel movements in the spent fuel pool.
Specification 2.8(9)
The current Specification 2.8(9) is inadequate. This
specification requires a minimum of 23 feet of water above the top
of the core. This does not meet the initial conditions assumed in
the fuel handling accident as documented in USAR Section 14.18. USAR
Section 14.18 assumes 23 feet of water above where the fuel could
land if dropped. In order to meet this initial condition, a minimum
of 23 feet of water above the reactor vessel flange is required, as
this is the highest point where a fuel bundle could land if dropped.
Procedures reflect the requirement to maintain 23 feet of water
above the reactor vessel flange during refueling operations. The
proposed revision is consistent with NUREG-1432, Specification
3.7.16.
Specification 2.8(11)
The current specification is inadequate. The specification
provides restrictions on storage of fuel in the spent fuel pool;
however, there are no required actions to address situations when
the specification is not met. It is proposed that TS 2.8(11) be
deleted and replaced with a new Specification 2.8.3(1) that requires
that a misloaded fuel assembly be moved immediately. Additionally,
it is being clarified that TS 2.0.1 is not applicable to this
activity, as reactor operation is independent of fuel movements in
the spent fuel pool.
Specification 2.8(12)
It is proposed that TS 2.8(12) be deleted and replaced with a
new Specification 2.8.3(3). The requirement to maintain 500 ppm
boron concentration in the spent fuel pool whenever unirradiated
fuel is stored there is consistent with the current TS and the
required actions are consistent with NUREG-1432, Specification
3.7.17.
Restriction on Movement of Irradiated Fuel from the Reactor Core
The restriction on irradiated fuel movement unless the core has
been subcritical for at least 72 hours if the reactor has been
operated at power levels above 2% is proposed for relocation to the
Bases of TS 2.8.2(2). This requirement does not meet any of the four
criteria contained in 10 CFR 50.36 for inclusion in the TS. This is
consistent with NUREG-1432, B 3.9.6.
Reactor Coolant System Boron Concentration
Currently, there is no specification for boron concentration.
Refueling boron concentration is included in the definition of Mode
5. However, there are no required actions to be taken if the boron
concentration should be below refueling concentration. Therefore, it
is proposed that a new Specification 2.8.1(1) be incorporated
consistent with NUREG-1432, Specification 3.9.1.
Spent Fuel Pool Water Level
Currently, there is no specification for spent fuel pool water
level. The water level of the spent fuel pool is an initial
condition assumed in USAR Section 14.18. It is proposed that a new
Specification 2.8.3(2) be incorporated into TS 2.8, which is
consistent with NUREG-1432, Specification 3.7.16.
The proposed changes for the RCS and containment during
shutdown, and requirements for refueling operations, consist of
providing additional restrictions on operation, and changes to make
the requirements of the TS Limiting Conditions for Operation
consistent with the initial conditions and assumptions of the fuel
handling accident as documented in USAR Section 14.18. Therefore,
the proposed changes do not increase the probability or consequences
of an accident previously evaluated.
SURVEILLANCE REQUIREMENTS
CONTROL ROOM
Specification 3.1, Table 3-3, Item 13.
Specification 3.1, Table 3-3, Item 13 requires that the
thermometer in the control room be compared with a calibrated
thermometer and replaced if out of tolerance on a refueling
frequency. It is proposed that this surveillance be deleted to
be consistent with deletion of the LCO requirement to maintain a
thermometer in the control room.
A new surveillance is proposed to verify that the control room
air conditioning system has the capability to remove the assumed
heat load. This surveillance will ensure the operability
requirements for TS 2.12 are met. The test and frequency is
consistent with NUREG-1432.
The air-operated CCW isolation valves to the A/C units fail
closed and are automatically closed on a VIAS to prevent CCW flow
through the waterside economizers in a post-accident situation.
These valves are currently tested in accordance with TS 3.3 (FCS
Inservice Testing Program). Prior to the modification, the valves
were tested as fail-open valves. No TS changes are necessary.
The control room air filtration system is currently tested on a
refueling frequency in accordance with TS 3.2, Table 3-5, Item 10a.
No TS changes are necessary.
REFUELING OPERATIONS
Reactor Coolant Boron Concentration During Refueling Operations
The Reactor Coolant System boron concentration is currently
sampled in accordance with TS 3.2, Table 3-4, Item 1(e). It is
proposed to revise the frequency from once per shift during
refueling operations to once per 3 days which is consistent with
NUREG-1432. As stated in the basis of TS 2.8 and USAR Section 14.18,
the reactor cavity is filled with over 200,000 gallons of borated
water prior to the start of refueling operations. The requirements
for sampling the reactor coolant during the remainder of Mode 5 is
performed once per 3 days in accordance with Table 3-4, Item 1(d).
This proposed change will make the sampling consistent with the
requirements of Item 1(d) and NUREG-1432.
Spent Fuel Pool Boron Concentration
The spent fuel pool boron concentration is currently sampled in
accordance with TS 3.2, Table 3-4, Item 5. It is proposed to revise
the frequency of the sampling to prior to movement of unirradiated
fuel in the spent fuel pool and once per week whenever unirradiated
fuel is stored there to be consistent with the requirements of the
LCO.
Source Range Neutron Monitors
Currently, a channel check and calibration of the wide range
neutron monitors is performed in accordance with TS 3.1, Table 3-1,
Item 2.
Containment Penetrations
Currently, there is no surveillance to determine the status of
containment penetrations during refueling operations. Therefore, a
new surveillance is proposed for TS 3.2, Table 3-5 to verify the
status of required containment penetrations once per 7 days
consistent with NUREG-1432.
The requirement of NUREG-1432 to verify that the containment
purge and exhaust valves actuate to the isolation position on a
refueling frequency is currently tested as part of the Containment
Radiation High Signal test required by TS 3.1, Table 3-2. Item 4.
Shutdown Cooling Loops
Currently, there is no surveillance requirement to verify that
the required
[[Page 30642]]
shutdown cooling loops are operable and in operation or to verify
correct breaker lineup for the shutdown cooling loop that is not in
operation. Therefore a new surveillance is proposed to be
incorporated into TS 3.2, Table 3-5 consistent with NUREG-1432.
Refueling Water Level
Currently, there is no surveillance requirement to verify the
refueling water level during refueling operations. Therefore, a new
surveillance is proposed for incorporation into TS 3.2, Table 3-5
consistent with NUREG-1432.
Spent Fuel Pool Water Level
Currently, there is no surveillance requirement to verify the
spent fuel pool water level during refueling operations. Therefore,
a new surveillance is proposed for incorporation into TS 3.2, Table
3-5 consistent with NUREG-1432.
Spent Fuel Initial Enrichment/Burnup Verification
Currently, the requirement to conduct a verification of initial
enrichment and burnup of spent fuel that will be stored in Region 2
is included as a general requirement of TS 2.8. It is proposed to
relocate this requirement into a surveillance in TS 3.2, Table 3-5,
consistent with NUREG-1432.
The proposed changes for the surveillance requirements consist
of providing additional testing requirements to ensure that the
Limiting Condition for Operations will be met. One surveillance
frequency related to the sampling of the reactor coolant system
boron concentration during refueling operations is being reduced
from a frequency of once per shift to once every 3 days. However,
this frequency is consistent with the frequency of sampling during
the remainder of Mode 5 when fuel is in the
reactor and is more than adequate due to the large volume (over
200,000 gallons) of borated water required during refueling
operations. Therefore, the proposed changes do not increase the
probability or consequences of an accident previously evaluated.
ADMINISTRATIVE CHANGES
The remainder of TS 2.8 requirements of refueling operations are
proposed to be reformatted into individual TS LCOs. It is also
proposed that sampling frequencies of items contained in TS 3.2,
Table 3-4, (page 3-19), be revised to incorporate frequencies
defined in TS 3.0.2. Therefore, frequencies stated as once per 31
days will be noted as ``M,'' and frequencies stated as once per 7
days will be noted as ``W.'' These proposed changes have no effect
on the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There will be no physical alterations to the plant
configuration. No changes in operating modes are proposed although
minor changes to the definitions of Cold Shutdown Condition and
Refueling Shutdown Condition are proposed for clarification
purposes. The proposed changes incorporate additional restrictions
on the operation and testing of equipment required to mitigate an
accident and to ensure the initial conditions and assumptions of the
design basis accidents are maintained and controlled by the
Technical Specifications.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes ensure that assumptions of the fuel
handling accident are maintained by Technical Specification Limiting
Condition for Operation and surveillance requirements. The
assumptions of the fuel handling accident that may affect a margin
of safety are not being changed. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502
NRC Project Director: William H. Bateman
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: February 25, 1997
Description of amendment request: The proposed Technical
Specifications (TS) changes would amend the Limerick Generating Station
(LGS) Unit 1 and Unit 2 Facility Operating Licenses (FOLs), and
Appendix B of the licenses (i.e., Environmental Protection Plan (EPP)),
reflecting a corporate name change from Philadelphia Electric Company
to PECO Energy Company. In addition, the application would make changes
to the LGS Units 1 and 2, FOL, and Appendix A (i.e., TS) of the
licenses, which would remove obsolete information and correct
typographical errors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The company name change and typographical corrections are
editorial and will not alter the operation of equipment assumed to
be an initiator of any analyzed event or transients previously
evaluated. The license provisions were satisfactorily completed, and
as such, have no effect on any previously evaluated accident
scenario. The changes will not alter the operation of equipment
assumed to be available for the mitigation of accidents or
transients, nor will they alter the operation of equipment important
to safety previously evaluated.
Therefore, the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The company name change and typographical corrections are
editorial and will not involve any physical changes to the plant
systems, structures, or components. The license provisions were
satisfactorily completed, and as such, have no effect on any
previously evaluated accident scenario. The proposed changes do not
allow plant operation in any mode that is not already evaluated. The
changes will not alter the operation of equipment important to
safety previously evaluated.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The company name change and typographical corrections are
editorial and will not affect the manner in which the facility is
operated, or change equipment or features which affect the
operational characteristics of the facility. There is no margin of
safety as defined in the bases of any TS regarding the name of the
company, or affected by the corrections or deletion of obsolete
license provisions.
Therefore, these proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
[[Page 30643]]
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: March 24, 1997
Description of amendment request: The proposed Technical
specifications (TS) changes would delete the Drywell and Suppression
Chamber Purge System operational time limit, and add a surveillance
requirement to ensure the purge system large supply and exhaust valves
are closed as required.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
This activity does not increase the probability of occurrence of
an accident previously evaluated in the SAR [Safety Analysis
Report]. This activity involves deleting the allowable operating
limit (180 hours each 365 days) for the Drywell and Suppression
Chamber Purge system, while maintaining specific criteria for when
the valves are allowed to be open. These changes do not increase the
probability that this system will be in service should a LOCA [loss-
of-coolant-accident] occur and does not increase the probability
that a LOCA will occur. These changes also do not impact the
probability of occurrence of any anticipated operational occurrence,
other postulated design basis accident, or other event in which the
plant was designed to respond.
This activity does not increase the consequences of an accident
previously evaluated in the SAR. UFSAR [Updated Final Safety
Analysis Report] Section 9.4.5.1.2.2 for high volume purging,
although limiting the operating time the vent and purge system is to
be in service, evaluates the consequences of a LOCA should the vent
and purge valves be open. System operating procedures for venting
and purging assure the availability of SGTS [standby gas treatment
system] should a LOCA occur.
This activity will not increase the probability of a LOCA
occurring during the time the Drywell and Suppression Chamber Purge
system is in operation as previously evaluated. The Improved TS do
not identify a specific time limit value as long as the valves are
operated under the stated conditions (inerting, de inerting,
pressure control, ALARA [as low as reasonably achievable] or air
quality considerations for personnel entry or Surveillances that
require that the valves be open). These proposed changes will
incorporate the ITS [Improved Technical Specifications] operational
controls which will result in the same order of magnitude of
equipment malfunction probability as that provided by limiting
purging to 180 hours per 365 days. A LGS [Limerick Generating
Station] Level 2 PSA [Probability Risk Assessment] Analysis was
performed to determine the additional risk associated with changing
the operating limit from 90 hours to a nominal 500 hours each 365
days. This analysis concluded that the increase in risk of
containment failure is well within the bounds of the EPRI [Electric
Power Research Institute] PSA Applications Guideline for permanent
changes and the NRC [Nuclear Regulatory Commission] Staff's safety
goal value of 1.0 E-6 per year of reactor operation. Industry and
LGS historical operating experience confirms that the purging lines
are opened only for the specified reasons stated in ITS and for
periods which do not exceed the current magnitude of equipment
malfunction probability. Therefore, earlier engineering judgment is
being replaced by operating experience.
Failure of the operating SGTS filter bank following a LOCA has
been found to be acceptable due to the limited benefit derived from
SGTS for accident sequences important to plant risk and the
possibility that the backup filter bank would be available.
Additionally, as discussed in UFSAR Section 9.4.5.1.2.2, the failure
of SGTS during a LOCA does not contribute to any significant
releases and is bounded by the analysis performed to address
containment overpressure rupture.
Deleting the time limit restriction that the vent and purge line
isolation valves may be open does not increase the probability that
these valves will not perform as designed (close upon isolation
signal) in response to a LOCA. Removing the 180 hour requirement
will not increase the likelihood that the vent and purge valves will
be called upon to close from that previously evaluated. UFSAR
Section 6.2 states that the containment purge valves have undergone
extensive testing and analyses to demonstrate the operability of
these valves following a LOCA.
These changes do not directly or indirectly degrade the
performance of any other safety system (assumed to function in the
accident analysis) design basis. The potential for other equipment
failures in the reactor enclosure due to duct impact, impingement,
and the resulting environmental conditions was previously evaluated
in the LGS SAR. It was concluded that the environmental
qualifications for the LGS equipment are sufficient to ensure
operability under the predicted environmental condition, and, the
potential does not exist for impact or impingement - related damage
to essential equipment. Maintaining the existing SAR analysis and
retaining operating criteria for opening the containment purge
valves, demonstrates that the risk of equipment failure and
resulting radiological consequences will not increase.
Therefore, deleting the TS operating limit for the Drywell and
Suppression Chamber Purge system from 180 hours each 365 days and
the addition of a TS Surveillance Requirement verifying that the
purge valves are closed under certain conditions does not increase
the probability or consequences of an accident previously evaluated.
2. The proposed Technical Specifications changes do not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
This activity does not change the function of the Drywell and
Suppression Chamber Purge system, the containment isolation system,
or SGTS as previously evaluated. Deleting the operational time limit
that the vent and purge system is in service and the addition of a
surveillance requirement does not create an accident initiator not
already considered.
In addition, this activity does not create a failure mode not
considered. All evaluated equipment failures that could occur as a
result of a LOCA during high volume purging have previously been
identified and evaluated. Therefore, these changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed Technical Specifications changes do not involve
a significant reduction in a margin of safety.
The bases of TS 3.6.1.8 state that the 180 hour each 365 day
operating limit for the Drywell and Suppression Chamber Purge system
is imposed to protect the integrity of the SGTS filters. The LGS
Offsite Dose Calculation Manual assures the availability of the
backup SGTS filter train during operation of the vent and purge
system. Furthermore, deleting the operating limit (180 hours each
365 days) does not reduce the margin of safety since specific
criteria for opening the purge valves is being maintained and does
not involve an increase in risk. Therefore, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
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: April 9, 1997
Description of amendment request: The proposed Technical
Specifications (TS) changes would clarify existing battery specific
gravity requirements, delete the requirement to correct specific
gravity values based on electrolyte level, and allow the use of
charging current measurements to verify the batterys state of charge.
Basis for proposed no significant hazards consideration
determination:
[[Page 30644]]
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 changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
Changes to Technical Specifications surveillance requirements
for specific gravity and Technical Specifications Bases commitments
do not change the frequency or consequences of any accident
previously evaluated. The proposed changes which commit to IEEE
Standard 450-1995 for specific gravity testing, providing battery
charging current as an alternate method to specific gravity
measurements, and eliminating the commitment to perform electrolyte
level correction do not prevent the DC system from performing its
intended safety function. The proposed changes to the Technical
Specification battery surveillance requirements and commitment to
IEEE Standard 450-1995 for specific gravity are in accordance with
current industry practices. These changes do not reduce the
readiness and performance of the 1E DC power system to perform its
intended function during a design basis event.
The proposed changes do not affect seismic specifications,
separation criteria or environmental qualifications. The proposed
changes do not impose an increase in or more severe test
requirements, an increase in the frequency of operation, reduce
independence or redundancy, modify the system or equipment
protective features, introduce new equipment failures or impose
additional loads than any previously evaluated. The Class 1E battery
system will continue to meet all of the design standards applicable
to the system and will not cause the system to operate outside of
its design or testing limits.
Batteries or battery chargers and their failure are not
initiators of the accidents previously evaluated. The proposed
changes do not affect, degrade or prevent the response of active or
passive systems described or assumed in the LGS accidents previously
evaluated. In addition, the proposed TS changes will improve the
availability of the station batteries.
Therefore, the changes will not increase the probability or
consequences of an accident previously evaluated.
2. The proposed Technical Specifications changes do not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The proposed Technical Specifications changes which will revise
the surveillance requirements and the TS Bases, do not increase the
failure rate of the battery. The proposed changes clarify and
enhance Operation's focus on the key battery parameters which will
improve the availability of the station batteries. The station
batteries are not accident initiators. The single failure of an
electrical component was previously evaluated in the LGS accident
analysis. Unexpected failures beyond the postulated single failure
are no more likely to occur under the clarified surveillance
requirements.
Therefore, these changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The proposed Technical Specifications changes do not involve
a significant reduction in a margin of safety.
The revision clarifies and reduces the battery surveillance
requirements for specific gravity. The revision eliminates the
possibility for misinterpretation and provides consistency of the
surveillance requirements. The specific gravity value for each
connected cell is being revised to reflect a discrete number which
meets the existing manufacturer's recommendations and does not
differ from the value described in the present bases. LGS is
currently committed to earlier revisions of IEEE Standard 450 (i.e.,
1975 and 1980), and the incorporation of IEEE Standard 450-1995 for
specific gravity will reflect current industry practices regarding
specific gravity.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: April 30, 1997 (TS 97-01)
Description of amendment request: The proposed amendment would
change the design features section of the Technical Specifications to
provide for insertion of Lead Test Assemblies (LTAs) containing Tritium
Producing Burnable Absorber Rods (TPBARs) in the Watts Bar Nuclear
Plant (WBN) reactor core during Cycle 2. The purpose of the change is
to provide irradiation services to support U.S. Department of Energy
(DOE) investigations into the feasibility of using commercial light
water reactors to maintain the DOE inventory of tritium.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
LTAs do not adversely affect reactor neutronic or thermal-
hydraulic performance; therefore, they do not significantly increase
the probability of accidents or equipment malfunctions while in the
reactor. The neutronic behavior of the LTAs mimics that of standard
burnable absorbers with only slight differences which are
accommodated in the core design. The reload safety analysis
performed for WBN Unit 1, Cycle 2 will confirm that any minor
effects of LTAs on the reload core will be within established fuel
design limits.
As described in DOE Technical Report PNNL-11419, Revision 1, the
LTA design is robust to all accident conditions except the large
loss of coolant accident where the rods are susceptible to failure.
However, the failure of the small number of TPBARS rods has been
determined to have an insignificant effect on the thermal hydraulic
response of the core to this event.
The impacts of LTAs on the radiological consequences for certain
postulated events [as shown in Table 6-1 of the licensee's
submittal, including Large Break LOCAs{time} are very small, and
they remain within 10 CFR 100 regulatory limits. The additional
offsite doses due to tritium leakage from the containment are small
with respect to loss of coolant accident source terms and are well
within regulatory limits.
The LTAs will not result in an increase in combustible gas
released to the containment. Therefore, the LTAs do not result in a
significant increase in the consequences of those previously
considered.
Analysis has shown that TPBARs will not fail during Condition I
through IV events, with the exception of a Large Break LOCA. The
radiological consequences of the non-Large-Break LOCA events are
essentially unchanged by the expected TPBAR tritium leakage to
reactor coolant, and doses remain within a small fraction of 10 CFR
100 regulatory limits. Therefore, there is no significant increase
in the consequences of these previously evaluated accidents.
The expected occupational and offsite doses, as reported in
Technical report PNNL-11419, Revision 1, resulting from release of
tritium from TPBARs over the plant operating cycle, including
refueling, are not significantly increased and are within applicable
regulatory limits.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
LTAs have been designed to be compatible with existing
Westinghouse 17x17 fuel assemblies and conventional Burnable Poison
Rod Assembly (BPRA) handling tools, equipment, and procedures, and
therefore no new accidents or equipment malfunctions are created by
the handling of LTAs.
LTAs use materials with known and predictable performance
characteristics and are compatible with PWR coolant. The LTA design
has specifically included material similar to those used in standard
burnable absorber rods with the exception of internal
[[Page 30645]]
assemblies used in the production and retention of tritium. As
described in the technical report, these materials are compatible
with the reactor coolant system and the core design. For the
irradiation proposed, the quantities of these materials is small.
Therefore, no new accidents or equipment malfunctions are created by
the presence of the LTAs in the reactor coolant system.
Thermal-hydraulic criteria have been established to ensure that
TPBARs will not fail during Condition I or II events. Analysis has
shown that TPBARs, appropriately positioned in the core, operate
within the established thermal-hydraulic criteria. Therefore, no new
accidents or equipment malfunctions are created by the presence of
the LTAs in the reactor.
Analysis has shown that TPBARs will not fail during Condition
III and IV events, with the exception of a large-break loss-of-
coolant-accident. The radiological consequences of these events are
small, with doses that are a small fraction of the 10 CFR 100
limits. Therefore there is no significant increase in consequences
of these previously evaluated accidents.
LTAs do not adversely affect reactor neutronic or thermal-
hydraulic performance; therefore, they do not create the possibility
of accidents or equipment malfunctions of a different type than
previously evaluated while in the reactor.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
LTAs do not adversely affect reactor neutronic or thermal-
hydraulic performance. Analysis indicates that reactor core behavior
and offsite doses remain relatively unchanged. TPBAR performance
under Condition I, II, III, and IV events are very similar to
standard burnable absorber rods previously evaluated. For these
reasons, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: April 18, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3/4.3.2, ``Safety System
Instrumentation,'' and TS Section 3/4.5.2, ``Emergency Core Cooling
Systems - ECCS Subsystems - Tavg (greater than or equal to)
280 deg.F.'' Certain surveillance intervals would be changed from 18
months to once each refueling interval, and certain setpoints would be
changed. The associated bases would also be changed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS) has
reviewed the proposed changes and determined that a significant
hazards consideration does not exist because operation of the DBNPS,
in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the initiation of such
accidents are not affected by the proposed revisions to increase the
surveillance test intervals from 18 to 24 months for TS 3/4.3.2.1,
``Safety Features Actuation System Instrumentation,'' and TS 3/
4.5.2, ``Emergency Core Cooling Systems - ECCS Subsystems - Tavg
(greater than or equal to) 280F.'' Initiating conditions and
assumptions remain as previously analyzed for accidents in the DBNPS
Updated Safety Analysis Report.
Results of the instrument drift study analysis and review of
historical 18-month surveillance data and applicable maintenance
records support an increase in the surveillance test intervals from
18 to 24 months (and up to 30 months on a non-routine basis)
because: the projected instrument errors caused by drift are bounded
by the existing setpoint analysis or a new analysis has been
performed incorporating a more conservative setpoint; and no
potential for a significant increase in a failure rate of a system
or component was identified during surveillance data and applicable
maintenance records reviews.
These proposed revisions are consistent with the NRC guidance on
evaluating and proposing such revisions as provided in Generic
Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2,
1991.
The proposed revisions to Allowable Values for Safety Features
Actuation System (SFAS) Reactor Coolant System (RCS) Pressure - Low,
RCS Pressure - Low-Low, RCS Pressure - Low-Low bypass permissive,
and Decay Heat Isolation Valve and Pressurizer Heater Interlocks
have no bearing on the probability of the initiation of an accident
previously evaluated.
The application of the Allowable Value to only the Channel
Functional Test and not the Channel Calibration, the proposed
deletion of the Trip Setpoints, the proposed revision of the TS
3.3.2.1 Limiting Condition for Operation (LCO) and Action Statement
3.3.2.1.a, and the proposed revisions to Actions 13 and 14 of TS
Table 3.3-3, are associated with the proposed revision of the
Allowable Values for SFAS RCS Pressure - Low, RCS Pressure - Low-
Low, and Decay Heat Isolation Valve and Pressurizer Heater
Interlocks, and are consistent with NUREG-1430, Revision 1,
``Standard Technical Specifications, Babcock and Wilcox Plants,''
dated April 1995. The proposed revisions have no bearing on the
probability of the initiation of an accident previously evaluated.
The proposed changes to TS Bases 3/4.3.1 and 3/4.3.2, ``Reactor
Protection System and Safety System Instrumentation,'' and TS Bases
3/4.5.2 and 3/4.5.3, ``ECCS Subsystems,'' are administrative changes
associated with the other proposed changes, and do not affect
previously analyzed accidents.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the slight increase in doses
due to a letdown line break event as a result of the proposed change
to the SFAS RCS Pressure - Low Allowable Value still satisfy the NRC
Standard Review Plan Section 15.6.2 acceptance criteria that doses
do not exceed a small fraction (10%) of the 10 CFR 100 guideline
values. The remaining proposed changes to Allowable Values, and the
other changes proposed by this License Amendment Request do not
increase the radiological consequences of previously analyzed
accidents because the source term, containment isolation, or
radiological releases are not being changed by the proposed
revisions.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated, for the reasons
discussed below.
No changes are being proposed to the type of testing currently
being performed, only to the length of the surveillance test
interval.
Results of the instrument drift study analysis and review of
historical 18-month surveillance data and maintenance records
support an increase in the surveillance test intervals from 18 to 24
months (and up to 30 months on a non-routine basis) because: the
projected instrument errors caused by drift are bounded by the
existing setpoint analysis or a new analysis has been performed
incorporating a more conservative setpoint; and no potential for a
significant increase in a failure rate of a system or component was
identified during surveillance data and applicable maintenance
records reviews.
The proposed revisions to Allowable Values for SFAS RCS Pressure
- Low, RCS Pressure - Low-Low, RCS Pressure Low-Low bypass
permissive, and Decay Heat Isolation Valve and Pressurizer Heater
Interlocks, do not alter the type of any testing currently being
performed.
The application of the Allowable Value to only the Channel
Functional Test and not the Channel Calibration, the proposed
deletion of the Trip Setpoints, revision of the TS 3.3.2.1 LCO and
Action Statement 3.3.2.1.a, and the proposed revisions to Actions 13
and 14 of
[[Page 30646]]
TS Table 3.3-3, are associated with the proposed revision to the
Allowable Values for SFAS RCS Pressure - Low, RCS Pressure - Low-
Low, RCS Pressure Low-Low bypass permissive, and Decay Heat
Isolation Valve and Pressurizer Heater Interlocks, and are
consistent with NUREG-1430, Revision 1, ``Standard Technical
Specifications, Babcock and Wilcox Plants,'' dated April 1995. The
proposed revisions do not alter the type of testing currently being
performed.
The proposed changes to TS Bases 3/4.3.1 and 3/4.3.2, ``Reactor
Protection System and Safety System Instrumentation,'' and TS Bases
3/4.5.2 and 3/4.5.3, ``ECCS Subsystems,'' are administrative changes
associated with the other proposed changes, and do not alter any
testing currently being performed.
3. Not involve a significant reduction in a margin of safety.
The results of the instrument drift study analysis and review of
historical 18-month surveillance data and applicable maintenance
records support an increase in the surveillance test intervals from
18 to 24 months (and up to 30 months on a non-routine basis)
because: the projected instrument errors caused by drift are bounded
by the existing setpoint analysis or a new analysis has been
performed incorporating a more conservative setpoint; and no
potential for a significant increase in a failure rate of a system
or component was identified during surveillance data and applicable
maintenance records reviews. Existing system and component
redundancy is not affected by these proposed changes.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences,
consequently there are no significant reductions in a margin of
safety.
The NRC staff has reviewed the licensees' 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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Attorney for licensees: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: April 18, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3/4.7.6, ``Plant Systems -
Control Room Emergency Ventilation System.'' Additional Limiting
Conditions for Operation (LCO) would be added related to the
availability of the station vent normal range radiation monitoring
instrumentation. The associated TS bases would also be modified
consistent with these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power Station
(DBNPS), Unit No. 1, in accordance with this change would not:
1a. Involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions, or assumptions are affected by the proposed changes.
The proposed change to LCO 3.7.6.1 would include new required
Action statements in the event that one or both channels of station
vent normal range radiation monitoring instrumentation become
inoperable. In the event that one channel is inoperable for greater
than 7 days, or in the event that both channels are inoperable, the
proposed Action statement would require that the control room normal
ventilation system be isolated and at least one Control Room
Emergency Ventilation System (CREVS) train be placed in operation.
Under the proposed actions, the ventilation systems would be
placed in a state equivalent to that which occurs were a high
radiation isolation to occur. These proposed changes have no bearing
on the probability of an accident.
The proposed change to the terminology utilized in Surveillance
Requirement (SR) 4.7.6.1.e is an administrative change made to make
the terminology consistent with the proposed new Action statements.
The proposed changes to Bases 3/4.7.6 are administrative changes
consistent with the proposed changes to LCO 3.7.6.1. These changes
have no bearing on the probability of an accident.
1b. Involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
change the source term, containment isolation, or allowable
releases.
As described above, under the proposed new LCO 3.7.6.1 Actions,
in the event that one station vent normal range radiation monitoring
instrumentation channel is inoperable for greater than 7 days, or in
the event that both channels are inoperable, the ventilation systems
would be placed in a state equivalent to that which occurs were a
high radiation isolation to occur. Therefore, in the unlikely event
of an accident requiring control room isolation while in this
condition, the dose consequences to control room operators would be
unchanged.
The proposed change to the terminology utilized in SR 4.7.6.1.e
is an administrative change made to make the terminology consistent
with the proposed new Action statements. The proposed changes to
Bases 3/4.7.6 are administrative changes consistent with the
proposed changes to LCO 3.7.6.1. These changes have no bearing on
the consequences of an accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because no new accident
initiators or assumptions are introduced by the proposed changes.
As described above, under the proposed new LCO 3.7.6.1 Actions,
in the event that one station vent normal range radiation
monitoring instrumentation channel is inoperable for greater
than 7 days, or in the event that both channels are inoperable,
theventilation systems would be placed in a state equivalent to that
which occurs were a high radiation isolation to occur. Operation of
the equipment and components in this manner would not introduce the
possibility of any new or different kinds of accidents.
The proposed change to the terminology utilized in SR 4.7.6.1.e
is an administrative change made to make the terminology consistent
with the proposed new Action statements. The proposed changes to
Bases 3/4.7.6 are administrative changes consistent with the
proposed changes to LCO 3.7.6.1. These changes would not introduce
the possibility of any new or different kinds of accidents.
3. Involve a significant reduction in a margin of safety because
the proposed changes to the Action under LCO 3.7.6.1 ensure that
control room isolation capability is maintained in the event a
station vent radiation monitor is inoperable. The proposed allowable
outage time of 7 days for one inoperable channel is consistent with
the presently allowable outage time for one inoperable CREVS. The
proposed Action to place at least one CREVS train in operation
within 1 hour, in the event both channels of radiation monitoring
become inoperable, is more conservative than the present Action
which requires that a plant shutdown commence within 1 hour, but
does not require the CREVS be placed in operation.
The proposed change to the terminology utilized in SR 4.7.6.1.e
is an administrative change made to make the terminology consistent
with the proposed new Action statements. The proposed changes to
Bases 3/4.7.6 are administrative changes consistent with the
proposed changes to LCO 3.7.6.1. These changes would not affect the
margin of safety.The NRC staff has reviewed the licensees' 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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Basis for proposed no significant hazards consideration
determination:
[[Page 30647]]
Attorney for licensees: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: August 22, 1996
Description of amendment request: The proposed change would remove
the action statement of Technical Specification (TS) Section 3.2.G,
Table 3.2.6, Note 7, requiring reactor shutdown after 30 days of
inoperability of the high range stack gas monitor and substitute an
action statement consistent with the guidance provided in NRC Generic
Letter 83-36.
The high range stack monitor provides an estimate of gross stack
activity that has exceeded the upper limit of the normal range
instrumentation. The high range monitor reading serves as input to dose
projection systems for initial estimation of off-site conditions. The
monitor reading would be used prior to the acquisition of stack
isotopic sample data which would provide a more accurate indication of
stack activity.
The licensee stated, among other things, that due to the
passivefunction of the instruments and the ability to monitor this
parameter utilizing alternate methods, it is not appropriate to impose
stringentrequirements on the operation of the unit. This monitor is
identified in the Vermont Yankee Regulatory Guide 1.97 submittal as
Category 2, Type E. This monitor provides post-accident information for
use in determining the magnitude of the release of radioactive
materials and for monitoring such release. However, the high range
stack monitor does not have any safety function associated with the
prevention or automatic mitigation of design-basis accidents, neither
does it provide primary information needed to permit the control room
operating personnel to take required manually controlled actions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91 (a),the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below.
[(1) The proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.]
The High Range Stack Monitor is a RG [Regulatory Guide] 1.97,
Category 2, Type E instrument with no specified safety function
associated with the prevention or automatic mitigation of design
basis accidents, neither does it provide primary information needed
to permit the control room operating personnel to take required
manually controlled actions. The proposed change to the action
statement associated with this monitor will not change the function
of this monitor, and since the monitor is not assumed to initiate
any accidents, nor function to mitigate any accidents, this change
will not significantly increase the probability or consequences of
any previously analyzed accident.
[(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 does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will still ensure effective methods
are available to assess post accident conditions. Thus, this 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 proposed change to the action statement associated with
this monitor will not change the function of this monitor, and since
the monitor is not assumed to function for the prevention or
mitigation of any previously evaluated accidents, this change will
not significantly reduce a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
Attorney for licensee: R. K. Gad, III, Ropes and Gray, One
International Place, Boston, MA 02110-2624
NRC Project Director: Patrick D. Milano, Acting
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of amendment request: January 24, 1997, as supplemented on May
15, 1997 (TSCR 193)
Description of amendment request: The proposed amendments (Point
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request
(TSCR) 193) would revise TS 15.5.4, ``Fuel Storage,'' to increase fuel
assembly enrichment limits to 5.0 w/o U-235 while maintaining
Keff in the storage pools (spent fuel pool and new fuel
storage racks) less than 0.95. The May 15, 1997, supplement provided a
revised no significant hazards consideration determination that
superseded the licensee's determination noticed in the Federal Register
on April 23, 1997 (62 FR 19837).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of this facility under the proposed Technical
Specifications will not create a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve a change to structures,
systems, or components that would affect the probability or
consequences of an accident previously evaluated in the PBNP Final
Safety Analysis Report (FSAR). The only relevant concern with
respect to increasing enrichment limits in the spent fuel pool and
new fuel storage racks is one of criticality. The proposed changes
use the same criticality limit used in the current Technical
Specifications. Therefore, margin to safe operation of Units 1 and 2
is maintained. The probability and consequences of an accident
previously evaluated are dependent on this criticality limit.
Because the limit will not change, the probability and consequences
of those accidents previously evaluated will not change.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve a change to the physical
structure of the spent fuel pool or of the plant. The proposed
increase in spent fuel pool and new fuel storage racks fuel assembly
enrichment limits maintains the margin to safe operation of Units 1
and 2 because the criticality limit for the spent fuel pool and new
fuel storage racks will not change. The enrichment increase does not
affect any of the parameters or conditions that contribute to the
initiation of any accidents. Because the criticality limit remains
the same, these changes have no effect on plant operation or on the
initiation of any accidents. Therefore, the proposed changes will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The proposed changes maintain the margin to safe operation of
Units 1 and 2. The margin of safety is based on the criticality
limit of the spent fuel pool and the new fuel storage racks. Because
this limit will not change, the margin of safety will not be
affected. Therefore, the proposed changes will not create a
significant reduction in a margin of safety.
[[Page 30648]]
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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John N. Hannon
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: April 23, 1997
Description of amendment request: This request proposes to revise
Technical Specification 3/4.9.4, Containment Building Penetrations, and
its associated Bases section, to allow selected containment isolation
valves to be opened under administrative controls during periods of
core alterations or movement of irradiated fuel inside containment.
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 involves changes to the Technical
Specification requirements for containment closure which is an
accident mitigating feature. The changes would not affect the
likelihood of occurrence of any accidents previously evaluated. The
proposed change does not involve any hardware or plant design
changes. The containment leakage value is not assumed to be an
initiator of any analyzed event. Containment isolation valves and
temporary closure devices serve to limit the radiological
consequences of accidents. The proposed change would ensure the
service air and breathing air manual isolation valves will perform
their required containment closure function and will serve to limit
the consequences of a fuel handling accident as described in the
USAR, such that the results of the analyses in the USAR remain
bounding. In considering the consequences of a design basis fuel
handling accident inside containment, the assumptions in the
analysis take no credit for the containment as a barrier to prevent
the postulated release of radioactivity. For events that could occur
during CORE ALTERATIONS or movement of irradiated fuel assemblies,
containment closure is considered a defense-in-depth boundary to
prevent uncontrolled release of radioactivity. Additionally, the
proposed change does not impose any new safety analyses limits or
alter the plant's ability to detect and mitigate events. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change involves reliance on manual actuation of
containment penetration valves (Service Air valves KA V-039 and KA
V-118 and Breathing Air valves KB V-001 and KB V-002 are manual
valves) to block the unimpeded flow of the containment atmosphere to
the environment under certain conditions. The proposed change would
not necessitate a physical alteration of the plant features that
provide core cooling or subcriticality (no new or different type of
equipment will be installed) or changes in parameters governing
plant operation during CORE ALTERATIONS or movement of irradiated
fuel in containment. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is similar to the use of administrative
controls to isolate an open containment airlock door. The use of
administrative controls in this manner has been approved by the NRC
(WCGS Technical Specification Amendment 95) for plant operations
that would not require the containment to maintain a pressure
boundary. This scenario is applicable during plant shutdown for
refueling when CORE ALTERATIONS and movement of irradiated fuel
assemblies in the containment occur. Accidental damage to spent fuel
during these operations is classified as a fuel handling accident.
The proposed change has been developed considering the importance of
the containment boundary in limiting the consequences of a design
basis fuel handling accident. The proposed change allows for
protection equivalent to that provided by previously approved
methods of containment closure. Considering the probability of an
event that would challenge the containment boundary, the alternative
protection provided by this change, and the operational requirements
to occasionally open these penetrations, the proposed change is
acceptable and any reduction in the margin of safety is
insignificant.
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: William H. Bateman
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: January 24, 1997, as
supplemented March 27, 1997
Brief description of amendment: The proposed amendment will update
the
[[Page 30649]]
Safety Limit Minimum Critical Power Ratio (SLMCPR) in Technical
Specification 2.1.2 and the associated Bases section to reflect the
results of the latest cycle-specific calculation performed for the
Pilgrim Nuclear Power Station Operating Cycle 12. In addition, the
values provided in Note 5 of Table 3.2.C.1, which are based on the
SLMCPR values, have been revised as a result of the changes to the
SLMCPR value.
Date of issuance: April 7, 1997
Effective date: April 7, 1997
Amendment No.: 171
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6568) The March 27, 1997, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated April 7, 1997 No
significant hazards consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of application for amendments: August 19, 1996, as
supplemented on February 5, March 13, April 29 and April 30, 1997.
Brief description of amendments: The amendment would revise
Technical Specification (TS) Section 4.4.5.2 to extend, for one
additional operating cycle (i.e., Cycle 7), the 1.0 volt and 3.0 volt
interim plugging criteria (IPC) which were added to the Braidwood, Unit
1, TSs by License Amendment No. 69, issued on November 9, 1995.
Additionally, this amendment to the Braidwood, Unit 1, license added
some definitions and reporting requirements to TS Section 4.4.5.2 and
modified the designations for the IPC models in TS Bases Section 3/
4.4.4.5. Braidwood, Unit 1, Cycle 7, will end in fall 1998. While there
are no revisions to the TS for Braidwood, Unit 2, both units are being
amended to maintain the continuity of the amendment numbers.
Date of issuance: May 14, 1997.
Date of effective: Immediately, to be implemented within 30 days.
Amendment Nos.: 82
Facility Operating License Nos. NPF-72 and NPF-77: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6570). The February 5, March 13, April 29 and April 30, 1997,
submittals provided clarifying technical information that did not
affect the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 14, 1997. No significant
hazards consideration comments received: No
Local Public Document Room location: Wilmington Public Library, 201
S. Kankakee Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: June 20, 1996, as supplemented
December 30, 1996, and March 5, 1997.
Brief description of amendments: The amendments would change the
TSs by incorporating an NRC-approved thermal limit licensing
methodology in the list of approved methodologies used in establishing
the fuel cycle-specific thermal limits. In addition, the proposed
amendments would change the TSs to reflect the use of Siemens Power
Corporation (SPC) ATRIUM-9B fuel for the first time at Dresden, Units 2
or 3. The proposed amendments would also correct minor editorial items
in the TSs.
In March 1997, the NRC staff performed an audit of the application
of Advanced Nuclear Fuel for Boiling Water Reactors (ANFB) to ATRIUM-9
fuel. The staff raised concerns associated with the ATRIUM-9B fuel
additive constant uncertainty used as input to the NRC-approved
methodology for the calculation of minimum critical power ratio (MCPR).
In response to the audit findings, by letter dated April 18, 1997, SPC
submitted a generic topical report (ANF-1125(P) Supplement 1 Appendix
D), which is currently under staff review, for the future reload
analysis in the safety limit MCPR calculation. The staff schedule for
the review of the topical report will not be timely enough for the
resolution of the ATRIUM-9B MCPR issue to support reload and restart of
Dresden, Unit 3. Therefore, by letters dated May 2 and May 6, 1997,
ComEd provided additional information concerning the MCPR issues and
how it will affect the Dresden, Unit 3, D3R15 fuel cycle and provided
additional information concerning the ATRIUM-9B fuel design and
shutdown margin that are applicable during refueling and shutdown.
The staff is currently reviewing the licensee's May 2 and May 6,
1997, letters. To be more timely and support the reload schedule for
Dresden, Unit 3 (currently scheduled for May 20, 1997), the staff has
chosen to split its consideration of the proposed amendments into two
parts. The first part of the amendment package now being evaluated
would modify Section 5.3.A, ``Design Features'' of the TSs to reflect
use of the ATRIUM-9B fuel design and would include two SPC topical
reports in TS Section 6.9.A.6, ``Core Operating Limits Report,'' to
reflect mechanical design criteria for this fuel. This change would
allow this fuel to be loaded into the core only under Operational Modes
3 (Hot Shutdown), 4 (Cold Shutdown), and 5 (Refueling) and does not
permit startup or power operation using the ATRIUM-9B fuel.
Date of issuance: May 16, 1997
Date of effective: Immediately, to be implemented within 30 days.
Amendment Nos.: 159 and 154
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17227). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 16, 1997 No significant
hazards consideration comments received: No
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Commonwealth Edison Company, Docket No. 50-265, Quad Cities Nuclear
Power Station, Unit 2, Rock Island County, Illinois
Date of application for amendment: April 21, 1997
Brief description of amendment: The amendment increases the minimum
critical power ratio safety limit for Unit 2 and adds a Siemens Power
Corporation reference to the Technical Specifications (TS) to allow
plant operation in Operational Modes 1 and 2.
Date of issuance: May 22, 1997
Date of effective: Immediately, to be implemented within 30 days.
Amendment No.: 174
Facility Operating License No. DPR-30: The amendment revised the
TSs. Public comments requested as to proposed no significant hazards
consideration: Yes (62 FR 23499 dated April 30, 1997). This notice
provided an opportunity to submit comments on the Commission's proposed
no significant
[[Page 30650]]
hazards consideration determination. No comments have been received.
The notice also provided for an opportunity to request a hearing by May
30, 1997, but indicated that if the Commission makes a final no
significant hazards consideration determination any such hearing would
take place after issuance of the amendment. The Commission's related
evaluation of the amendment, finding of exigent circumstances, and
final no significant hazards consideration determination are contained
in a Safety Evaluation dated May 22, 1997.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan Date of application for amendment: December 2, 1996 (NRC-
96-0134)
Brief description of amendment: The amendment revises TS 3.1.4.3,
TS Table 3.3.6-1, and TS Table 4.3.6-1 to change the operability
requirements for the Rod Block Monitor (RBM). Specifically, the
revision requires the RBM to be operable when reactor thermal power is
greater than or equal to 30 percent of rated thermal power.
Date of issuance: May 15, 1997
Date of effective: May 15, 1997, with full implementation within 60
days
Amendment No.: 112
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
124) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 15, 1997. No significant hazards
consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
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: April 29, 1997
Brief description of amendments: The amendments incorporate a
license condition that will allow revisions to the Oconee Updated Final
Safety Analysis Report (UFSAR) that clarifies the main turbine-
generated missile protection criteria.
Date of issuance: May 16, 1997
Date of effective: As of the date of issuance and implementation is
the incorporation in the UFSAR the changes described in Duke Power
Company's application dated April 29, 1997
Amendment Nos.: 224, 224, and 221
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the UFSAR and added a new License Condition. Public
comments requested as to proposed no significant hazards consideration:
Yes. (62 FR 24512 dated May 5, 1997). The notice provided an
opportunity to submit comments on the Commission's proposed no
significant hazards consideration determination. No comments have been
received as of the date of issuance. The notice also provided for an
opportunity to request a hearing by June 9, 1997, but indicated that if
the Commission makes a final no significant hazards consideration
determination, any such hearing would take place after issuance of the
amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated May
16, 1997.
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: November 26, 1996, as
supplemented February 12, 1997.
Brief description of amendment: The amendment changes the allowable
primary-to-secondary leak rate and in the Surveillance Requirements
section of the TSs it changes the acceptance criteria for steam
generator tubes. The amendment changes the reference that is included
in the tube acceptance criteria from Combustion Engineering topical
report CEN-601-P Revision 01-P to CEN-630-P, Revision 01.
Date of issuance: May 20, 1997
Date of effective: May 20, 1997, to be implemented within 30 days.
Amendment No.: 184
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64376) The February 12, 1997, submittal provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated May 20, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: April 4, 1995, as supplemented
by letters dated August 25, 1995, and April 18, 1997.
Brief description of amendment: The amendment changes the required
frequency for inspecting reactor coolant pump flywheels.
Date of issuance: May 20, 1997
Date of effective: May 20, 1997, to be implemented within 30 days.
Amendment No.: 185
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995, (60 FR
35069) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 20, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear
One,Unit No. 2, Pope County, Arkansas
Date of application for amendment: October 7, 1996, as supplemented
February 10, and May 8, 1997
Brief description of amendment: The amendment changes the channel
functional testing frequency for most of the Reactor Protection System
(RPS) and Engineered Safety Feature Actuation System (ESFAS)
instrumentation from monthly to every four months. In addition, the
amendment allows the use of Cycle Independent Shape Annealing Matrix
(CISAM) methodology in the Core Protection Calculators (CPCs). Finally,
the amendment makes a number of administrative changes to the Technical
Specifications (TS) to clarify the existing TS or correct previous
errors in the TS.
Date of issuance: May 21, 1997
Date of effective: May 21, 1997
Amendment No.: 186
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4346) The Commission's related evaluation of the amendment is contained
in a Safety
[[Page 30651]]
Evaluation dated May 21, 1997 No significant hazards consideration
comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 27, 1997
Brief description of amendment: The amendment changes TSs
surveillance requirements 4.5.2.d.3 and 4.5.2.d.4 by increasing the
required amount of trisodium phosphate dodecahydrate (TSP) stored in
the containment sump from 97.5 cubic feet to 380 cubic feet, and
adjusts the TSP sampling requirement accordingly.
Date of issuance: May 15, 1997
Date of effective: May 15, 1997, to be implemented within 60 days.
Amendment No.: 127
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17234) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 15, 1997 No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 21, 1996, as supplemented by
letter dated March 17, 1997
Brief description of amendment: The amendment approves revision of
Attachment 1 to the operating license concerning design and testing
modifications in the Containment Vacuum Relief System (CVR) that
penetrate the primary containment at Waterford Steam Electric Station,
Unit 3. The penetrations affected are commonly referred to as
Penetrations 53 and 65.
Date of issuance: May 20, 1997
Date of effective: May 20, 1997, to be implemented within 90 days.
Amendment No.: 128
Facility Operating License No. NPF-38: Amendment revised the
Operating License.
Date of initial notice in Federal Register: November 6, 1996 (61 FR
57484) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 20, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: September 13, 1996, as
supplemented by letter dated January 15, 1997.
Brief description of amendment: The amendment revised the Technical
Specifications to permit the use of 10 CFR Part 50, Appendix J, Option
B, performance-based containment leakage rate testing.
Date of issuance: May 19, 1997
Date of effective: May 19, 1997, to be implemented within 60 days
of the date of issuance.
Amendment No.: 158
Facility Operating License No. DPR-36: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 6, 1996 (61 FR
57487) The January 15, 1997, supplemental letter provided additional
clarifying information and did not change the initial no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
May 19, 1997. No significant hazards consideration comments received:
Yes. Comments were submitted by Patrick J. Dostie on behalf of the
State of Maine by letter dated April 15, 1997. The staff responded by
letter dated May 19, 1997.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: February 18, 1997, as supplemented by
letter dated February 26, 1997.
Description of amendment request: The amendment revises the
Appendix A Technical Specifications relating to the reactor core fuel
assembly design features requirements contained in Technical
Specification 5.3.1, Fuel Assemblies. The changes made by this
amendment allow for the limited replacement of failed or damaged fuel
rods in fuel assemblies with solid stainless steel or zirconium alloy
filler rods.
Date of issuance: May 13, 1997
Date of effective: May 13, 1997
Amendment No.: 51
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 12, 1997 (62 FR
11496) The licensee's letter dated February 26, 1997, provided a
correction to a typographical error in the original application but
does not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated May 13, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833
Portland General Electric Company, et al., Docket No. 50-344,
Trojan Nuclear Plant, Columbia County, Oregon
Date of application for amendment: November 2, 1995.
Brief description of amendment: This amendment changes the TS to
reflect changes in the organization as they apply to oversite and
management of the Trojan Nuclear Plant.
Date of issuance: October 31, 1996
Date of effective: October 31, 1996
Amendment No.: 195
Facility Operating License No. NPF-1: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58404) No significant hazards consideration comments received: No.
Local Public Document Room location: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: January 7, 1997
Brief description of amendments: These amendments revise Technical
Specification (TS) 3/4.2.5 to incorporate an exception to the
provisions of TS 4.0.4 and to clarify the time at which the
surveillance can be performed by adding that the surveillance is to be
performed within 24 hours after attaining steady state conditions at or
above 90% rated thermal power. The revised surveillance contains
editorial enhancements that clarify the
[[Page 30652]]
surveillance requirement. Salem Unit 1 TS Table 3.2-1 is also being
revised to delete reference to three loop operation.
Date of issuance: May 8, 1997
Date of effective: Both units, as of date of issuance, to be
implemented prior to entry into Mode 1 from the current outage.
Amendment Nos. 193 and 176
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4353) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 8, 1997. No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079
Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama
Date of amendment request: March 25, 1997
Brief description of amendments: The amendment changes Technical
Specification 3/4.4.9, ``Specific Activity,'' and the associated Bases
to reduce the limit associated with dose equivalent iodine-131. The
steady-state dose equivalent iodine-131 limit would be reduced by 40
percent from 0.5 [micro]Ci/gram to 0.3 [micro]Ci/gram and the maximum
instantaneous value would be reduced by 40 percent from 30 [micro]Ci/
gram to 18 [micro]Ci/gram.
Date of issuance: May 19, 1997
Date of effective: As of the date of issuance to be implemented
within 30 days
Amendment Nos.: 128
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: April 4, 1997 (62 FR
16201) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 19, 1997. No significant
hazards consideration comments received: No
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Dated at Rockville, Maryland, this 28th day of May, 1997.
For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear
Reactor Regulation
[Doc. 97-14395 Filed 6-3-97; 8:45 am]
BILLING CODE 7590-01-F
