[Federal Register Volume 64, Number 96 (Wednesday, May 19, 1999)]
[Notices]
[Pages 27315-27339]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-12494]
-----------------------------------------------------------------------
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 April 24, 1999, through May 7, 1999. The
last biweekly notice was published on May 5, 1999.
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed no Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By June 18, 1999, 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
[[Page 27316]]
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: March 3, 1999.
Description of amendment request: The proposed amendment would
change the reactor vessel (RV) surveillance capsule pull interval from
approximately 15 effective full power (EFPY) years to 18 EFPY in
Technical Specification (TS) Table 4.6-3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below: The operation of Pilgrim in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
Pilgrim plant's physical configuration and operational practices are
not changed by this proposed change. The licensee is only proposing to
change the TS withdrawal schedule for the RV surveillance capsule. This
change does not affect any of the current accident mitigation features
of the facility or the sequence of any accidents previously analyzed.
For the reasons given above, deferral of withdrawal of Pilgrim's second
capsule for at least one additional cycle (or 3 EFPY) does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The operation of Pilgrim in accordance with the proposed amendment
will not create the possibility of a new or different kind of accident
from any accident previously evaluated. As discussed in the above
narrative, the deferral of the second capsule pull at Pilgrim does not
change any of the design features or operation of the facility but does
defer a TS surveillance. Pilgrim's current TS pressure-temperature (P-
T) curves are conservative and will remain so even if the RV
surveillance capsule is not pulled this outage. The data from the first
RV capsule supports this conclusion. Because the RV capsule pull
schedule is being deferred, the P-T curves, which can be modified based
on the data from the RV capsule surveillance, will not be changed. The
deferral of the withdrawal of Pilgrim's second RV surveillance capsule
does not change the design features or operation of the facility and
the existing P-T curves have not changed, therefore, the TS change will
not create the possibility of a new or different kind of accident from
any accident previously evaluated.
The operation of Pilgrim in accordance with the proposed amendment
will not involve a significant reduction in the margin of safety.
The capsule pull is a surveillance technique that provides data for
modification of the P-T curves. The methods used to develop the
temperatures associated with these curves are regarded as conservative.
The data from the first RV capsule supported this conclusion. Because
the P-T curves have not changed and have been determined to be
conservative, the margins of safety that were previously established
have not changed. Therefore, deferral of the withdrawal of Pilgrim's
second RV surveillance capsule will not involve a significant reduction
in the margin of safety.
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
[[Page 27317]]
amendment request involves no significant hazards consideration.
Local Public Document Room location: Plymouth Public Library, 132
South Street, Plymouth, Massachusetts 02360.
Attorney for licensee: J. Fulton, Boston Edison Company, 800
Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Section Chief: James W. Clifford.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: March 30, 1999.
Description of amendment request: The proposed amendment would
revise Section 4.0, Surveillance Requirements, of the Technical
Specifications (TSs). Specifically, Section 4.0.2 would be added to
allow a 24-hour grace period for performing inadvertently missed
surveillance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No. This proposed change will result in either the
plant condition either remaining unchanged (i.e., the system or
component is declared operable) or in the plant proceeding to a
shutdown condition (i.e., the system or component is declared
operable). If at the end of the 24-hour interval, it is necessary to
proceed to shutdown, this shutdown is indistinguishable from any
shutdown where a system or component is declared inoperable.
Allowing an additional 24 hours to perform the surveillance balances
the risks associated with an allowance for completing the
surveillance within this 24-hour period against the risks associated
with the potential for a plant upset and challenge to safety systems
when the alternative is a shutdown to comply with the action
requirements before the surveillance can be completed. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No. This proposed change will result in either the
plant condition either remaining unchanged (i.e., the system or
component is declared operable) or in the plant proceeding to a
shutdown condition (i.e., the system or component is declared
operable). If at the end of the 24-hour interval, it is necessary to
proceed to shutdown, this shutdown is indistinguishable from any
shutdown where a system or component is declared inoperable.
Allowing an additional 24 hours to perform the surveillance balances
the risks associated with an allowance for completing the
surveillance within this 24-hour period against the risks associated
with the potential for a plant upset and challenge to safety systems
when the alternative is a shutdown to comply with the action
requirements before the surveillance can be completed. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No. This proposed change will result in either the
plant condition either remaining unchanged (i.e., the system or
component is declared operable) or in the plant proceeding to a
shutdown condition (i.e., the system or component is declared
operable). If at the end of the 24-hour interval, it is necessary to
proceed to shutdown, this shutdown is indistinguishable from any
shutdown where a system or component is declared inoperable.
Allowing an additional 24 hours to perform the surveillance within
this 24-hour period against the risks associated with the potential
for a plant upset and challenge to safety systems when the
alternative is a shutdown to comply with the action requirements
before the surveillance can be completed . Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Section Chief: S. Singh Bajwa.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: July 22 and October 22, 1998; May 6,
1999.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) to reflect the licensee's planned use of
fuel supplied by Westinghouse. The staff has published a Notice of
Consideration of Issuance of Amendments and Proposed No Significant
Hazards Consideration Determination on November 18, 1998 (63 FR 64108)
covering the July 22 and October 22, 1998, submittals. In the May 6,
1999, submittal the licensee proposed to expand the original amendment
request, revising Section 5.6.5 of the Technical Specifications.
Section 5.6.5 specifies a list of NRC-approved topical reports that the
licensee is required to use to determine reactor core operating limits.
The licensee proposed to update this list to show the current approval
status of these topical reports.
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 for the proposed changes conveyed by the May 6, 1999,
submittal. The NRC staff has reviewed the licensee's analyses against
the standards of 10 CFR 50.92(c). The NRC-staff's analysis is presented
below.
First Standard
No. The proposed changes to Section 5.6.5 will not affect the
safety function and will not involve any change to the design or
operation of any plant system or component. The topical reports were
previously approved by the NRC staff under separate licensing actions.
The use of methodologies in these approved topical reports will ensure
that previously evaluated accidents remain bounding. Therefore, no
accident probabilities or consequences will be impacted.
Second Standard
No. The proposed changes would not lead to any hardware or
operating procedure change. Hence, no new equipment failure modes or
accidents from those previously evaluated will be created.
Third Standard
No. Margin of safety is associated with confidence in the design
and operation of the plant; specifically, the ability of the fission
product barriers to perform their design functions during and following
an accident. The proposed changes to Section 5.6.5 do not involve any
change to plant design, operation, or analysis. Thus, the margin of
safety previously analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for the proposed changes to Section 5.6.5.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
[[Page 27318]]
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: April 5, 1999.
Description of amendment request: The proposed amendments would
provide revised spent fuel pool storage configurations, revised spent
fuel pool storage criteria, and revised fuel enrichment and burnup
requirements which take credit for soluble boron in maintaining
acceptable margins of subcriticality in the spent fuel storage pools.
Also, the proposed amendments would provide additional criteria for
ensuring acceptable levels of subcriticality in the spent fuel storage
pools.
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 the change involve a significant increase in the
probability or consequence of an accident previously evaluated?
No, based upon the following:
Dropped Fuel Assembly
There is no significant increase in the probability of a fuel
assembly drop accident in the spent fuel pools when considering the
degradation of the Boraflex panels in the spent fuel pool racks
coupled with the presence of soluble boron in the spent fuel pool
water for criticality control. The handling of the fuel assemblies
in the spent fuel pool has always been performed in borated water,
and the quantity of Boraflex remaining in the racks has no affect on
the probability of such a drop accident.
The criticality analysis showed that the consequences of a fuel
assembly drop accident in the spent fuel pools are not affected when
considering the degradation of the Boraflex in the spent fuel pool
racks and the presence of soluble boron.
Fuel Misloading
There is no significant increase in the probability of the
accidental misloading of spent fuel assemblies into the spent fuel
pool racks when considering the degradation of the Boraflex in the
spent fuel pool racks and the presence of soluble boron in the pool
water for criticality control. Fuel assembly placement and storage
will continue to be controlled pursuant to approved fuel handling
procedures to ensure compliance with the Technical Specification
requirements. These procedures will be revised as needed to comply
with the revised requirements which would be imposed by the proposed
Technical Specification changes.
There is no increase in the consequences of the accidental
misloading of spent fuel assemblies into the spent fuel pool racks
because criticality analyses demonstrate that the pool will remain
subcritical following an accidental misloading if the pool contains
an adequate boron concentration. Current Technical Specification
3.7.14 will ensure that an adequate spent fuel pool boron
concentration is maintained in the McGuire spent fuel storage pools.
A McGuire Station UFSAR change will revise Chapter 16, ``Selected
Licensee Commitments'', to provide for adequate monitoring of the
remaining Boraflex in the spent fuel pool racks. If that monitoring
identifies further reductions in the Boraflex panels which would not
support the conclusions of the McGuire Criticality Analysis, then
the McGuire TS's and design bases would be revised as needed to
ensure that acceptable subcriticality are maintained in the McGuire
spent fuel storage pools.
Significant Change in Spent Fuel Pool Temperature
There is no significant increase in the probability of either
the loss of normal cooling to the spent fuel pool water or a
decrease in pool water temperature from a large emergency makeup
when considering the degradation of the Boraflex in the spent fuel
pool racks and the presence of soluble boron in the pool water for
subcriticality control since a high concentration of soluble boron
has always been maintained in the spent fuel pool water. Current
Technical Specification 3.7.14 will ensure that an adequate spent
fuel pool boron concentration is maintained in the McGuire spent
fuel storage pools.
A loss of normal cooling to the spent fuel pool water causes an
increase in the temperature of the water passing through the stored
fuel assemblies. This causes a decrease in water density that would
result in a decrease in reactivity when Boraflex neutron absorber
panels are present in the racks. However, since a reduction in the
amount of Boraflex present in the racks is considered, and the spent
fuel pool water has a high concentration of boron, a density
decrease causes a positive reactivity addition. However, the
additional negative reactivity provided by the current boron
concentration limit, above that provided by the concentration
required to maintain keff less than or equal to 0.95
(1170 ppm), will compensate for the increased reactivity which could
result from a loss of spent fuel pool cooling event. Because
adequate soluble boron will be maintained in the spent fuel pool
water, the consequences of a loss of normal cooling to the spent
fuel pool will not be increased. Current Technical Specification
3.7.14 will ensure that an adequate spent fuel pool boron
concentration is maintained in the McGuire spent fuel storage pools.
A decrease in pool water temperature from a large emergency
makeup causes an increase in water density that would result in an
increase in reactivity when Boraflex neutron absorber panels are
present in the racks. However, the additional negative reactivity
provided by the current boron concentration limit, above that
provided by the concentration required to maintain keff
less than or equal to 0.95 (1170 ppm), will compensate for the
increased reactivity which could result from a decrease in spent
fuel pool water temperature. Because adequate soluble boron will be
maintained in the spent fuel pool water, the consequences of a
decrease in pool water temperature will not be increased. Current
Technical Specification 3.7.14 will ensure that an adequate spent
fuel pool boron concentration is maintained in the McGuire spent
fuel storage pools.
2. Will the change create the possibility of a new or different
kind of accident from any previously evaluated?
No. Criticality accidents in the spent fuel pool are not new or
different types of accidents. They have been analyzed in Section
9.1.2.3 of the Updated Final Safety Analysis Report and in
Criticality Analysis reports associated with specific licensing
amendments for fuel enrichments up to 4.75 weight percent U-235.
Specific accidents considered and evaluated include fuel assembly
drop, accidental misloading of spent fuel assemblies into the spent
fuel pool racks, and significant changes in spent fuel pool water
temperature. The accident analysis in the Updated Final Safety
Analysis Report remains bounding.
The possibility for creating a new or different kind of accident
is not credible. The amendment proposes to take credit for the
soluble boron in the spent fuel pool water for reactivity control in
the spent fuel pool while maintaining the necessary margin of
safety. Because soluble boron has always been present in the spent
fuel pool, a dilution of the spent fuel pool soluble boron has
always been a possibility, however this accident was not considered
credible. For the proposed amendment, the spent fuel pool dilution
evaluation (Attachment 7) demonstrates that a dilution of the boron
concentration in the spent fuel pool water which could increase the
rack keff to greater than 0.95 (constituting a reduction
of the required margin to criticality) is not a credible event. The
requirement to maintain boron concentration in the spent fuel pool
water for reactivity control will have no effect on normal pool
operations and maintenance. There are no changes in equipment design
or in plant configuration. This new requirement will not result in
the installation of any new equipment or modification of any
existing equipment. Therefore, the proposed amendment will not
result in the possibility of a new or different kind of accident.
3. Will the change involve a significant reduction in a margin
of safety?
No. The proposed Technical Specification changes and the
resulting spent fuel storage operating limits will provide adequate
safety margin to ensure that the stored fuel assembly array will
always remain subcritical. Those limits are based on a plant
[[Page 27319]]
specific criticality analysis (Attachment 6) based on the
``Westinghouse Spent Fuel Rack Criticality Analysis Methodology''
described in Reference 1. The Westinghouse methodology for taking
credit for soluble boron in the spent fuel pool has been reviewed
and approved by the NRC (Reference 6). This methodology takes
partial credit for soluble boron in the spent fuel pool and requires
conformance with the following NRC Acceptance criteria for
preventing criticality outside the reactor:
(1) keff shall be less than 1.0 if fully flooded with
unborated water which includes an allowance for uncertainties at a
95% probability, 95% confidence (95/95) level; and
(2) keff shall be less than or equal to 0.95 if fully
flooded with borated water, which includes an allowance for
uncertainties at a 95/95 level.
The criticality analysis utilized credit for soluble boron to
ensure keff will be less than or equal to 0.95 under
normal circumstances, and storage configurations have been defined
using a 95/95 keff calculation to ensure that the spent
fuel rack keff will be less than 1.0 with no soluble
boron. Soluble boron credit is used to provide safety margin by
maintaining keff less than or equal to 0.95 including
uncertainties, tolerances and accident conditions in the presence of
spent fuel pool soluble boron. The loss of substantial amounts of
soluble boron from the spent fuel pool which could lead to exceeding
a keff of 0.95 has been evaluated (Attachment 7) and
shown to be not credible. Accordingly, the required margin to
criticality is not reduced.
The evaluations in Attachment 7, which show that the dilution of
the spent fuel pool boron concentration from the conservative
assumed initial boron concentration (2475 ppm) to the minimum boron
concentration required to maintain keff [less than or
equal to] 0.95 (440 ppm) is not credible, combined with the 95/95
calculation which shows that the spent fuel rack keff
will remain less than 1.0 when flooded with unborated water, provide
a level of safety comparable to the conservative criticality
analysis methodology required by References 2, 3 and 4.
Therefore the proposed changes in this license amendment will
not result in a significant reduction in the plant's margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: J. Murray Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: April 6, 1999.
Description of amendment request: The proposed amendments would
expand the allowable values for Interlocks P-6 (Intermediate Range
Neutron Flux) and P-10 (Power Range Neutron Flux) in TS 3.3.1, Table
3.3.1-1, Function 16, Reactor Trip System Interlocks, as recommended by
Westinghouse.
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; 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.
Criterion 1--Would operation of the facility in accordance with
the requested amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
The reactor protection interlocks are provided to ensure reactor
trips are in the correct configuration for the current unit status.
They back up operator actions to ensure protection system functions
are not bypassed during unit conditions under which the safety
analysis assumes the functions are not bypassed. The proposed
changes involve changing the lower value of the P-10 permissive
(power range (PR) neutron flux) allowable values from [greater than
or equal to] 9% RTP to [greater than or equal to] 7% RTP, and
changing the P-6 permissive (intermediate range (IR) neutron flux)
allowable value from [greater than or equal to] 6E11 amp to [greater
than or equal to] 4E-11 amp. Changing the P-10 allowable value would
allow for tripping and resetting of the permissive at a lower
reactor power level. Changing the P-6 allowable value would allow
the source range (SR) channels to be blocked at a lower increasing
reactor power level and delay resetting of the permissive at a lower
decreasing reactor power level.
A review of the UFSAR Chapter 15 accident analyses determined
that no credit is taken for the SR reactor trip or the IR reactor
trip for any of the UFSAR accidents. Credit is taken for the PR low
setpoint trip for a feedwater system malfunction causing an increase
in feedwater flow accident (15.1.2), uncontrolled rod cluster
control assembly bank withdrawal from a subcritical or low power
startup condition accident (15.4.1), and spectrum of rod cluster
control assembly ejection accidents (15.4.8). All three of these
accident scenarios are bounded by cases at 0% RTP taking credit for
the PR low setpoint trip and cases at [greater than or equal to] 10%
RTP taking credit for the PR high setpoint trip. The uncontrolled
rod cluster control assembly bank withdrawal from power accident
(15.4.2) analyses are performed at initial power levels of 10%, 50%,
and 100% RTP to demonstrate that acceptable results are obtained for
a range of initial power levels. For this accident, the PR neutron
flux high setpoint trip, high pressurizer pressure trip, overpower
delta-T (OPDT) trip and overtemperature delta-T (OTDT) trip provide
core protection. With the P-10 reset function changed to as low as
7% RTP, the conclusions of Section 15.4.2 analysis would not change.
Since the uncontrolled bank withdrawal event is analyzed from both
zero power and 10% RTP, all low power initial conditions are
adequately bounded. Therefore, the proposed changes will not
increase the probability or consequences of an accident previously
evaluated.
Criterion 2--Would operation of the facility in accordance with
the requested amendment create the possibility of a new or different
kind of accident from any previously evaluated?
The proposed changes to the allowable values will provide
adequate deadbands between the trip and reset setpoints as well as
adequate margin for instrument drift. The reactor trip system
overpower trips continue to perform their safety function as assumed
in safety analyses. Only the permissives (P-6 and P-10) for blocking
and unblocking of overpower reactor trips are changed. The proposed
changes will not invalidate any of the UFSAR accident analyses. The
proposed changes will not introduce any new failure modes.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any previously evaluated.
Criterion 3--Would operation of the facility in accordance with
the requested amendment involve a significant reduction in a margin
of safety?
The proposed changes involve lowering the Technical
Specification allowable values associated with the P-10 and P-6
permissives for blocking and unblocking of reactor overpower trips.
The lowering of these allowable values is not considered a
significant reduction since it is just enough to accommodate a
deadband recommended by Westinghouse and a margin for instrument
drift. The proposed changes will not invalidate any UFSAR Chapter 15
accident analyses. Therefore, the proposed changes will not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: J. Murrey Atkins Library,
[[Page 27320]]
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: April 26, 1999.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to provide a method for obtaining a
Nuclear Regulatory Commission review of (a) the analytical details
regarding a revised methodology for determining steam generator tube
loads following a main steam line break, and (b) the crediting of the
main steam line break detection and feedwater isolation instrumentation
as a means for providing runout protection for the turbine-driven
emergency feedwater pump.
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?
No. The proposed changes involve: (a) revising the methodology
utilized to determine steam generator tube loads following a main
steam line break (MSLB); and (b) utilizing the MSLB detection and
feedwater isolation instrumentation as an additional means of
providing runout protection of the turbine-driven emergency
feedwater (EFW) pump.
The revised methodology utilized to determine steam generator
tube loads following a MSLB is consistent with the methodology
utilized in the MSLB containment response analysis which has
received Nuclear Regulatory Commission (NRC) approval. The revised
MSLB analysis reaches the same conclusion as the original analysis
(i.e., steam generator tube integrity is maintained). The new
analysis takes into consideration the operation of the MSLB
detection and feedwater isolation instrumentation to terminate main
feedwater (MFW) flow and inhibit the auto-start of or auto-stop the
turbine-driven EFW pump. This instrumentation is QA-1, whereas the
Integrated Control System (ICS) is non-safety. Furthermore, the
revised MSLB analysis results in a greater temperature difference
between the steam generator tube and shell, thus, more conservative
steam generator tube loads than those identified in the original
MSLB analysis.
Also, in the event that the MSLB detection and feedwater
isolation instrumentation does not function properly, the non-safety
ICS is still available to maintain steam generator water level at
the post-trip minimum level as assumed in the original analysis.
Currently, operator action is the only credited means to protect
the turbine-driven RFW pump from runout. The MSLB detection and
feedwater isolation instrumentation provides an additional method to
protect the turbine-driven EFW pump from runout. Crediting the MSLB
detection and feedwater isolation instrumentation simply adds
defense in depth.
There are no physical changes to the plant structures, systems,
or components (SSCs) or operating procedures, nor are there any
changes to safety limits or set points. Also, no new radiological
release pathways are created.
Thus, the proposed change does not significantly increase the
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from the accidents previously evaluated?
No. The reanalysis of the steam generator tube loads following a
MSLB accident is limited to an accident that is already evaluated in
the UFSAR. The methodology is similar to the current analysis for
the MSLB containment response. The effects of the MSLB on steam
generator tube integrity are the same as in the original analysis--
tube integrity is maintained.
The revised analysis takes into consideration the operation of
the MSLB detection and feedwater isolation instrumentation, which
terminates MFW flow and inhibits the auto-start of or auto-stops the
turbine-driven EFW pump following a MSLB. As assumed in the original
analysis, the non-safety ICS will remain available to control steam
generator water level at the post-trip minimum level should a
malfunction occur in the MSLB detection and mitigation circuit.
Should this malfunction occur, the resulting tube stresses would
decrease relative to the revised analysis.
Crediting the MSLB detection and feedwater isolation
instrumentation as a means to protect the turbine-driven EFW pump
from runout simply adds defense in depth.
There are no physical changes to the plant SSCs or operating
procedures. There are no new hazardous materials or potential
missiles. It does not introduce the possibility of any new or
different malfunctions. No safety limits or set points are changed.
Thus, the proposed change does not create the possibility of a
new or different kind of accident.
3. Involve a significant reduction in a margin of safety?
No. The reanalysis of the steam generator tube loads following a
MSLB accident is similar to the current analysis for the previously
NRC approved MSLB containment response. The conclusion of the
revised MSLB steam generator tube load analysis is the same as the
conclusion in the original analysis--steam generator tube integrity
is maintained.
Crediting the MSLB detection and feedwater isolation
instrumentation as a means to protect the turbine-driven EFW pump
from runout simply adds defense in depth.
There are no safety limit, set point, design parameters, or
operating procedure changes required. The integrity of the fuel
cladding, reactor coolant system, and containment are preserved.
Thus, the proposed change does not involve a significant
reduction in a margin of safety.
Duke has concluded based on the above information that there are
no significant hazards involved in this LAR.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: April 9, 1999.
Description of amendment request: The proposed amendment would
revise the requirements affecting the surveillance methods for the
containment tendons, the conduct of containment visual inspections, and
the reporting methods employed in disseminating the results of these
inspections to the Nuclear Regulatory Commission.
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 proposed change to the ANO-1 [Arkansas Nuclear One, Unit 1]
TS [Technical Specifications] replaces previous requirements and
commitments to establish a containment inspection program based on
the guidance provided in Regulatory Guide 1.35, Revision 2 in favor
of regulations depicted in [Title] 10 [of the] CFR [Code of
[[Page 27321]]
Federal Regulations] 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix). ANO-1
is implementing a containment inspection program to comply with
these new regulatory requirements. The final rule specifies
requirements to assure that the critical areas of the containment
structure are routinely inspected to detect and take corrective
action for defects that could compromise structural integrity.
Maintaining reactor building structural integrity is independent
of the operation of the reactor coolant system (RCS), the reactor
protection system (RPS) and emergency core cooling system (ECCS).
The reactor building is not considered to be the initiator of any
accident previously evaluated. The physical location of inspection
details does not prevent or inhibit the reactor building from
functioning as designed to provide an acceptable barrier against
release of radioactive materials to the environment. Through
appropriate inspections and implementation of corrective actions for
any degradation discovered during the inspections that might lead to
containment structural failures, the probability or consequences of
accidents will not be increased.
Therefore, the removal of inspection details from the TS does
not involve a significant increase in the probability or
consequences of any accident previously evaluated.
Criterion 2--Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
Maintaining containment structural integrity is independent of
the operation of the RCS, the RPS and ECCS. The proposed changes do
not change the design, configuration, or method of operation of the
plant. By implementing corrective actions for any degradation
discovered during the required inspections of the containment, the
possibility of a new or different kind of accident will not be
created. Implementation of the requirements of Subsection IWL of the
ASME [American Society of Mechanical Engineers] code and those of 10
CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix) provide an equally
acceptable containment inspection program.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the
Margin of Safety.
The removal of the level of detail currently found in the ANO-1
TS regarding reactor building inspections and incorporating the
applicable requirements of Subsection IWL of the ASME code and of 10
CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix) into the ANO-1
containment inspection program has no impact on any safety analysis
assumptions. Requirements associated with containment inspections
are controlled by safety related procedure 5220.011. Sufficient
controls exist under the procedure change process at ANO-1 to ensure
current and future regulations and commitments are properly
addressed when making revisions to the containment inspection
procedure. The addition of structural integrity requirements to ANO-
1 TS Specification 3.6.1 imposes consistent requirements with those
previously specified in the ANO-1 TSs. The containment inspection
program ensures that the containment will function as designed to
provide an acceptable barrier against release of radioactive
materials to the environment. Through the implementation of the
containment inspection program, the existing margin of safety is
preserved.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: April 9, 1999.
Description of amendment request: The proposed amendment would
revise the requirements associated with the station batteries and the
direct current (dc) sources to the 125 volt dc switchyard distribution
system.
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 switchyard 125V DC control power source requirements do not
meet the criteria for inclusion in Technical Specifications (TSs) as
evaluated with respect to the selection criteria of [Title] 10 [of
the] CFR [Code of Federal Regulations] 50.36. These control power
sources are not assumed to mitigate accident or transient events.
The effects of a loss of these control power sources are enveloped
by the Loss of Offsite Power (LOOP) event and relocation is
considered to have a non-significant impact on the probability or
severity of a LOOP event. These requirements will be relocated from
the TSs to an appropriate administratively controlled document and
maintained pursuant to 10 CFR 50.59.
Proposed changes incorporating the requirements of TS 3.7.1.D,
3.7.2.E, 3.7.2.F, and 3.7.2.A, as related to the DC electrical power
subsystems, in the new TS 3.7.3 results in a more stringent
requirement for the ANO-1 [Arkansas Nuclear One, Unit 1] TSs in that
reductions to lower conditions of operation in shorter periods of
time are now required. These more stringent requirements are not
assumed to be initiators of any analyzed events and will not alter
assumptions relative to mitigation of accident or transient events.
The proposed addition of TS 3.7.4 allowing continued operation
for a limited period of time with battery cell parameters not within
limits under certain conditions clarifies an allowance that
currently exists in the ANO-1 TS due to the absence of acceptance
criteria for the battery cell parameter surveillances.
Proposed changes in Surveillance Requirements and Frequencies
reflect current industry guidance on maintenance and testing of the
station batteries. These requirements, in themselves, are not
considered to be initiators of any analyzed accident condition.
Although some frequencies have been extended, continued performance
of maintenance activities in accordance with IEEE-450 [Institute of
Electrical and Electronic Engineers, ``Recommended Practice for
Maintenance Testing and Replacement of Vented Lead-Acid Batteries
for Stationary Applications], in addition to the required
Surveillance Requirements, ensures that corrective maintenance can
be performed prior to a condition challenging an operability limit.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed changes revise the surveillance requirements, and
required actions associated with the 125VDC distribution system and
the battery cell parameters. The requirements associated with the
ANO-1 switchyard DC sources have been relocated to licensee control.
The proposed changes do not change the design, configuration, or
method of operation of the plant.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the
Margin of Safety.
Relocation of the switchyard 125V DC control power source
requirements has no impact on any safety analysis assumptions. In
addition, the requirements associated with these control power
sources are relocated to an owner controlled document for which
future changes will be evaluated pursuant to the requirements of 10
CFR 50.59.
Proposed changes incorporating the requirements of TS 3.7.1.D,
3.7.2.E, 3.7.2.F, and 3.7.2.A, as related to the DC electrical power
subsystems, in the new TS 3.7.3 impose more stringent requirements
than previously specified for ANO-1.
The proposed addition of TS 3.7.4 allowing continued operation
for a limited period of time with battery cell parameters not within
limits under certain conditions clarifies an allowance that
currently exists in the ANO-1 TS due to the absence of acceptance
criteria for the battery cell parameter surveillances.
[[Page 27322]]
Proposed changes in Surveillance Requirements and Frequencies
reflect current industry guidance on maintenance and testing of the
station batteries. Although some frequencies have been extended,
continued performance of maintenance activities in accordance with
IEEE-450, in addition to the required Surveillance Requirements,
ensures that corrective maintenance can be performed prior to a
condition challenging an operability limit.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: March 17, 1999.
Description of amendment request: The proposed amendment changes
the Perry Nuclear Power Plant as described in the Updated Safety
Analysis Report. The change incorporates a leak-off line in the
residual heat removal system. The leak-off line is designed to
eliminate an operator work around, which will significantly reduce the
collective dose to plant operations personnel.
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 modification has been described, and will be
procured and installed in accordance with the original design codes
and standards. The safety functions of the RHR [residual heat
removal] system have not been impacted by the change. Systems
supporting the operation of the RHR system have not been affected by
this modification. Though the modification affects the Containment
System, the containment remains capable of performing its associated
safety functions to the same level as the original design.
The accidents of concern are the Loss-Of-Coolant (LOCA) and the
Loss of Shutdown Cooling. The proposed change has been designed in
accordance with the original codes and standards. The proposed
change will not alter the operation of any plant equipment assumed
to function in response to the aforementioned analyzed events or
otherwise increase their failure probability. Therefore, the
probability of occurrence or the consequences of an accident
previously evaluated remains unchanged.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed modification has been designed, and will be
procured and installed in accordance with the original RHR system
design codes and standards. RHR system functions have not been
impacted by the change. Systems supporting the operation of the RHR
system have not been affected. Failure of the modification to
perform its design function due to leak-off line failure or blockage
would be identical to the current RHR system performance. Improper
operation of the valves associated with the modification have been
evaluated and will not prevent or otherwise inhibit the RHR or
Containment systems from performing their applicable safety
functions.
Missile generation is not a concern since no mechanisms
conducive to missile generation have been introduced. Electrical
analyses have shown there is no adverse effect upon the diesel
generator loadings. A single failure of the new configuration will
not result in more than the loss of a single RHR loop which is
already analyzed. Therefore, the possibility of a new or different
kind of accident from any previously evaluated has not been created.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The proposed modification has been designed, and will be
procured and installed in accordance with the original RHR system
design codes and standards. The RHR and Containment systems remain
capable of performing their safety functions. Systems supporting the
operation of the RHR system have not been affected. Hence, the RHR
system margin of safety with respect to safety classification,
protection, redundancy, and seismic classification remains
unaffected.
The margins of safety contained in the Technical Specifications
and the associated Bases also remain unaffected by this
modification. Specifically, Technical Specifications 3.4.6,
``Reactor Coolant System Pressure Isolation Valve Leakage'; 3.4.9,
``RHR Shutdown Cooling System--Hot Shutdown'; 3.4.10, ``RHR Shutdown
Cooling System--Cold Shutdown'; 3.6.2.1, ``Suppression Pool Average
Temperature'; and 3.6.2.2, ``Suppression Pool Water Level'; and the
associated Bases remain unchanged and fully applicable. Hence, the
margins of safety defined in the Technical Specifications remains
unaffected.
Therefore, the proposed modification 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: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Anthony J. Mendiola.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: March 31, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to (1) increase the minimum reactor
coolant system (RCS) flow rate limit, (2) delete the reactor coolant
flow rate footnote, and (3) change the minimum frequency surveillance
for RCS flow rate.
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.
Combustion Engineering (ABB/CE) in Thermal-Hydraulic Report CR-
94-19-CSE95-1131, Revision 0 performed a comprehensive evaluation of
the effects the removal of the orifice plates would have on steam
generator tube degradation. It was concluded that the removal of the
orifice plates would increase the primary flow rate by approximately
5%.
The removal of the orifice plates was estimated to increase the
probability of tubes requiring repair over the lifetime of the
plant. However, the presence of the orifice plates had prevented
inspection of approximately 22% of the steam generator tubes for
circumferential cracks on the hot-leg side. Therefore, it was
concluded that the removal of the orifice plates did not increase
the probability of steam generator tube failure, given that the
tubes previously covered by the plates are now inspected each outage
in accordance with the Electrical Power Research Institute
Pressurized Water Reactor (EPRI PWR) steam generator examination
guidelines. Fort Calhoun Station is using the eddy current
inspection technology to ensure that tubes showing evidence of a
crack exceeding the present plugging criteria will be repaired or
removed from service. Industry experience has shown that even in
cases of severely degraded tubes, the
[[Page 27323]]
resulting primary to secondary leak rates are insignificant compared
to those analyzed in the design basis steam generator tube rupture
event.
Calculation of the Reactor Coolant Flow Rate using the heat
balance methodology once every refueling outage is consistent with
requirements contained in the NUREG 1432, Improved Technical
Specifications for Combustion Engineering Plants' surveillance
requirement 3.4.1.4.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The original orifice plates were installed on each steam
generator hot leg tube sheet in the primary inlet plenum as a field
modification prior to the initial fuel load in the year 1973. The
orifice plates were designed to increase the hydraulic resistance of
the primary coolant flow rate in the associated tubes, thereby
reducing the primary coolant temperature inside the tubes. Reduction
of the primary coolant temperature and flow rate would decrease the
heat flux, thus improving the steam quality and reducing the
potential for dry-out and surface deposits on the outer surface of
the tubes. However, due to inaccessibility, these originally
installed orifice plates had prevented tube inspection in the hot
leg tube sheet area, even with the latest state-of-the-art eddy
current probe technology. The orifice plates also prevented normal
repair techniques such as steam generator tube plugging and
sleeving.
The original orifice plates were removed during the 1996
refueling outage. However, there were concerns related to
Westinghouse fuel failures as a result of flow-induced vibration. To
address those concerns, new ``removable'' orifice plates were
installed to maintain the RCS flow rate at the previous level. Since
then, the remaining batches of the Westinghouse fuel considered most
susceptible to flow-induced vibration were replaced during the 1998
refueling outage, thus minimizing the concerns and allowing the
permanent removal of the ``removable'' orifice plates.
The removal of the ``removable'' orifice plates returned the
steam generators to their original design configuration. RCS flow
rate has increased by virtue of decreased hydraulic resistance
through the steam generators. No other systems or components other
than the steam generators have been affected. The resulting change
in operational parameters (decreased reactor coolant Thot
temperature and increased flow rate) has been evaluated for the
Updated Safety Analysis Report Chapter 14. Potential adverse
consequences of the modifications were (1) increase in reactor
vessel component vibration, (2) increase in hydraulic loading, and
(3) increase in steam generator tube degradation for row 1-18 tubes.
The potential adverse consequences were evaluated and found to be
acceptable.
Calculation of the Reactor Coolant Flow Rate using the heat
balance methodology once every refueling outage is consistent with
requirements contained in the NUREG 1432, Improved Technical
Specifications for Combustion Engineering Plants' surveillance
requirement 3.4.1.4.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The removal of the orifice plates has resulted in approximately
a 5% increase in the reactor coolant flow rate. This has increased
the margin for minimum reactor coolant system flow rate specified in
Technical Specifications Section 2.10.4, Power Distribution Limits,
Item (5), DNBR Margin During Power Operation Above 15% of Rated
Power. Steam Generator tube inspections performed in accordance with
Technical Specifications Section 3.17, Steam Generator Tubes, have
not been adversely affected.
The increased flow rate has been analyzed for the thermal
hydraulic effects on the reactor core and was found acceptable.
Calculation of the Reactor Coolant Flow Rate using the heat
balance methodology once every refueling outage is consistent with
requirements contained in the NUREG 1432 [Improved Technical
Specifications for Combustion Engineering Plants] surveillance
requirement 3.4.1.4.
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: Stuart A. Richards.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: January 28, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to
remove two lists of Containment Isolation Valves (CIVs) in Tables 3.6-1
and 4.4-1 and make related changes to TSs 1.10, 3.6.A.1, and 4.4 and
the associated bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No. Operation of Indian Point 3 in accordance with the
proposed license amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The removal of the two component listings (i.e., Tables
3.6-1 and 4.4-1) and the TS references to them from the TS requested
by this submittal is performed in accordance with the guidance
provided by the NRC in GL 91-08 [Generic Letter 91-08]. As
established by the NRC, in the aforementioned GL, such a change will
not alter existing TS requirements or those components to which they
apply. Required information contained in the two tables being
removed is duplicated in the FSAR [final safety analysis report] and
other appropriate plant procedures. Any subsequent changes regarding
the individual components (i.e., the containment isolation valves)
or their operation (e.g., valve positioning under administrative
controls) would be addressed in accordance with the requirements
specified in the Administrative Controls section of the TS regarding
changes to plant procedures and/or changes to the FSAR (i.e., 10 CFR
50.59). These changes will not alter any structure, system, or
component and, therefore, will not result in the possibility of an
increase in [the] probability or consequence of an accident
previously evaluated.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: No. The proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The deletion of two component listings (i.e., Tables 3.6-
1 and 4.4-1) and the TS references to them from the Technical
Specifications and the removal of all references made in the TS
regarding these two listings will not alter how the individual
components (i.e.--the containment isolation valves) identified in
the tables are designed, operated, tested, or maintained. Testing of
CIVs will be performed as required by 10 CFR part 50, Appendix J and
IP3 TS 6.14.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No. The proposed license amendment does not involve a
significant reduction in a margin of safety. The proposed changes
are in accordance with recommendations provided by NRC in Generic
Letter 91-08 and the Standard Technical Specifications, NUREG 1431.
These changes will maintain current safety margins while reducing
the regulatory/administrative burdens to both the NRC and to the
Power Authority. As stated, the changes will not result in changes
to the design, operation, or maintenance of the ClVs, and the
testing of the CIVs will be in accordance with 10 CFR 50 Appendix J
and IP3 TS 6.14.
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
[[Page 27324]]
involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: April 12, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to
remove the footnote restriction found on page 3.1-36 which states that
the departure from nucleate boiling (DNB) analysis contains adequate
margin for Cycle 10, but needs to be reviewed/approved prior to Cycle
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. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response: The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously analyzed. The removal of the footnote on TS page 3.1-36
is an administrative change in that it does not affect the DNB
limits of the current TS. The footnote was added to the TS as part
of Amendment 175, which permitted the use of V+ fuel at IP3. The
footnote required the Authority to demonstrate that sufficient DNB
margin existed for Cycle 11, prior to achieving criticality for that
cycle. The NRC requested this DNB limitation because the
applicability of the WRB-1 correlation to predict DNB performance
for the V+ fuel had not been adequately proven by fuel tests.
Westinghouse has completed fuel tests which verify that the use of
the WRB-1 correlation with the 15 x 15 V+ fuel is conservative.
Therefore, this DNB limitation is no longer applicable and the
footnote can be removed.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Response: The proposed change does not create the possibility of
a new or different kind of accident, as the removal of the footnote
on TS page 3.1-36 does not affect the current TS DNB limits, plant
equipment, or the way the plant is operated. This footnote was
inserted into the TS as part of Amendment 175, which permitted the
use of 15 x 15 V+ fuel at IP3. Westinghouse had used scaling
techniques to demonstrate that the WRB-1 correlation correctly
predicted the critical heat flux performance of the 15 x 15 V+
fuel. Since no fuel tests had been performed on this fuel design,
the NRC was concerned that the use of this correlation may be
unconservative. Therefore, approval to use the V+ fuel at IP3 was
granted based upon the DNB margin available during Cycle 10. This
limitation was contained in the footnote on TS page 3.1-36.
Westinghouse has recently completed fuel tests on 15 x 15 V+ fuel
which verify that the use of the WRB-1 correlation is conservative.
Therefore, the use of V+ fuel at IP3 is no longer dependent on the
amount of DNB margin available and the footnote can be removed.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: The proposed deletion of the footnote on TS page 3.1-
36 does not involve a significant reduction in a margin of safety.
The footnote was introduced as part of Amendment 175, which
permitted the use of V+ fuel at IP3. The footnote required the
Authority to demonstrate that sufficient DNB margin existed for
Cycle 11, prior to achieving criticality for that cycle. The NRC
requested this DNB limitation because the applicability of the WRB-1
correlation to predict DNB performance for the V+ fuel had not been
adequately proven by fuel tests. Westinghouse has completed fuel
tests which verify that the use of the WRB-1 correlation with the 15
x 15 V+ fuel is conservative. Therefore, this DNB limitation is no
longer applicable and the footnote can be removed. The removal of
the footnote is an administrative change as deleting it does not
alter the current DNB margin or future DNB margins.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: March 29, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) by relocating the procedural
details of the Radiological Effluent Technical Specifications (RETS) to
the Offsite Dose Calculation Manual (ODCM). The TSs would also be
revised to relocate procedural details associated with solid
radioactive wastes to the Process Control Program (PCP). In addition,
the Administrative Controls section of the TSs would be revised to
incorporate programmatic controls for radioactive effluents and
environmental monitoring. The proposed changes are consistent with the
guidance provided in Generic Letter 89-01, ``Implementation of
Programmatic Controls for Radiological Effluent Technical
Specifications in the Administrative Controls Section of the Technical
Specifications and the Relocation of Procedural Details of RETS to the
Offsite Dose Calculation Manual or to the Process Control Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes do not affect accident initiators or
precursors and do not alter the design assumptions, conditions,
configuration of the facility or the manner in which the plant is
operated. The proposed changes do not alter or prevent the ability
of structures, systems, or components to perform their intended
function to mitigate the consequences of an initiating event within
the acceptance limits assumed in the Updated Final Safety Analysis
Report (UFSAR). The proposed changes are administrative in nature
and do not change the level of programmatic controls and procedural
details relative to radiological effluents.
Implementation of programmatic controls for RETS in TS will
assure that the applicable regulatory requirements pertaining to the
control of radioactive effluents will continue to be maintained.
Since there are no changes to previous accident analysis, the
radiological consequences associated with these analyses remain
unchanged, therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter the design assumptions,
conditions, configuration of the facility or the manner in which the
plant is operated. The proposed changes have no impact on component
or system interactions. The proposed changes are administrative in
nature and do not change the level of programmatic controls and
procedural details relative to radiological
[[Page 27325]]
effluents. Therefore, these changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) The proposed change does not involve a significant reduction
in a margin of safety.
There is no impact on equipment design or operation and there
are no changes being made to the TS required safety limits or safety
system settings that would adversely affect plant safety as a result
of the proposed changes. The proposed changes are administrative in
nature and do not change the level of programmatic controls and
procedural details relative to radiological effluents. A comparable
level of administrative control will continue to be applied to those
design conditions and associated surveillances being relocated to
the ODCM or PCP. 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: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: April 29, 1999 (TS 99-04).
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) for Sequoyah (SQN) Units 1 and
2 by deleting the Auxiliary Feedwater (AFW) suction pressure low
channel functional surveillance test. The licensee's analysis of the
performance history revealed that the monthly functional test of this
instrument channel does not provide an increased assurance of
operability that justifies the monthly 7 hours per unit system
unavailability that it creates.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The probability of occurrence or the consequences for an
accident is not increased by this request. The proposal to delete
the monthly channel functional test for the auxiliary feedwater
(AFW) suction pressure low functions does not alter the way any
structure, system or component functions, does not modify the manner
in which the plant is operated, and reduces equipment out-of-service
time. This request does not degrade the ability of AFW to perform
its intended function. Therefore, the pressure switches will be
available to perform their intended function.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
A possibility for an accident or malfunction of a different type
than any evaluated previously in SQN's FSAR [Final Safety Analysis
Report] is not created. The proposal does not alter the way any
structure, system or component functions and does not modify the
manner in which the plant is operated. Therefore, the possibility of
a new or different kind of accident previously evaluated is not
created by the proposed change to delete the monthly functional test
of the AFW pressure switches.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The margin of safety has not been reduced since the test
methodologies are not being changed. Increasing the surveillance
interval does not change the results of accident analysis by this
request. The proposed change to delete the AFW system pressure low
channel functional test does not involve a significant reduction in
the margin of safety. The new frequency will not reduce the
reliability of the system and increases overall system availability.
Therefore, changing the frequency of the surveillance does not
reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 3740.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H Knoxville, Tennessee 37902.
NRC Section Chief: Sheri Peterson.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: April 29, 1999 (TS 99-03).
Description of amendment request: The proposed amendment would add
new actions to Technical Specification (TS) Limiting Condition for
Operations (LCOs) 3.3.3.1 and 3.7.7 to address the situation when one
channel of radiation monitoring control room emergency ventilation
system actuation equipment is inoperable and would expand the mode of
applicability for LCOs 3.3.3.1 and 3.7.7 to include periods when
movement of irradiated fuel assemblies are involved and defines actions
to take in these instances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed revision does not change any plant functions or
equipment operating practices for the radiation monitoring system
and control room emergency ventilation system (CREVS). The radiation
monitoring instruments and the CREVS are not considered to be the
source of any accident evaluated in the Final Safety Analysis
Report. These features provide accident mitigation functions that
will be utilized in response to postulated accident conditions. The
activities and failures that could contribute to the initiation of
an accident are not affected by the implementation of this revision.
This revision provides for more stringent requirements for operation
of the facility (additional limiting condition for operation [LCO]
actions and applicability requirements). Therefore the proposed
activity will not increase the probability of an accident.
The proposed activity does not affect accident mitigation
capabilities or the radiation release amounts for postulated
accidents. This TS change will not affect requirements that the
radiation monitoring system and CREVS be maintained to support
accident mitigation. The functions and testing will remain the same
while operability requirements will become more stringent. This TS
change enhances the requirements associated with CREVS and the
initiation of this system such that inoperabilities are
appropriately handled to reduce the safety impact of component
inoperabilities. Therefore, the proposed change will not increase
the consequences of an accident and could reduce the consequences by
limiting operation with inoperable components and requiring the
application of appropriate actions for all conditions that could
result in a postulated accident that CREVS was designed to mitigate.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change provides more stringent operating
requirements for operation of the facility. The proposed
[[Page 27326]]
activity will not change any plant function or operating practice
that could impact accident initiators. Therefore, these more
stringent requirements do not result in operation that will increase
the probability of any postulated accidents. In addition, CREVS and
the associated actuation features are not considered to be the
source of an accident. Therefore, the proposed activity will not
create the possibility of an accident of a different kind.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed activity does not impact plant setpoints designed
to maintain the assumptions in the safety analysis or limits for the
actuation of systems to mitigate accidents. Plant functions and
operating practices will not be altered by the implementation of
more stringent requirements for operation of the facility. These
requirements, by definition, provide additional restrictions to
enhance plant safety. Therefore, the proposed activity will not
reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H Knoxville, Tennessee 37902.
NRC Section Chief: Sheri Peterson.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: February 1, 1999, as supplemented on
April 19 and April 23, 1999.
Description of amendment request: The amendment request proposes a
total replacement of current Technical Specifications Section 6,
``Administrative Controls.'' Administrative changes to certain other
sections of Technical Specifications are also being made to conform to
the changes resulting from the re-write of Section 6.
The proposed changes represent a comprehensive upgrade of Section 6
of the Vermont Yankee Technical Specifications, incorporating
improvements in content and format based on industry standards. In
accordance with industry practice some Technical Specifications
requirements are being relocated to the recently implemented Vermont
Yankee Technical Requirements Manual (TRM), Offsite Dose Calculation
Manual (ODCM), or Vermont Yankee Operational Quality Assurance Manual
(VOQAM) and will be eliminated from the Technical Specification upon
NRC approval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:
The proposed changes have no effect on plant hardware, plant
design, safety limit setting, or plant system operation and
therefore do not modify or add any initiating parameters that would
significantly increase the probability or consequences of an
accident previously evaluated.
No new modes of operation are introduced by the proposed changes
such that additional adverse consequences would result. Accordingly,
the consequences of previously analyzed accidents are not
deleteriously affected by this proposed license amendment.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, because:
The proposed changes do not involve any physical alteration of
the plant (no new or different type of equipment will be installed)
or any change in the methods governing normal plant operation. These
changes do not affect the operation of any systems or components,
nor do they involve any potential initiating events that would
create any new or different kind of accident. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated for VYNPS.
3. Involve a significant reduction in a margin of safety,
because:
The proposed changes have no impact on any safety analysis
assumptions. Consequently, no margin of safety as described in the
Final Safety Analysis Report and defined in the basis of any
Technical Specification is reduced as a result of these changes.
These proposed changes do not detrimentally affect the ability
of structures, systems and components important to safety to fulfill
their intended safety functions. Therefore, it is concluded that the
proposed changes do no[t] involve a significant reduction in a
margin of safety.
Additional Safety Considerations for Specific Changes Deemed to be
``Less Restrictive''
In accordance with the criteria set forth in 10 CFR 50.92, Vermont
Yankee has evaluated the proposed changes to the [Vermont Yankee
Nuclear Power Station] VYNPS Technical Specifications and determined
that they do not involve a significant hazards consideration. Those
changes which are deemed to be ``less restrictive'' have been subject
to the following additional consideration:
(a) Changes which are deemed to be ``less restrictive'' based
solely upon removal from the Technical Specifications and relocated
in VYNPC-controlled documents:
NRC's Technical Specifications Branch has conducted reviews of
the Administrative Controls section of Standard Technical
Specifications and concluded that certain provisions historically
contained in Technical Specifications can be relocated to other
licensee documents for which changes to those provisions are
adequately controlled by other regulatory requirements. In general,
Administrative Controls are those requirements not covered by other
Technical Specifications, but are considered necessary to assure
operation of the facility in a safe manner. Application of this
criterion can be based on two categories or requirements: (a)
requirements not covered by other regulatory requirements, but are
considered necessary to assure the safe operation of the facility or
(b) specific requirements that are broadly covered by regulations or
other regulatory controls, for which details need to be specified in
the Technical Specifications to ensure safe plant operation. In
general, however, Technical Specifications need not duplicate other
regulatory requirements.
As identified in Attachment A hereto, certain portions of the
current Technical Specifications are to be relocated to the
Technical Requirements Manual (TRM), Offsite Dose Calculation Manual
(ODCM), or the Vermont Yankee Operational Quality Assurance Manual
(VOQAM) and removed from the Technical Specifications. As an initial
step in this process, the subject requirements are being duplicated
in the TRM, ODCM, or VOQAM. Removal from the Technical
Specifications will occur upon NRC approval. The ability to relocate
these requirements is based on regulations and standards that
contain these provisions such that duplication in the Technical
Specifications is not necessary.
[1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:]
The TRM is a[n] FSAR level document and is incorporated by
reference into the FSAR. Changes to the TRM will be strictly
controlled by the 10 CFR 50.59 process to ensure that proper reviews
are conducted. The relocation of requirements to the VYNPC-
controlled TRM will not diminish the effectiveness of compliance
withthe relocated provisions. Since any changes to the TRM will be
evaluated per the requirements of 10 CFR 50.59, no increase
(significant or insignificant) in the probability or consequences of
an accident previously analyzed will be allowed. Therefore, these
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Changes to the ODCM are controlled by current Technical
Specifications and require the reporting to the NRC of changes to
the
[[Page 27327]]
ODCM with sufficient information to support the changes together
with appropriate analyses or evaluations justifying the changes. The
relocation of these details to the ODCM is thus acceptable
considering the controls provided by existing regulations and the
controls remaining in Technical Specifications for ODCM changes.
Therefore, these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Relocation of the Technical Specification Administrative
Controls related to quality assurance from the Technical
Specifications to the VOQAM is consistent with the guidance provided
by the NRC in Administrative Letter 95-06, ``Relocation of Technical
Specification Administrative Controls Related to Quality
Assurance.'' Changes to the VOQAM are subject to the change control
process in 10CFR50.54(a). These provisions are adequate to ensure
that quality assurance program commitments are not reduced without
prior NRC approval. Therefore, these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
[2. Create the possibility of a new or different kind of
accident from any accident previously evaluated, because:]
The proposed changes do not involve any physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change will not impose or eliminate any requirements, and
adequate control of the information will be maintained. Thus, this
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,
because:]
The proposed changes will not reduce a margin of safety because
they have no impact on any safety analysis assumption. In addition,
the details to be transposed from the Technical Specifications to
the TRM, ODCM, and VOQAM are the same as the existing Technical
Specifications. Since any future changes to these provisions in the
TRM will be evaluated per the requirements of 10CFR50.59 and
Technical Specifications already requires supporting information be
submitted to the NRC for ODCM changes, no reduction (significant or
insignificant) in a margin of safety will be allowed. The provisions
of 10CFR50.54(a) are adequate to control changes to the VOQAM and
maintain current margins of safety.
Based on 10CFR50.92, the existing requirement for NRC review and
approval of revisions (to the Technical Specifications provisions
proposed for relocation) does not have a specific margin of safety
upon which to evaluate. However, since the proposed changes are
consistent with industry standards, approved by the NRC, revising
the Technical Specifications to relocate these provisions will not
diminish administrative controls necessary to assure the safe
operation of the facility.
(b) Change [9] identified in Attachments A and D [of the
February 1, 1999, submittal]:
This change proposes to relax the requirement to have an
individual qualified in radiation protection procedures onsite at
all times. The proposed change will allow the position to be vacant
for up to two hours in order to provide for unexpected absence.
[1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:]
The proposed change does not affect the probability of an
accident. The actions of an individual qualified in radiation
protection procedures are not assumed to be an initiator of an
accident. Also, the consequences of an accident are not affected by
the presence of an individual qualified in radiation protection
procedures. This proposed change does not impact the assumptions of
any design basis accident. This change will not alter assumptions
relative to the mitigation of an accident or transient event. This
change will not have any impact on the safe operation of the plant
because the presence of a person qualified in radiation protection
procedures is not required for the mitigation of any accident.
Therefore, this change 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, because:]
This change will not physically alter the plant (no new or
different type of equipment will be installed). The changes in
methods governing normal plant operation are consistent with the
current safety analysis assumptions. Therefore, this change will not
create the possibility of a new or different type of accident from
any accident previously evaluated.
[3. Involve a significant reduction in a margin of safety,
because:]
The margin of safety is not affected by the presence or absence
onsite of an individual qualified in radiation protection
procedures. This proposed change has no effect on the assumptions of
any design basis accident. This change has no impact on the safe
operation of the plant since the presence onsite of an individual
qualified in radiation protection procedures is not required for the
mitigation of an accident. This change does not affect any plant
equipment or requirements for maintaining plant equipment. The
safety analysis assumptions will still be maintained, thus no
question of safety exists. Therefore, this change does not involve a
significant reduction in a margin of safety.
(c) Change [10] identified in Attachments A and D [of the
February 1, 1999, submittal]:
This change proposes to incorporate the allowances of a
temporary deviation from the shift staffing levels of
10CFR50.54(m)(2)(i) for up to two hours. In addition, this change
proposes to apply these same allowances to the positions of Shift
Engineer and non-licensed operators.
[1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:]
The proposed change does not affect the probability of an
accident. The shift staffing level requirements are not assumed to
be an initiator or any analyzed event. Also, the consequences of an
accident are not affected by these temporary deviations to the shift
staffing levels. This proposed change does not impact the
assumptions of any design basis accident. This change will not alter
assumptions relative to the mitigation of an accident or transient
event, since 10CFR50.54(m) (ii) and (iii) still maintain the
requirements for the presence of licensed operators and senior
operators. This change has no impact on the safe operation of the
plant. The level of shift staffing will still be maintained as
required by 10CFR50.54(m) (ii) and (iii) and does not affect any
plant equipment or requirements for maintaining plant equipment. The
temporary deviations from the shift staffing level for up to two
hours to provide for unexpected absence, provided immediate action
is taken to fill the required position is acceptable in terms of
staffing requirements for the mitigation of an accident due to the
low probability of an accident occurring during these short-term,
infrequent deviations and the remaining licensed operators and
senior operators. Therefore, this change 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, because:]
This change will not physically alter the plant (no new or
different type of equipment will be installed). The temporary
deviations from shift staffing levels are consistent with the
current safety analysis assumptions. Therefore, this change will not
create the possibility of a new or different type of accident from
any accident previously evaluated.
[3. Involve a significant reduction in a margin of safety,
because:]
The margin of safety in not reduced by allowing these temporary
deviations from shift staffing levels due to unforeseen events. This
proposed change has no effect on the assumptions of any design basis
accident. This change has no impact on the safe operation of the
plant since 10CFR50.54(m) (ii) and (iii) still maintain the
requirements for the minimum number of licensed operators and senior
operators necessary to safely operate the plant. This change does
not affect any plant equipment or requirements for maintaining plant
equipment. The safety analysis assumptions will still be maintained,
thus no question of safety exists. Therefore, this change does not
involve a significant reduction in a margin of safety.
(d) Changes [38] and [39] identified in Attachments A and D [of
the February 1, 1999, submittal]:
In accordance with 10CFR20.1601 (c), these changes propose
alternative methods for controlling access to high radiation areas
consistent with the intent of 10CFR20.1601 (a) and (b).
[1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:]
The proposed changes do not affect the probability of an
accident. The controls used for access to high radiation areas are
not assumed in the initiation of any analyzed event. Also, the
consequences of an accident are not affected by these changes. These
changes are both consistent with good
[[Page 27328]]
radiological practices and will provide an adequate level of
radiation protection. These proposed changes do not impact the
assumptions of any design basis accident. These changes will not
alter assumptions relative to the mitigation of an accident or
transient event. These changes have no impact on safe operation of
the plant. Therefore, these 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, because:]
The proposed changes will not create the possibility of an
accident. These changes will not physically alter the plant (no new
or different type of equipment or system will be installed). The
changes in methods governing normal plant operations are consistent
with the current safety analysis assumptions and deal only with
personnel exposure to radiation, not reactor safety. Therefore,
these changes will 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,
because:]
The margin of safety is not reduced due to these proposed
changes. These changes are both consistent with good radiological
safety practice and have been found to provide adequate levels of
radiation protection. In addition, these changes provide the benefit
of ensuring radiation dose to workers can be minimized by providing
the flexibility to select the best means of providing access control
to a high radiation area, given the plant area and radiological
conditions. These proposed changes have no impact on the safe
operation of the plant. No change in analytic limits or setpoints is
introduced by these changes. The safety analysis assumptions will
still be maintained, thus no question of nuclear safety exits.
Therefore, these changes do not involve a significant reduction in a
margin of safety.
(e) Change [49] identified in Attachments A and D [of the
February 1, 1999, submittal]:
This change proposes to relax the requirement for submitting the
(now-named) Occupational Radiation Exposure Report from the
currently required date of March 1 to April 30 of each year. April
30 is now the industry standard date for submittal of such reports.
[1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:]
The proposed change does not affect the probability of an
accident. The submittal date of the Occupational Radiation Exposure
Report is not assumed to be an initiator of any analyzed event.
Also, the consequences of an accident are not affected by the
submittal date of this report. This proposed change does not impact
the assumptions of any design basis accident. This change will not
alter assumptions relative to the mitigation of an accident or
transient event. This change has no impact on the safe operation of
the plant. The report will still be required to be submitted each
year and does not affect any plant equipment or requirements for
maintaining plant equipment. The submittal date of this report is
not required for the mitigation of any accident. Therefore, this
change 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, because:]
The proposed change will not create the possibility of an
accident. This change will not physically alter the plant (no new or
different type of equipment will be installed). The change in method
governing submittal of this report does not affect current safety
analysis assumptions. Therefore, this change will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
[3. Involve a significant reduction in a margin of safety,
because:]
The margin of safety i[s] not reduced by allowing the report to
be submitted 60 days later. This proposed change has no effect on
the assumptions of the design basis accident. This change has no
impact on the safe operation of the plant. The report will still be
required to be submitted each year and does not affect any plant
equipment or requirements for maintaining plant equipment. The
safety analysis assumptions will still be maintained, thus no
question of safety exists. Therefore, this change does not involve a
significant reduction in a margin of safety.
(f) [Change [64] identified in Attachments A and D [of the
February 1, 1999, submittal]:
This change proposes to relax the requirement for submitting the
(now-named) Annual Radiological Environmental Operating Report from
the currently required date of May 1 to May 15 of each year. May 15
is now the industry standard date for submittal of such reports.
[1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:]
The proposed change does not affect the probability of an
accident. The submittal date of this report is not assumed to be an
initiator of any analyzed event. Also, the consequences of an
accident are not affected by the submittal date of this report. This
proposed change does not impact the assumptions of any design basis
accident. This change will not alter assumptions relative to the
mitigation of an accident or transient event. This change has no
impact on the safe operation of the plant. The report will still be
required to be submitted each year and does not affect any plant
equipment or requirements for maintaining plant equipment. The
submittal date of this report is not required for the mitigation of
any accident. Therefore, this change 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, because:]
The proposed change will not create the possibility of an
accident. This change will not physically alter the plant (no new or
different type of equipment will be installed). The change in method
governing submittal of this report does not affect current safety
analysis assumptions. Therefore, this change will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
[3. Involve a significant reduction in a margin of safety,
because:]
The margin of safety i[s] not reduced by allowing the report to
be submitted 14 days later. This proposed change has no effect on
the assumptions of the design basis accident. This change has no
impact on the safe operation of the plant. The report will still be
required to be submitted each year and does not affect any plant
equipment or requirements for maintaining plant equipment. The
safety analysis assumptions will still be maintained, thus no
question of safety exists. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: April 20, 1999.
Description of amendment request: The amendment request proposes
changes to the existing requirements associated with the unloading and
loading of fuel in the reactor vessel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
VY has determined that the proposed change to reload the reactor
core in a spiral pattern beginning around a Source Range Monitor
(SRM) does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The design basis
accident associated with refueling is the Refueling Accident; i.e.,
the accidental dropping of a fuel bundle onto the top of the core.
There is no assumption as to
[[Page 27329]]
the core loading pattern in the analysis of this accident. The
analyzed abnormal operational transients associated with refueling
are: (1) the Control Rod Removal Error During Refueling, and (2) the
Fuel Assembly Insertion Error During Refueling. There is no
assumption as to the core loading pattern in the analyses of these
transients. The Fuel Assembly Insertion Error During Refueling
transient involves mislocated and rotated fuel assembly loading
errors. However, a change in the approved core loading pattern has
no impact on the probability of mislocating or rotating a bundle
while following that pattern. Furthermore, the proposed change
implements a core loading pattern that provides improved flux
monitoring as compared to the pattern prescribed by the current
Technical Specifications. When loading the core in accordance with
the proposed change, the SRM indication will be indicative of the
true flux of the loaded fuel, as the creation of flux traps
(moderator filled cavities surrounded on all sides by fuel) is
precluded.
The Technical Specification Bases are under the purview of
10CFR50.59. As such, subsequent changes made via 10CFR50.59 to the
information relocated to the Bases are not allowed to increase the
probability or consequences of an accident previously evaluated.
Therefore, relocating the details of the core loading pattern to the
Bases does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The SRMs and the core loading pattern are not initiators of any
accident previously evaluated. As such, the subject changes cannot
affect the probability of an accident previously evaluated. The core
loading pattern is not assumed in the mitigation of any accident.
Since the proposed change provides improved flux monitoring by the
SRMs, operators will have more accurate indication and SRM automatic
trip functions will actuate more accurately. As such, any event
mitigation function provided by the SRMs is enhanced by this change.
Therefore, the associated changes do not involve a significant
increase in the consequences of an accident previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
VY has determined that the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated. VY proposes to change the core reloading and
offloading patterns to start and stop, respectively, at an SRM
versus the geometric center of the core as prescribed by current
Technical Specifications. This ensures that flux monitoring
instrumentation is always OPERABLE in the fueled region of the
vessel. There is no separation of the monitoring device from the
fuel by cavities of water as is the case with the pattern prescribed
by the current Technical Specifications. As such, flux monitoring is
enhanced during core reloading and offloading. This change is
conservative relative to the current requirements. Therefore, no new
categories or types of accidents are created.
Additionally, the Technical Specification Bases are under the
purview of 10CFR50.59. As such, subsequent changes made via
10CFR50.59 to the information relocated to the Bases are not allowed
to create the possibility for an accident or malfunction of a
different type than any evaluated previously in the safety analysis
report. Therefore, relocating the details of the core loading
pattern to the Bases does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
VY has determined that the proposed change does not involve a
significant reduction in a margin of safety. Loading around the
geometric center of the core as prescribed by the current Technical
Specifications results in cells of moderator separating the fuel
from the instrumentation monitoring its flux. This change requires
the flux monitoring instrumentation to be in the fueled region, and,
in so doing, provides for more accurate monitoring of core flux
during core reloading and offloading. As such, the operators will
have more accurate indication and SRM automatic trip functions will
actuate when the actual flux reaches the trip setpoints. This
corrects non-conservatisms that result from cells of moderator
separating the fuel from the instrumentation. Therefore, this change
will not result in a significant reduction in a margin of safety.
Additionally, the details of the loading pattern are relocated
from the Technical Specifications to the Bases. Since any future
changes to the Bases will be evaluated per the requirements of 10
CFR 50.59, no reduction in a margin of safety will be allowed.
Therefore, relocating the core loading pattern details to the Bases
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: April 7, 1999.
Description of amendment request: The proposed amendment would
revise the minimum critical power ratio (MCPR) limit in Technical
Specification (TS) 2.1.1.2, for the ATRIUM-9X and the SVEA-96 fuel for
one and two recirculation loop operation. The proposed amendment would
add a new reference in TS 5.6.5, ``Core Operating Limits Report.'' The
reference cites ANFB Critical Power Correlation Uncertainty for Limited
Data Sets, ANF1125(P)(A), Supplement 1, Appendix D, Siemens Power
Corporation-Nuclear Division, July 1998.
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 probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
Technical Specifications amendment uses conservatively established
SLMCPR [safety limit minimum critical power ratio] values for WNP-2
such that the fuel is protected during normal operation as well as
during plant transients or anticipated operational occurrences.
The probability of an evaluated accident is not increased by the
use of the ATRIUM-9X MCPR safety limit of 1.10 (two loop operation)
or 1.11 (single loop operation). The ATRIUM-9X fuel was evaluated by
SPC (Reference 5) [Letter KVW:98:148 dated July 8, 1998, KV Walters,
(Siemens Power Corporation), to RA Vopalensky (Supply System),
``MCPR Safety Limit Reanalysis for WNP-2 Cycle 11''] using the
additive constant uncertainty for ATRIUM-9X fuel of 0.0201 which is
contained in the NRC safety evaluation approval of Reference 4 [ANFB
Critical Power Correlation Uncertainty for Limited Data Sets, ANF-
1125(P)(A), Supplement 1, Appendix D, Siemens Power Corporation--
Nuclear Division, July 1998]. Based upon the NRC approved additive
constant of uncertainty of 0.0201, as documented in Reference 5, at
least 99.9% of the SPC ATRIUM-9X fuel rods would be expected to
avoid boiling transition with a SLMCPR of 1.10 during two loop
operation and 1.11 during single loop operation.
The probability of an evaluated accident is not increased by the
use of the ABB SVEA-96 SLMCPRs of 1.10 (two loop operation) or 1.12
(single loop operation). NRC approved
[[Page 27330]]
methodology documented in CENPD-300-P-A, ``Reference Safety Report
for Boiling Water Reactor Reload Fuel'', July 1996 (Reference 3) was
used in deriving these ABB SVEA-96 SLMCPR values. The ABB evaluation
as a function of cycle exposure established that late in Cycle 15
conservative two loop and single loop SLMCPRs of 1.10 and 1.12,
respectively, can be used to represent the entire cycle.
The SLMCPR changes do not require any physical plant
modifications, physically affect any plant component, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
Since the operability of plant systems designed to mitigate any
consequences of accidents have not changed, the consequences of an
accident previously evaluated are not expected to increase.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. This Technical Specification submittal
does not involve any modifications of the plant configuration or
allowable modes of operation. This Technical Specification change
establishes SLMCPRs for SPC fuel based upon the NRC approved
additive constant of uncertainty of 0.0201, as documented in
Reference 5. At least 99.9% of the SPC ATRIUM-9X fuel rods would be
expected to avoid boiling transition with an SLMCPR of 1.10 during
two loop operation or 1.11 during single loop operation.
Additionally, the ABB SVEA-96 SLMCPRs of 1.10 (two loop operation)
or 1.12 (single loop operation) were derived using the NRC approved
methodology documented in CENPD-300-P-P, ``Reference Safety Report
for Boiling Water Reactor Reload Fuel'', July 1996 (Reference 3).
Therefore, no new precursors of an accident are created and no new
or different kinds of accidents are created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Implementation of SLMCPRs derived by proven analytical methods
provides a margin of safety by ensuring that less than 0.1% of the
rods are expected to be in boiling transition if the MCPR limit is
not violated. Because the fuel design safety criteria of more than
99.9% of the fuel rods avoiding transition boiling during normal
operation as well as anticipated operational occurrences is met,
there is not 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, NW, Washington, DC 20005-3502.
NRC Project Director: Stuart Richards.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: April 20, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.4.11, ``RCS Pressure and
Temperature Limits,'' to update the curves that set forth the pressure
temperature limit lines. The curves provide the pressure temperature
limits for the operation of the reactor coolant system for heatup and
cooldown during inservice leak and hydrostatic testing, non-nuclear
heating and cooldown, and nuclear heating and cooldown.
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 pressure temperature shift is well within the operating
margins of plant equipment. Using the new non-nuclear and nuclear
heating and cooldown curves, higher temperature values for
corresponding pressures at temperatures which are closest to RT
NDT, further reduce the potential for brittle fracture.
The proposed 32 EFPY [effective full power years] curves were
developed using methodology that is consistent with the guidance in
Regulatory Guide 1.99, Revision 2, Appendix G of the ASME Code and
Appendix G of 10 CFR part 50. This methodology is recognized by the
NRC and the industry as providing acceptable margin.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will 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 has no impact on the previously analyzed
accidents or transients. The proposed change does not introduce any
credible mechanisms for unacceptable radiation release nor does it
require physical modification to the plant. The 32 EFPY curves are
calculated using a published methodology that was discussed with the
NRC.
The proposed change is also within any upper bound limit. The
only impact on plant operation is that the plant will be operated
with new pressure temperature limits derived from the proposed
alternative calculational methodology in place of the previously
approved model based on actual plant data.
Therefore, the operation of WNP-2 in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The results of testing reflected 30 ft-lb shifts and changes in
uppershelf energy of the base plate and the weld material. However,
the results are well within the values predicted by Regulatory Guide
1.99, Revision 2. Furthermore, the adjusted reference temperature
values and the upper shelf energy of the reactor beltline materials
are expected to remain within the limits of 10 CFR part 50, Appendix
G, for at least 32 effective full power years of reactor operation.
For the non-nuclear and nuclear heating and cooldown curves
(with a calculated through wall T), lower temperatures
which are closest to RTNDT, have an increased margin of
safety due to the higher required temperature values for a given
pressure than is required by current curve calculation methodology.
Thus additional margin to brittle fracture is achieved for non-
nuclear and nuclear heating.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Stuart Richards.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait
[[Page 27331]]
for this biweekly notice or because the action involved exigent
circumstances. They are repeated here because the biweekly notice lists
all amendments issued or proposed to be issued involving no significant
hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: December 31, 1997, as supplemented May
15, September 15, November 25, 1998 and January 28, 1999.
Description of amendment request: Revise the St. Lucie, Unit 2,
Technical Specifications to increase the capacity of the spent fuel
storage pool, in part, by allowing a credit for a certain soluble boron
concentration in the spent fuel pool.
Date of publication of individual notice in the Federal Register:
April 5, 1999 (64 FR 16502).
Expiration date of individual notice: May 5, 1999.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: April 19, 1999.
Brief description of amendments: The amendments would revise
Technical Specification Section 3/4.8.1.2, ``Electrical Power Systems,
Shutdown,'' and its associated bases to provide a one-time extension of
the 18-month surveillance interval for specific surveillance
requirements for Units 1 and 2. This surveillance will be performed
prior to the first entry into Mode 4 subsequent to receipt of the
requested T/S amendment. In addition, for Unit 2 only, a minor
administrative change is included to delete a reference to T/S 4.0.8,
which is no longer applicable. For Unit 1 only, an editorial change is
made to add the word ``or'' to action statement 3.8.1.2.
Date of publication of individual notice in Federal Register: April
29, 1999 (64 FR 23129).
Expiration date of individual notice: June 1, 1999.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: April 6, 1999.
Description of amendment request: The proposed amendments would
allow an increase of 168 fuel assemblies in the storage capacity of
Unit 1's Spent Fuel Pool and an increase of 88 fuel assemblies in the
storage capacity of Unit 2's Spent Fuel Pool.
Date of publication of individual notice in Federal Register: May
4, 1999 (64 FR 23877).
Expiration date of individual notice: June 3, 1999.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
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.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: December 29, 1998.
Brief description of amendments: The amendments change Technical
Specification Tables 3.3.1-1 and 3.3.2-1 to revise the Allowable Values
for 12 functions of the Reactor Trip System and Engineered Safety
Features Actuation System.
Date of issuance: April 23, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 107, 107, 100 and 100.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9186). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: October 30, 1998.
Brief description of amendments: The amendments revised the
Technical Specification (TS) requirements for
[[Page 27332]]
spent fuel pool inadvertent draindown elevation.
Date of issuance: May 3, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 108, 108 101, and 101.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69335). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 0481.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of application of amendment: June 2, 1998, and as supplemented
by letters dated January 18 and March 9, 1999.
Brief description of amendment: The amendment relocates
requirements related to seismic monitoring instrumentation from the
Technical Specifications to the Technical Requirements Manual.
Date of issuance: April 28, 1999.
Effective date: Immediately; and shall be implemented within 60
days of issuance.
Amendment No.: 194.
Facility Operating License No. DPR-61: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50936). The January 18 and March 9, 1999, supplements contained
revised TS pages to account for TS changes issued by the NRC since the
original June 2, 1998, submittal, pages from the Updated Final Safety
Analysis Report and TRM, which were revised to support the June 2,
1998, request, and additional clarifications. The supplemental
information did not change the staff's initial proposed no significant
hazards consideration determination or expand the scope of the original
notice. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 28, 1999.
No significant hazards consideration received: No
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: August 21, 1996, as supplemented
May 2, 1997.
Brief description of amendment: The amendment revised Section 3.3.G
(Hydrogen Recombiner System and Post-Accident Containment Venting
System), the basis for Section 3.3.G, and Section 4.4, Table 4.4-1
(Containment Isolation Valves). This change permits removal of the
existing flame-type hydrogen recombiners, its supporting equipment, and
replacement with passive autocatalytic recombiners.
Date of issuance: April 27, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 200.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: Janaury 29, 1997 (62 FR
4345). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 27, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of application for amendment: September 3, 1997.
Brief description of amendment: The amendment revises TS 3.14,
Control Room Ventilation, to be consistent with NUREG-1432, Standard
Technical Specifications, Combustion Engineering Plants.
Date of issuance: May 6, 1999.
Effective date: May 6, 1999.
Amendment No.: 186.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14281). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 6, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423-3698.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application of amendments: March 1, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications by adding a Note to Improved Technical
Specification (ITS) 3.9, ``Refueling Operations,'' Subsection 3.9.3,
``Containment Penetrations,'' Limiting Condition for Operation 3.9.3.b,
to state that the emergency air lock door is not required to be closed
when it is sealed with the temporary cover plate.
Date of Issuance: April 28, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: Unit 1-303; Unit 2-303; Unit 3-303.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: 64 FR 14282 (March 24,
1999). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 28, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: March 1, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications by changing the number of required channels
shown in TS Table 3.3.8-1, ``Post Accident Monitoring Instrumentation''
for the Reactor Coolant System Hot Leg Temperature function from ``2
per loop'' to ``2.''
Date of Issuance: April 28, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: Unit 1-304; Unit 2-304; Unit 3-304
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14281). The Commission's related evaluation of
[[Page 27333]]
the amendments is contained in a Safety Evaluation dated April 28,
1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: April 30, 1998.
Brief description of amendment: The amendment revises the single
largest post-accident load capable of being supplied by the diesel
generators and relocates this value to the Bases for Technical
Specification (TS) Surveillance 4.8.1.1.2.c.3. TS Surveillance
4.8.1.1.2.c.3 has been revised to refer to ``the single largest post-
accident load'' rather than a specific numerical value for diesel
generator load reject testing. This change is consistent with the
guidance provided in NUREG-1432 , ``Improved Standard Technical
Specifications for Combustion Engineering Plants.''
Date of issuance: April 21, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 204.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56241). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 21, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: July 21, 1995.
Brief description of amendment: The amendment extends the
expiration date of Operating License NPF-29 for Grand Gulf Nuclear
Station, Unit 1, from June 16, 2022, to November 1, 2024. The extended
date is 40 years from the date the full-power license was issued for
the plant on November 1, 1984.
Date of issuance: April 26, 1999.
Effective date: As of the date of issuance to be implemented within
30 days of issuance.
Amendment No: 137.
Facility Operating License No. NPF-29: Amendment revises Operating
License No. NPF-29.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42605). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 16, 1998.
Brief description of amendment: The amendment changes Technical
Specification (TS) Section 2.1.1.2, ``Reactor Core [Safety Limits],''
by revising the two recirculation loop Minimum Critical Power Ratio
(MCPR) limit from 1.13 to 1.12 and the single recirculation loop MCPR
limit from 1.14 to 1.13. The revised limits are required to address the
River Bend Cycle 9 core design and operation. The proposed TS changes
are scheduled to be implemented following refueling outage 8, currently
scheduled to begin in April 1999.
Date of issuance: April 27, 1999.
Effective date: As of the date of issuance to be implemented prior
to the startup following refueling outage 8.
Amendment No.: 105.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9190). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 27, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 8, 1998, as supplemented April
15, 1999.
Brief description of amendment: The amendment implements the
Boiling Water Reactor Owners Group Enhanced Option I-A for the reactor
stability long-term solution to the neutronic and thermal hydraulic
instability that is documented in NEDO-32339, Revision 1, ``Reactor
Stability Long-Term Solution, Enhanced Option I-A.''
Date of issuance: May 5, 1999.
Effective date: As of the date of issuance and shall be implemented
during refueling outage 8.
Amendment No.: 106.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64112). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 5, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: October 30, 1998, as
supplemented March 31, 1999.
Brief description of amendment: The amendment proposed to revise
the Final Safety Analysis Report (FSAR) and associated Improved
Technical Specification (ITS) Bases to reflect changes in the
methodology for the B spent fuel pool criticality analysis. The
proposed change is necessary due to Boraflex degradation in the B spent
fuel pool storage racks.
Date of issuance: April 27, 1999.
Effective date: April 27, 1999.
Amendment No.: 175.
Facility Operating License No. DPR-72: Amendment approves changes
to the FSAR and ITS Bases.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71966). The supplemental letter dated March 31, 1999, did not change
the original no significant hazards consideration determination.
[[Page 27334]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 27, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: January 27, 1999.
Brief description of amendment: The change would allow a one-time
extension of approximately 2 months of the steam generator tube
inspection interval in order for the inspection to coincide with the
next planned refueling outage.
Date of issuance: May 5, 1999.
Effective date: May 5, 1999.
Amendment No.: 176.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 10, 1999 (64 FR
11962). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 5, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: December 31, 1997, as
supplemented May 15, 1998, September 15, 1998, November 25, 1998, and
January 25, 1998.
Brief description of amendment: This change modified the St. Lucie
Unit 2 Technical Specifications to increase the capacity of the spent
fuel storage pool, in part, by allowing a credit for a certain soluble
boron concentration in the spent fuel pool.
Date of Issuance: May 6, 1999.
Effective Date: Upon issuance of license amendment package with
implementation by the end of the next scheduled refueling outage,
currently scheduled for April of 2000.
Amendment No.: 101.
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 11, 1998 (63
FR 6985) and December 16, 1998 (63 FR 69340). Following the receipt of
the supplement dated November 25, 1998, and the staff's subsequent no
significant hazards consideration determination (63 FR 69340), the
supplement dated January 28, 1999, contained clarifying information
that did not change the no significant hazards consideration
determination. An additional notice was required, in accordance with 10
CFR 2.1107, due to an oversight (64 FR 16502, April 5, 1999). An
environmental assessment has been published in the Federal Register (64
FR 23133, April 29, 1999). In that assessment, the Commission
determined that the issuance of this amendment will not result in any
environmental impacts other than those evaluated in the Final
Environmental Statement.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 6, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: February 24, 1999.
Brief description of amendments: The amendments changed Technical
Specification (TS) 3/4.7.4 to permit the option of monitoring the
ultimate heat sink temperature afer the intake cooling water (ICW)
pumps but before the component cooling water heat exchangers which is
considered to be equivalent to temperature monitoring before the ICW
pumps.
Date of issuance: May 5, 1999.
Effective date: May 5, 1999.
Amendment Nos.: 200 and 194.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the TS.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14282). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 5, 1999.
No significant hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: July 14, 1998
Brief description of amendment: The proposed amendment changed the
Technical Specifications to revise the liquid and gaseous release rate
limits to reflect revisions to 10 CFR Part 20, ``Standards for
Protection Against Radiation.''
Date of issuance: May 3, 1999.
Effective date: May 3, 1999, to be implemented within 30 days from
the date of issuance.
Amendment No.: 163.
Facility Operating License No. DPR-36: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 13, 1999 (64 FR
2249). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: September 30, 1997.
Brief description of amendment: The proposed amendment revises
portions of Facility Operating License No. DPR-36 to delete License
Conditions 2.B.6.c, 2.B.6.e, 2.B.6.f, 2.b.6.g, 2.b.7(a), and 2.B.7(b)
which are no longer applicable due to the permanently shutdown and
defueled condition of the Maine Yankee Atomic Power Station. Orders
dated May 23, 1980, August 29, 1980, and September 19, 1980, are
rescinded due to their being superseded by the equipment qualification
rule (10 CFR 50.49).
Date of issuance: May 5, 1999.
Effective date: May 5, 1999, and shall be implemented within 30
days from the date of issuance.
Amendment No.: 164.
Facility Operating License No. DPR-36: The amendment revised the
Operating License.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63978). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 5, 1999.
[[Page 27335]]
No significant hazards consideration comments received: No.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: January 14, 1998, as
supplemented by letters dated May 19, 1998, September 28, 1998, and
three letters dated February 5, 1999.
Brief description of amendments: The amendments authorize revisions
to the licensing basis as described in the Final Safety Analysis Report
(FSAR) Update to incorporate the modification to the 230 kV offsite
power system.
Date of issuance: April 29, 1999.
Effective date: April 29, 1999, and shall be implemented in the
next periodic update to the FSAR Update in accordance with 10 CFR
50.71(e).
Amendment Nos.: Unit 1-132; Unit 2-130.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Final Safety Analysis Report Update.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53952). The supplemental letters dated September 28, 1998, and the
three letters dated February 5, 1999, provided additional clarifying
information, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 29, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: September 3, 1998, as
supplemented by letters dated January 22, 1999, February 5, 1999, and
March 17, 1999.
Brief description of amendments: The amendments change the
Technical Specifications to revise TS 3/4.4.9.1 Figures for heatup and
cooldown to extend their applicability to 16 effective full power
years.
Date of issuance: May 3, 1999.
Effective date: May 3, 1999, to be implemented within 30 days from
the date of issuance.
Amendment Nos.: Unit 1-133; Unit 2-131.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR69345). The supplemental letters dated January 22, 1999, February 5,
1999, and March 17, 1999 provided additional clarifying information and
did not change the staff's initial no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 7, 1999.
Brief description of amendment: The amendment allows loading and
handling of spent fuel transfer and storage casks in the Trojan fuel
building.
Date of issuance: April 23, 1999.
Effective date: April 23, 1999.
Amendment No.: 199.
Facility Operating License No. NPF-1: The amendment changes the
Operating License.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9197). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 23, 1999.
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.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 27, 1999.
Brief description of amendment: This proposed amendment would allow
unloading of spent fuel transfer casks in the Trojan Fuel Building.
Date of issuance: April 23, 1999.
Effective date: April 23, 1999.
Amendment No.: 200.
Facility Operating License No. NPF-1: The amendment revises the
Operating License.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9198). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 23, 1999.
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.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: February 12, 1997.
Brief description of amendment: The amendment deletes the
Independent Spent Fuel Storage Installation area from the Permanently
Defueled Technical Specifications.
Date of issuance: May 5, 1999.
Effective date: May 5, 1999.
Amendment No.: 201.
Facility Operating License No. NPF-1: The amendment changes the
Permanently Defueled Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9196). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 5, 1999.
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 No. 50-272, Salem Nuclear
Generating Station, Unit No. 1, Salem County, New Jersey
Date of application for amendment: January 15, 1999, as
supplemented on March 31, 1999.
Brief description of amendment: The amendment allows a one-time
extension of the Technical Specification (TS) surveillance interval to
the end of fuel Cycle 13 (IR13) for certain TS
[[Page 27336]]
surveillance requirements (SRs). Specifically, the amendment extends
the surveillance interval in (a) SR 4.3.2.1.3 for the instrumentation
response time and sequence testing of each engineered safety features
actuation system (ESFAS) function; (b) SRs 4.8.2.3.2.f and 4.8.2.5.2.d
for service testing of the 125-volt DC and the 28-volt DC distribution
system batteries, respectively; (c) SR 4.8.2.5.2.c.2 for verification
of the condition of the 125-volt DC battery connections; (d) SR
4.8.3.1.a.1.a and 4.8.3.1.a.1.b for channel calibration and integrated
system functional test for containment penetration conductor
protection; (e) SR 4.1.2.2.c for verification that each automatic valve
in the reactivity control system flow path actuates on a safety
injection (SI) test signal; (f) SRs 4.3.1.1.1,Table 4.3-1, 4.3.2.1.1,
Table 4.3-2, 4.3.3.5, Table 4.36, and 4.3.3.7, Table 4.3-11 for the
channel calibration of containment water level-wide range, the manual
solid-state protection system (SSPS) functional input check, and the
ESFAS manual initiation channel functional test; (g) SR 4.5.1.d for
verification that each accumulator isolation valve opens automatically
on an SI test signal; (h) SR 4.5.2.e.1 for verification that each
automatic valve in the ECCS flow path actuates on an SI test signal,
(i) SR 4.7.6.1.d.2 for verification that the control room emergency air
conditioning system automatically actuates in the pressurization mode
on an SI test signal or control room intake high radiation test signal;
(j) SR 4.7.10.b for verification that each automatic valve in the
chilled water loop actuates on an SI signal; and (k) SR 4.8.1.1.2.d.7
which requires a test to verify that each emergency diesel generator
operates for at least 24 hours. The SRs are to be completed during the
next refueling outage (1R13), prior to returning the unit to Mode 4
(hot shutdown) upon outage completion. The amendment also makes some
administrative and editorial changes on some of the pages that will be
affected by the above SR interval extensions.
Date of issuance: May 4, 1999.
Effective date: May 4, 1999.
Amendment No.: 222.
Facility Operating License No. DPR-70: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6709). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 4, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
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: February 8, 1999.
Brief description of amendments: The amendments revise Technical
Specification 4.5.3.2.b to allow the option of using closed and
disabled automatic valves to provide the necessary isolation function
when performing safety injection and charging pump testing in Modes 4,
5, and 6 (i.e., hot shutdown, cold shutdown, and refueling) for low
temperature overpressurization protection.
Date of issuance: April 26, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 220 and 202.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14284). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
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: March 26, 1998.
Brief description of amendments: The amendments revise Technical
Specification 3/4.8.2.1, ``AC Distribution--Operating,'' to add
operability conditions and associated action statements for the 115-
volt vital instrument bus (VIB) D and inverter. The amendments complete
the recommended action from NRC Generic Letter 91-11, Resolution of
Generic Issues 48, ``LCOs for Class 1E Vital Instrument Buses,'' and
49, ``Interlocks and LCOs for Class 1E Tie Breakers,'' pursuant to 10
CFR 50.54(f), dated July 18, 1991.
Date of issuance: April 30, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 221 and 203.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 6, 1998 (63 FR
25117). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 30, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: March 1, 1999.
Brief description of amendment: The amendment revises the Ginna
Station Improved Technical Specifications battery cell parameters limit
for specific gravity (Surveillance Requirement (SR) 3.8.6.3 and SR
3.8.6.6).
Date of issuance: April 23, 1999.
Effective date: April 23, 1999.
Amendment No.: 74.
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14284). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: January 15,1999 (TS 98-07).
Brief description of amendments: The amendments change the
Technical specifications (TS) by adding a new action statement to TS
3.1.3.2, ``Position Indicating Systems--Operating,'' that eliminates
the need to enter TS 3.0.3 whenever two or more individual rod position
indications per bank may be inoperable. It also allows additional time
to determine the position of the non indicating rod(s).
Date of issuance: May 4, 1999.
Effective date: May 4, 1999.
Amendment Nos.: 244 and 235.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TS.
Date of initial notice in Federal Register: February 24, 1999 (64
FR
[[Page 27337]]
9201). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 4, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: November 3, 1998.
Brief description of amendment: The amendment makes changes to the
Technical Specifications to more clearly describe the emergency core
cooling system actuation instrumentation for the low pressure coolant
injection and core spray systems.
Date of Issuance: April 26, 1999.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 170.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6714). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated April 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: July 28, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications Section 4.6.2.2.1.b for Units 1 and 2 casing
cooling and outside recirculation spray pumps surveillance testing
criteria.
Date of issuance: April 22, 1999.
Effective date: April 22, 1999.
Amendment Nos.: 219 and 200.
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48272). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 22, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 23, 1998.
Brief description of amendment: The amendment revises Technical
Specification 3/4.5.1, ``Emergency Core Cooling Systems--
Accumulators,'' by increasing the allowed outage time with one
accumulator inoperable for reasons other than boron concentration
deficiencies from 1 hour to 24 hours. The corresponding Bases section
was also revised.
Date of issuance: April 27, 1999.
Effective date: April 27, 1999, to be implemented within 30 days
from the date of issuance.
Amendment No.: 124.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64127). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 27, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of no Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action.
[[Page 27338]]
Accordingly, the amendments have been issued and made effective as
indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see: (1) The
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By June 18, 1999, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Southern California Edison Company, et al., Docket No. 50-361, San
Onofre Nuclear Generating Station, Unit No. 2, San Diego County,
California
Date of application for amendment: April 24, 1999.
Brief description of amendment: This one-time temporary amendment
allows the facility to be outside the licensing basis regarding remote
shutdown capability of the shutdown cooling system as described in the
Updated Safety Analysis Report, Section 5.4.7.1.2, during the period of
the repair. The amendment is effective for 7 days from the date of
issuance or until the repair of the check valves is completed,
whichever occurs first.
Date of issuance: April 26, 1999.
Effective date: April 26, 1999, and is effective for 7 days from
the date of issuance or until the check valves repair is completed,
whichever occurs first.
Amendment No.: 152.
Facility Operating License Nos. NPF-10: This amendment approved a
one-time change to the design basis as described in the Updated Safety
Analysis Report.
Public comments requested as to proposed no significant hazards
consideration: No.
[[Page 27339]]
The Commission's related evaluation of the amendment, finding of
emergency circumstances, consultation with the State of California, and
final no significant hazards consideration determination are contained
in a Safety Evaluation dated April 26, 1999.
Attorney for Licensee: T.E. Qubre, Esquire, Southern California
Edison Company, P.O. Box 800, Rosemead, California 91770
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
NRC Section Chief: Stephen Dembek.
Dated at Rockville, Maryland, this 12th day of May 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-12494 Filed 5-18-99; 8:45 am]
BILLING CODE 7590-01-P