[Federal Register Volume 63, Number 77 (Wednesday, April 22, 1998)]
[Notices]
[Pages 19964-19989]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-10470]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 30, 1998, through April 10, 1998. The
last biweekly notice was published on April 8, 1998 (63 FR 17219).
[[Page 19965]]
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 May 22, 1998, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's
[[Page 19966]]
Public Document Room, the Gelman Building, 2120 L Street, NW.,
Washington, DC, by the above date. A copy of the petition should also
be sent to the Office of the General Counsel, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, and to the attorney for the
licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: March 14, 1997.
Description of amendment request: The proposed amendment would
delete license conditions which have been satisfied, revise others to
delete parts which are no longer applicable or to revise references,
and make editorial changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The initial conditions and methodologies used in the accident
analyses remain unchanged. The proposed changes do not change or
alter the design assumptions for the systems or components used to
mitigate the consequences of an accident. Therefore, accident
analyses results are not impacted.
The license conditions were one-time commitments that have been
satisfied. There are no physical changes to the facility, and all
operating procedures, limiting conditions for operation, limiting
safety system settings, and safety limits are unchanged. Removal of
these license conditions is appropriate and safe.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Many of the proposed changes delete references to items that
have been completed. The NRC required these items as a condition of
granting the license. Since they have been satisfied as intended,
deleting them is administrative.
None of the proposed changes affect the design or operation of
any system, structure, or component in the plant. The safety
functions of the related structures, systems, or components are not
changed in any manner, nor is the reliability of any structure,
system, or component reduced by the revised surveillance or testing
requirements. The changes do not affect the manner by which the
facility is operated and do not change any facility design feature,
structure, system, or component. No new or different type of
equipment will be installed. Since there is no change to the
facility or operating procedures, and the safety functions and
reliability of structures, systems or components are not affected,
the proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The remaining changes are editorial in nature and have no impact on
plant operation or design.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes to the Operating License are generally
administrative in nature and have no impact on the margin of safety
of any Technical Specification. There is no impact on safety limits
or limiting safety system settings. The changes do not affect any
plant safety parameters or setpoints. The operating license
conditions have been satisfied, as required. There are no changes to
the conditions themselves. Therefore, the proposed changes do not
involve a significant reduction in a margin of safety.
Therefore, based on the above evaluation, Commonwealth Edison
has concluded that these changes do not involve significant hazards
considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Stuart A. Richards.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: October 16, 1997.
Description of amendment request: The proposed amendment would add
an exemption from 10 CFR 70.24(a) to the Unit 1 license consistent with
the Unit 2 license.
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 change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The initial conditions and methodologies used in the accident
analyses remain unchanged. The proposed change does not change or
alter the design assumptions for the systems or components used to
mitigate the consequences of an accident. Therefore, accidents
analysis results are not impacted.
There are no physical changes to the facility, and all operating
procedures, limiting conditions for operation, limiting safety
system settings, and safety limits are unchanged.
The specific requirements for granting an exemption from 10 CFR
70.24(a) have been met. The request is authorized by law, will not
endanger life or property or the common defense and security, and is
in the public interest.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not affect the design or operation of
any system, structure, or component in the plant. The safety
functions of structures, systems, or components are not changed in
any manner, nor is the reliability of any structure, system, or
component reduced by the revised surveillance or testing
requirements. The change does not affect the manner by which the
facility is operated and does not change any facility design
feature, structure, system, or component. No new or different type
of equipment will be installed. Since there is no change to the
facility or operating procedures, and the safety functions and
reliability of structures, systems, or components are not affected,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
C. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change to the Operating License has no impact on
the margin of safety of any Technical Specification. There is not
[[Page 19967]]
impact on safety limits or limiting safety system settings. The
change does not affect any plant safety parameters or setpoints.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Therefore, based on the above evaluation, Commonwealth Edison
has concluded that the proposed change does not involve significant
hazards considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Stuart A. Richards.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: March 13, 1998, as supplemented March
30, 1998.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to allow any two auxiliary
feedwater (AFW) flow control valves to be inoperable concurrently for
up to 72 hours, provided the corresponding redundant flow control
valves and a pump in the other AFW train are operable.
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:
Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change would only alter the allowance for specific
AFW flow control valves to be inoperable. It would not affect any
operating limits, any plant operating conditions, or the physical
capability of any plant equipment. Therefore, it would not affect
the probability of any accident previously evaluated.
The proposed change would not reduce the AFW flow capability to
the steam generators during operation under the affected Action
Statement. It would allow more operational flexibility in plant
operation when two AFW flow control valves in the same train were
concurrently inoperable. The specified AOT [allowed outage time] of
72 hours would remain unchanged. Current TS allow continued
operation for 72 hours with one of the three AFW pumps inoperable,
or with one flow control valve in each train inoperable (provided
the corresponding redundant flow control valve and a pump in the
other pipe train are operable), but do not allow continued operation
with both valves in the same train inoperable. The proposed change
would allow any two valves to be inoperable, with the same provision
that the corresponding redundant flow control valve and a pump in
the other pipe train are operable.
Since, with the proposed change there would be no reduction in
the ability to provide AFW flow to either steam generator, operation
of the Facility in accordance with the proposed changes would not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
The changes do not alter the plant configuration (no new or
different type of equipment will be installed) or make changes in
the methods governing normal plant operation. The changes do allow
different sets of AFW flow control valves to be inoperable, however,
these changes retain a consistent level of AFW capability during
operation under the Action Statement. Therefore, the changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Therefore, operation of the Facility in accordance with the
proposed TS change would not create the possibility of a new or
different kind of accident from any previously evaluated.
Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed change would not reduce the AFW flow capability to
the steam generators during operation under the affected Action
Statement. It would allow more operational flexibility in plant
operation when two AFW flow control valves were concurrently
inoperable. The specified AOT of 72 hours would remain unchanged.
Therefore, operation of the Facility in accordance with the
proposed TS change would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: Cynthia A. Carpenter.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: March 27, 1998 (NRC-98-0033).
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.5.2, ``ECCS--Shutdown,'' and TS
3.5.3, ``Suppression Chamber,'' raising the minimum water level
required in the condensate storage tank (CST) to support the core spray
system (CSS) when the suppression pool (the normal supply for CSS) is
unavailable. The amendment would also eliminate incorrect information
concerning CST inventory reserved for the high pressure coolant
injection (HPCI) and reactor core isolation cooling (RCIC) systems. The
associated Bases are also revised.
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 changes will not affect the performance or reliability of
the Condensate Storage System which could lead to an accident
because the Condensate Storage Tank (CST) is not involved as an
initiator of any accident previously evaluated. The proposed change
meets the design standards of the Condensate Storage System by
providing assurance that sufficient water volume is available for
the Core Spray System. This change also removes [an] erroneous
discussion of water inventory for HPCI/RCIC Systems while in
Operating [Operational] Conditions 4 and 5. The removal of
information is acceptable since HPCI/RCIC Systems are not operable
in these modes and will therefore not increase the probability of an
accident. The increase in volume provides for vortex/air entrainment
avoidance in the Core Spray System and will not increase
consequences. Furthermore, the elimination of HPCI/RCIC information
will not increase consequences of an accident previously evaluated
because these systems are not credited for accident mitigation in
Operating [Operational] Conditions 4 and 5.
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 does not add or modify any equipment or
components related to the Condensate Storage System and will
therefore not create any new failure modes or common failure modes.
This proposed change raises the water level within the CST to ensure
sufficient water volume is maintained and updates the TS by removing
descriptive information with respect to CST water inventory for
HPCI/RCIC Systems
[[Page 19968]]
while in Operating [Operational] Conditions 4 and 5. The Condensate
Storage System will continue to operate as intended and as designed.
This change will therefore not create the possibility of a new or
different kind of accident.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed change increases the required CST water level to
provide at least 150,000 gallons of water available for the Core
Spray System while maintaining adequate submergence of the Core
Spray standpipe for avoiding vortex and air entrainment. As such,
the proposed change involves no reduction on any margin of safety.
Revision to TS Bases concerning discussion of reserve volume in CST
for HPCI and RCIC, does not alter the requirement for Core Spray or
Suppression Pool operability and does not involve a reduction in any
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: March 27, 1998 (NRC-98-0034).
Description of amendment request: The proposed amendment would
clarify a footnote in Technical Specification (TS) 3.5.1, ``ECCS--
Operating,'' and 3.5.2, ``ECCS--Shutdown,'' to indicate that a low
pressure coolant injection system loop may be considered operable
during alignment and operation for decay heat removal if it is capable
of being manually realigned and is not otherwise inoperable. The
associated Bases would also be revised.
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 changes involve actions required to realign the Low Pressure
Coolant Injection (LPCI) system for LPCI injection if LPCI is
required when operating in the Shutdown Cooling (SDC) mode. The
additional actions described involve resetting isolations and trips
which could occur prior to LPCI initiation. Resetting these logics
does not initiate any valve operation or pump start; the LPCI
initiation signals and interlocks remain in control of valve and
pump logic.
The equipment interlocks that provide the isolation signal for
the LPCI injection valves were designed to prevent drain down of the
Reactor Pressure Vessel (RPV) when in SDC. The injection valve
closure is the most conservative action in response to an RPV drain
event. The current TS acknowledges that operator action to realign
the suction path is necessary. The proposed change acknowledges that
operator action to reset injection valve logic and pump trips is
necessary. The time required to realign LPCI is not significantly
different than the existing actions to realign the suction path.
No changes in either system design or operating strategies will
be made as a result of these changes, thus no opportunity exists to
increase the probability or consequences of a previously analyzed
accident. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of a
previously evaluated accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The manual realignment of the LPCI system from SDC following an
isolation signal does not affect the accident analysis described in
Chapter 15 of the UFSAR [updated final safety analysis report]. No
new limiting single failure has been identified as a result of the
proposed changes. The possibility of a new or different kind of
accident from those previously analyzed will not be created by the
change to the TS footnote or Bases, because the proposed change
merely clarifies the actions necessary to realign the LPCI system.
The time required to realign the system is not significantly
different than the time necessary to realign the suction path.
Therefore, no new or different types of failures or accident
initiators are introduced by the proposed changes.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed change described above affects the plant's ability
to enter Operational Conditions 3, 4, and 5, and to achieve and
maintain COLD SHUTDOWN conditions when shutting down the plant. The
proposed change in combination with existing restrictions within the
TS provide assurance that there is no credible mechanism to inhibit
running the LPCI system. The minor additional operator action
required to realign LPCI from SDC requires minimum time and effort
considering controls for each division are located on their
respective control panel. As a result of this change, there will be
no changes in either system design or operating strategies because
the proposed changes merely clarify existing TS requirements and
actions necessary to meet TS requirements. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: April 2, 1998 (NRC-98-0057).
Description of amendment request: The proposed amendment would
permit entering Operational Conditions 1 and 2 prior to completion of
Surveillance Requirements for the primary containment hydrogen and
oxygen monitors in order to establish the conditions necessary (inerted
containment) to properly perform the calibrations. The amendment would
also increase the frequency of the calibration for the oxygen monitors
from every 18 months to quarterly in accordance with vendor
recommendations and correct the nomenclature for the hydrogen and
oxygen monitors in tables 3.3.7.5-1 and 4.3.7.5-1.
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 changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change will permit delaying the performance of
calibrations of the hydrogen and oxygen monitors until after the
containment is inerted following a plant startup. The proposed
change will also increase the calibration frequency for the oxygen
monitors from once per 18 months to once per quarter, and change the
nomenclature for the hydrogen and oxygen monitors.
The primary containment hydrogen and oxygen monitors are passive
instruments that provide indication to control room operators of
hydrogen and oxygen concentration in the primary containment.
Because they perform only a passive monitoring function, the
[[Page 19969]]
hydrogen and oxygen monitors are not associated with the initiation
of any previously evaluated accident. The indication provided by the
monitors is used by the control room operators to ensure oxygen
concentration remains below limits and to make decisions regarding
the use of the Combustible Gas Control System, if necessary. The
allowance to permit entry into applicable operational conditions
before calibration ensures that the conditions (nitrogen
environment) are appropriate for accurate calibration of the
instrument. Delaying the calibration does not cause the instrument
to cease to function. Calibrations verify and adjust, as necessary,
the accuracy of the instrument to compensate for drift that may
occur since the last calibration. Thus, even with a delayed
calibration, the instruments still would provide valuable
information to the operators. Consequently, this change will not
involve a significant increase in the consequences of a previously
evaluated accident because the monitors will still function and
provide meaningful information until the calibration is completed.
The change to reduce the interval for calibration of the oxygen
monitors from once per 18 months to once per quarter provides
increased assurance of monitor accuracy and is consistent with the
manufacturer's recommendations. Therefore, because this instrument
is not associated with the initiation of an accident and the change
improves the functionality of the instrument, the probability and
consequences of previously evaluated accidents are not significantly
affected.
The change in nomenclature is editorial, and, as such does not
affect the probability or consequences of a previously evaluated
accident.
2. The changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
As discussed above, the hydrogen and oxygen monitors are
passive, indication-only instruments which provide information to
control room operators. The proposed changes do not introduce a new
mode of operation or involve a physical modification to the plant.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The changes do not involve a significant reduction in the
margin of safety.
The proposed changes involve the containment hydrogen and oxygen
monitors which do not affect any parameters or assumptions used in
the calculation of any safety margin with regard to Technical
Specification Safety Limits, Limiting Safety System Settings,
Limiting Control Settings or Limiting Conditions for Operation, or
other previously defined margins for any structure, system, or
component. 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: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, (BVPS-1 and BVPS-2),
Shippingport, Pennsylvania
Date of amendment request: March 16, 1998.
Description of amendment request: The proposed amendment would
revise technical specification (TS) Table 4.3-1 to add footnote 6 to
the channel calibration requirement for all instrument channels that
are provided with an input from neutron flux detectors. Footnote 6
provides that neutron detectors may be excluded from channel
calibrations. Additional changes are proposed for BVPS-1 to provide
consistency between BVPS-1 and BVPS-2. These additional changes would
add channel calibration requirements to BVPS-1 TS Table 4.3-1 items
2.b. (Power Range, Neutron Flux, Low Setpoint), 5. (Intermediate Range,
Neutron Flux), 6. (Source Range, Neutron Flux (Below P-10), and 23.
(Reactor Trip System Interlocks P-6, P-8, P-9, and P-10). Furthermore,
changes would be made to correct page numbers in the BVPS-2 Index and
to add corresponding changes to the bases for both units.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The protection systems will continue
to function in a manner consistent with the plant design basis. The
proposed changes will not affect any of the analysis assumptions for
any of the accidents previously evaluated. The proposed changes will
not affect the probability of any event initiators nor will the
proposed changes affect the ability of any safety-related equipment
to perform its intended function. There will be no degradation in
the performance of nor an increase in the challenges imposed on
safety-related equipment assumed to function during an accident.
There will be no change to normal plant operating parameters or
accident mitigation capabilities. Therefore, the proposed changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
There are no hardware changes associated with this license
amendment nor are there any changes in the method by which any
safety-related plant system performs its safety function. The normal
manner of plant operation is unchanged.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). Maintaining the SAL preserves the margin of safety.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. 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: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of amendment request: March 17, 1998.
Description of amendment request: The proposed amendment would
revise Action 34 of technical specification (TS)
[[Page 19970]]
Table 3.3-3, ``Engineered Safety Feature Actuation System
Instrumentation.'' Action 34 applies to Functional Units 6.b., ``Grid
Degraded Voltage (4.16 kV Bus),'' and 6.c. ``Grid Degraded Voltage (480
v Bus).'' The proposed revision would require that with one channel
inoperable, the inoperable channel be placed in the tripped condition
within one hour; otherwise, the applicable action statement(s) for the
associated emergency diesel generator made inoperable by the degraded
voltage start instrumentation be entered immediately. The proposed
revision would also require that with two channels inoperable, at least
one of the two channels be restored to operable status and the other
channel be placed in the tripped condition within one hour; otherwise,
the associated emergency diesel generator shall be declared inoperable
and its applicable action statement(s) shall be entered. In addition,
corresponding changes would be made to the bases for TS 3/4.3.2 and the
BVPS-2 Index would be revised to reflect changed page numbers.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Engineered Safety Feature Actuation System (ESFAS) will continue
to function in a manner consistent with the plant design basis. The
proposed change will not affect any of the analysis assumptions for
any of the accidents previously evaluated. The proposed changes will
not affect the probability of any event initiators. There will be no
change to normal plant operating parameters. The emergency bus
degraded voltage protection system is utilized for accident
mitigation and is not considered to be the source of accidents
previously evaluated.
Implementation of the proposed changes will now provide viable
corrective actions which do not significantly increase the
probability of failure of safety related equipment to perform its
intended function. The proposed Action 34 permits a one hour time
frame before the affected diesel generator(s) is required to be
declared inoperable. This one hour period allows for repairs of most
failures and takes into account the low probability of an event
which would require the degraded voltage protection system to
function. If adequate protection is not restored within this one
hour period, the diesel generator(s) allowable outage time is
invoked. The diesel generator(s) allowable outage time has been
previously evaluated and determined to be an acceptable period of
time during which plant operation may continue without an emergency
backup power source. The loss of emergency bus degraded voltage
protection is similar to the loss of the ability of an emergency
diesel generator to provide electrical power to the safety related
loads on the emergency buses. In both situations, a loss of offsite
power, due to a total loss or a degraded condition, will result in
the safety related loads not being capable of mitigating a design
basis accident. The proposed changes to the Index page are
administrative in nature and do not affect plant safety.
Therefore, the proposed changes do not result in a significant
increase in probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The emergency bus degraded voltage protection system is utilized
for accident mitigation. The proposed changes will now provide
viable corrective actions which do not result in a change in the
manner in which the emergency bus loads are protected from a
degraded voltage condition. These changes do not alter the function
of the degraded voltage protection system. The proposed changes will
continue to require that at least one of the two redundant 4160 volt
or 480 volt emergency buses is protected from a degraded voltage
condition assuming a single active failure of the opposite emergency
bus degraded voltage protection system. This action will ensure that
at least one train of engineered safety feature (ESF) equipment is
not damaged due to a sustained bus undervoltage condition. The
proposed addition of the requirement to enter the action statement
for the inoperable diesel generator, if the one hour requirements of
Action 34 cannot be met, will ensure that adequate compensatory
actions to assure plant safety are taken. These requirements include
the demonstration of the operability of the A.C. offsite sources by
performing a specific surveillance within one hour and at least once
per eight hours thereafter. If both diesel generators are
inoperable, at least one diesel generator must be restored to
operable status within two hours or the plant must be placed in cold
shutdown within the following 36 hours.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of these changes. There
will be no adverse effect or challenges imposed on any safety-
related system as a result of these changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The margin of safety is not significantly reduced because the
A.C. electrical power sources will continue to provide sufficient
capability, redundancy, and reliability to ensure availability of
necessary power to ESF systems. The ESF systems will continue to
function, as assumed in the safety analyses, to ensure that fuel,
reactor coolant system and containment design limits are not
exceeded. The proposed revisions to Action 34 will continue to
require that at least one of the two redundant 4160 volt or 480 volt
emergency buses is protected from a degraded voltage condition
assuming a single active failure of the opposite emergency bus
degraded voltage protection system. This action will ensure that at
least one train of ESF equipment is not damaged due to a sustained
bus undervoltage condition. The emergency loads, which are powered
from that train of emergency buses, will continue to be available to
perform their safety related functions. If the one hour requirements
of Action 34 cannot be met, the affected emergency diesel generator
will be declared inoperable. This will ensure that adequate
compensatory actions to ensure plant safety are taken. The loss of
emergency bus protection from a degraded voltage condition is
similar to the loss of the ability of an emergency diesel generator
to provide electrical power to the safety related loads on the
emergency buses. In both situations, a loss of the offsite power,
due to a total loss or a degraded condition, will result in the
safety related loads not being capable of mitigating a design basis
accident.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: March 23, 1998.
Description of amendment request: The proposed amendment modifies
Section 3.1.2 of the Technical Specifications (TS) to incorporate new
pressure/temperature limits regarding reactor vessel pressurization
heatup, cooldown, and inservice leak and hydrostatic leak test
limitations in accordance with 10 CFR 50, Appendix G. These new limits
would be applicable through the period of 17.7 effective full power
years (EFPY) of operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 19971]]
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated. The design basis event related to this change is
nonductile failure of the reactor coolant pressure boundary. The
updated pressure/temperature limits have been established in
accordance with the requirements of 10 CFR 50, Appendix G. Revision
of these curves for an applicability period of 17.7 EFPY is based on
maintaining the required design margin. Operation of the facility in
accordance with the proposed amendment provides assurance of
protection against nonductile failure of the reactor coolant
pressure boundary for operation through 17.7 EFPY. Therefore,
operation in accordance with the proposed amendment does not involve
a significant increase in the probability of occurrence or
consequences of an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. The design basis
event related to the change is nonductile failure of the reactor
coolant boundary. The proposed amendment provides assurance of
protection against nonductile failure of the reactor coolant
boundary for operation through 17.7 EFPY and is unrelated to the
possibility of creating a new or different kind of accident.
3. Operation of the facility in accordance with the proposed
amendment would not involve any reduction in a margin of safety
since the design methodology has maintained the existing margins.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 27, 1997.
Description of amendment request: The proposed amendment, included
as part of the proposed conversion from the current Technical
Specifications (TS) to improved TS, would revise the Limiting
Conditions for Operation in the event that one 250 V DC electrical
power subsystem is inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The DC electrical power sources are used to support mitigation
of the consequences of an accident; however, they are not considered
the initiator of any previously analyzed accident. The proposed
change merely provides direction to the operator to declare
equipment associated with a 250 V DC electrical power subsystem
inoperable if the subsystem becomes inoperable. This provides
assurance that all affected features are immediately recognized as
incapable of performing their safety functions, and requires
immediate actions equivalent to those determined appropriate in the
Technical Specifications for the affected features. Therefore, the
proposed change does not involve an increase in the probability or
consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not introduce a new mode of plant
operation and does not involve physical modification to the plant.
Therefore, the possibility of a new or different kind of accident
from any accident previously evaluated is not created.
3. Does this change involve a significant reduction in a margin
of safety?
This change does not involve a significant reduction in a margin
of safety, since the proposed change results in establishing the
level of safety for the loss of a 250 V DC electrical power
subsystem equivalent to the level of safety that exists in the
Technical Specifications for components and systems that are
supplied by the 250 V DC electrical power system.
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: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Project Director: John N. Hannon.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: March 2, 1998.
Description of amendment request: The proposed changes would revise
the frequency for the performance of specific surveillances associated
with the emergency diesel generators (EDGs) and delete the requirements
contained in the current Technical Specifications for accelerated
testing whenever the number of valid test failures associated with the
EDGs is met or exceeded. In addition, the special requirements for
reporting valid or invalid EDG failures would be deleted.
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 affecting the
ability of the EDGs to mitigate the consequences of an accident.
Industry experience has indicated that excessive testing
requirements have proven to be a contributor to increased EDG
unavailability and equipment degradation. Removing inappropriate
testing requirements increases EDG reliability and enhances the
ability of EDGs to mitigate the consequences of an accident.
Implementing the maintenance rule in accordance with 10 CFR 50.65,
Regulatory Guide 1.160, and NUMARC 93-01 for the EDGs provides
additional assurance that high EDG performance and availability will
be maintained.
Deleting the special reporting requirements from the Technical
Specifications is an administrative change that does not affect the
ability of the EDGs to perform their specified safety function.
North Atlantic will continue to notify the NRC of significant EDG
failures in accordance with the provisions of 10 CFR 50.72 and
50.73.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed changes do not alter the ability of the EDGs 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 have no
impact on component or system interactions, or the plant design
basis. Instrumentation setpoints, starting,
[[Page 19972]]
sequencing and loading functions associated with the EDGs are not
affected by the proposed changes. Furthermore, combining the
implementation of the maintenance rule program with the proposed
amendment will enhance both the availability and the performance of
the EDGs.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. The proposed changes do 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 Technical Specification required
safety limits or safety system settings that would adversely affect
plant safety. The proposed changes do not affect the EDG's ability
to ensure that sufficient power is available to supply the safety
related equipment required for: 1) the safe shutdown of the
facility, and 2) the mitigation and control of accident conditions
within the facility. In addition, the proposed changes do not affect
the EDG's ability to ensure that: 1) the facility can be maintained
in a shutdown or refueling condition for extended periods of time,
and 2) sufficient instrumentation and control capability is
available for monitoring and maintaining the unit status.
EDG reliability and availability are expected to be improved by
the proposed changes. Eliminating excessive testing requirements can
improve safety by reducing challenges to plant systems and reducing
equipment wear and degradation. While the proposed changes affect
surveillance intervals there are no changes to the methods used to
perform the surveillances. The surveillances will continue to
demonstrate the ability of the EDGs to perform their intended
function of providing electrical power to the emergency safety
systems needed to mitigate design basis transients and accidents.
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 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: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Cecil O. Thomas.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: March 5, 1998.
Description of amendment request: The proposed changes would revise
the Seabrook Station Radiological Effluent Technical Specifications
(TS) and Administrative Controls section of the Technical
Specifications, as authorized by NRC Generic Letter (GL) 89-01,
``Implementation Of Programmatic Controls For Radiological Effluent
Technical Specifications (RETS) 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.'' The proposed amendment would incorporate
programmatic controls in the TSs for radioactive effluents and for
environmental monitoring conforming to the applicable regulatory
requirements and would relocate the existing procedural details of the
current RETS to the Offsite Dose Calculation Manual (ODCM). Procedural
details associated with solid radioactive wastes would be relocated to
the Process Control Program (PCP).
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 (SSCs) 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.
Incorporation of programmatic controls for RETS in TSs 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 analyses, 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 changes do not create the possibility of a new
or different kind of accident from any previously analyzed.
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 effluents. Therefore,
the proposed changes do not create the possibility of a new or
different kind of accident from any previously analyzed.
3. The proposed changes do 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 Technical Specification required
safety limits or safety system settings that would adversely affect
plant safety. 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 any margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 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: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Cecil O. Thomas.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: March 23, 1998.
Description of amendment request: The proposed change would revise
the Seabrook Station Technical Specifications (TSs) to add a new TS
3.0.5 that would provide an exception to TSs 3.0.1 and 3.0.2 to allow
the performance of required testing to demonstrate the operability of
the equipment being returned to service or the operability of other
equipment.
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 design basis accidents are not affected by the proposed
administrative changes.
[[Page 19973]]
Specification 3.0.5 provides the administrative controls to
ensure the time the equipment is returned to service in conflict
with the requirements of the ACTIONS is limited to the time
absolutely necessary to perform the allowed required testing.
Specification 3.0.5 was incorporated in NUREG-1431, ``Standard
Technical Specifications--Westinghouse Plants,'' (as modified by
approved Technical Specification Task Force (TSTF) generic change
Traveler TSTF-165), to address these, and other similar situations,
that conflict with the requirements with the ACTIONS when equipment
is returned to service. Specification 3.0.5 does not provide time to
perform other preventative or corrective maintenance.
Inclusion of Specification 3.0.5 into the Seabrook Station
Technical Specifications will provide operational flexibility with
the restrictive compliance requirements of the other Applicability
Specifications (3.0.1 and 3.0.2) and allow the performance of post-
maintenance/surveillance activities to facilitate returning
equipment to service or to allow other equipment to be tested.
Therefore, inclusion of Specification 3.0.5 into the Seabrook
Station Technical Specifications enhances plant safety by minimizing
the potential for plant trip and/or transients. A qualitative risk
assessment concerning returning components to service for post-
maintenance testing was performed and concluded that the
configurations allowed by Specification 3.0.5 have a negligible
effect on the Seabrook Station risk profile. The components involved
will have either completed calibration or maintenance, and can
reasonably be expected to be able to perform their required safety
function when returned to service for testing purposes. Therefore,
the proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed changes do not introduce new features or modify
plant structures, systems and components or procedures that could
possibly affect station operations under normal or abnormal
conditions, thus, the potential for an unanalyzed accident is not
created. The proposed administrative changes have no adverse affect
on the safety limits or design basis accidents. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
There are no changes being made to the Technical Specification
safety limits or safety system settings that would adversely affect
plant safety. The changes do not affect the operation of structures,
systems or components (SSCs) nor do they introduce administrative
changes to plant procedures that could affect operator response
during normal, abnormal or emergency situations. Inclusion of
Specification 3.0.5 into the Seabrook Station Technical
Specifications enhances plant safety by minimizing the potential for
plant trip and/or transients by allowing equipment to be returned to
service. A qualitative risk assessment concerning the return of
components to service for post-maintenance testing was performed and
concluded that the configurations allowed by Specification 3.0.5
have a negligible effect on the Seabrook Station risk profile. The
components involved will have either completed calibration or
maintenance, and can reasonably be expected to be able to perform
their required safety function when returned to service for testing
purposes. 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 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: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Cecil O. Thomas.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: March 27, 1998.
Description of amendment request: The proposed change would revise
Technical Specification (TS) 3.7.6, ``Control Room Emergency Makeup Air
and Filtration (CREMAFS).'' The proposed change would modify the
existing required action when both trains of CREMAFS are inoperable in
Modes 5 and 6 by eliminating the restriction of suspending positive
reactivity changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes have no impact on the probability or
consequences of an accident previously evaluated in the UFSAR. The
control room ventilation systems are support systems which have a
role in the detection and mitigation of accidents but do not
contribute to the initiation of any accident previously evaluated.
The removal of the positive reactivity addition restriction in Mode
5 and 6 has no impact on the course of any accidents previously
evaluated. There are no presently evaluated positive reactivity or
boron dilution accidents that credit the CREMAFS to mitigate its
consequences or provide radiological protection. The positive
reactivity restriction is overly restrictive in that it does not
allow cooldown below 200 deg. F when Mode 5 is entered as a result
of both trains of CREMAFS being inoperable nor does it allow Reactor
Coolant System temperature to vary.
The restriction is also redundant to Technical Specification
3.1.1.2 ``Reactivity Control Systems Shutdown Margin-Tavg
less than or equal to 200 deg. F'' in Mode 5 and Technical
Specification 3.9.1 ``Refueling Operations Boron Concentration'' in
Mode 6. Technical Specification 3.1.1.2 action, with shutdown margin
less than the limit specified in the Core Operating Limits Report or
with the Reactor Coolant System boron concentration less than 2000
ppm boron, requires immediate and continued boration until the
restoration of the required shutdown margin or boron concentration.
Similarly, Technical Specification 3.9.1 actions require suspension
of core alterations or positive reactivity changes in addition to
immediate and continued boration until the restoration of the
required shutdown margin (Keff) or boron concentration
while in Mode 6. Sufficient shutdown margin ensures that (1) the
reactor can be made subcritical from all operating conditions, (2)
the reactivity transients associated with the postulated accident
conditions are controllable within acceptable limits and (3) the
reactor will be maintained sufficiently subcritical to preclude
inadvertent criticality in the shutdown condition. The above
referenced reactivity control system specifications provide the
necessary protection for postulated reactivity addition accident
conditions. Therefore, modifying the Technical Specification action
that requires the suspension of positive reactivity changes and core
alterations with both trains of the CREMAFS inoperable does not
involve a significant increase in the consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed change that removes the positive reactivity
addition restriction in Mode 5 and 6 does not create the possibility
of a new accident nor does it create the possibility of a different
kind of accident previously evaluated. There are no presently
evaluated positive reactivity or boron dilution accidents that
credit the CREMAFS to mitigate its consequences or provide
radiological protection. The addition of positive reactivity during
the above described situation is overly restrictive and furthermore
redundant to Technical Specification 3.1.1.2 ``Reactivity Control
Systems Shutdown Margin-Tavg less than or equal to
200 deg. F'' in Mode 5 and Technical Specification 3.9.1 ``Refueling
Operations Boron Concentration'' in Mode 6. The above referenced
reactivity control system specifications provide the necessary
protection for postulated reactivity addition accident conditions.
Therefore, modifying the Technical Specification action that
requires the suspension of positive
[[Page 19974]]
reactivity changes and core alterations with both trains of the
CREMAFS inoperable does not create the possibility of a new or
different accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The changes being proposed do not revise equipment design or
operation nor do they make changes to Technical Specification
required safety limits or safety system settings. In addition, they
do not alter the environmental conditions which are to be maintained
in the control room during normal operation and following an
accident and they do not revise the accident analyses. Therefore,
the proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 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: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Cecil O. Thomas.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: April 3, 1998.
Description of amendment request: The proposed change would revise
the Seabrook Station Technical Specifications (TSs) with administrative
changes to support phased implementation of 24-month fuel cycle
surveillance interval extensions. Specifically, the proposed change
would: (1) provide wording changes in the Bases Section of TS 4.0.2
necessary to support 24-month surveillance interval extensions, (2)
revise TS 4.0.5.b to provide revised terminology for inservice
inspection and testing activities and their associated frequencies, (3)
revise TS Table 1.1 to clarify current and future refueling intervals
and their associated surveillance requirements and frequencies, and (4)
delete the ``during shutdown'' restriction from the performance
requirements of certain surveillance requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The design basis accidents are not affected by the proposed
editorial and administrative changes. The proposed changes do not
change the level of programmatic controls or the procedural details
currently in place. Furthermore, these changes have no adverse
affect to the safe operation of the station. Performance of certain
maintenance and testing activities during conditions or modes other
than shutdown will be evaluated by North Atlantic to ensure proper
regard to their effect on safe operation of the plant is given prior
to conduct of a particular surveillance, or portion thereof.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed changes do not introduce new features or modify
plant structures, systems and components or procedures that could
possibly affect station operations under normal or abnormal
conditions, thus, the potential for an unanalyzed accident is not
created. Performance of maintenance and testing activities on-line,
as well as shutdown, are controlled by North Atlantic's procedures
and policies to perform reviews and assessments of these activities
to determine the affect on safe operation of the facility. The
proposed editorial and administrative changes have no adverse affect
on the safety limits or design basis accidents. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
There are no changes being made to the Technical Specification
safety limits or safety system settings that would adversely affect
plant safety. The changes do not affect the operation of structures,
systems or components nor do they introduce administrative changes
to plant procedures that could affect operator response during
normal, abnormal or emergency situations. Performance of certain
maintenance and testing activities during conditions or modes other
than shutdown will be evaluated by North Atlantic to ensure proper
regard to their effect on safe operation of the plant is given prior
to conduct of a particular surveillance, or portion thereof.
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 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: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Cecil O. Thomas.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: April 1, 1998.
Description of amendment request: The proposed revision to the
Millstone Unit 3 licensing basis would add a new sump pump subsystem to
address groundwater inleakage through the containment basemat.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with 10
CFR 50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10 CFR 50.92(c) are not
satisfied. The proposed revision does not involve an SHC because the
revision would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The current FSAR [Final Safety Analysis Report] credits the
waterproof membrane for assuring that groundwater inleakage is not
significant and would have no impact on safety related structures
and components. However, degradation of the waterproof membrane has
been detected, and it is now concluded that groundwater inleakage
can be significant in that it could affect the operability of the
RSS [recirculation spray system] pumps. The original plant design
had only nonsafety-related RSS sump pumps available for pumping the
groundwater from the RSS sumps. These pumps are not powered from the
emergency busses and would not be accessible during a design basis
LOCA [loss-of-coolant accident].
Thus, it is assumed that they would not be available to mitigate
a design basis accident. Two independent safety-related air-driven
sump pumps have been installed to eliminate the potential for
groundwater inleakage that would affect the RSS pumps.
Air-driven sump pumps have been installed with the air supply
line routed to a connection outside the ESF [engineered safety
features] building. This allows the installation of an air
compressor in an area that is accessible during a design basis
accident such as a LOCA. Two air compressors have been staged in
designated locations, and will be maintained and
[[Page 19975]]
periodically tested to ensure their availability. Periodic testing
of the sump pumps will also be performed. The surveillance
requirements have been incorporated into the Technical Requirements
Manual.
EOP [Emergency Operating Procedure] 35-ES1.3 has been modified
to add a step to install the compressors and start the sump pumps.
It is estimated that these sump pumps would be needed approximately
ten hours after a design basis accident. Thus, there is sufficient
time for the operators to perform this action. Since sufficient time
is available, the action has been incorporated into procedures and
the environmental conditions allow access to the area, it is
concluded that credit for operator action can be taken.
Thus, the new system is single failure proof and meets the
requirements of Standard Review Plan 3.4.1 which stated the
following:
``If safety-related structures are protected from below-grade
groundwater seepage by means of a permanent dewatering system, then
the system should be designed as a safety-related system and meet
the single failure proof criterion.''
This provides assurance that the RSS pumps and other safety-
related structures and components will perform the required safety
function as assumed in the accident analysis.
The current nonsafety-related RSS sump pump system will continue
to provide protection from groundwater inleakage during normal
operation. Thus, there is no impact on the probability of occurrence
of a transient because of equipment or structural failure due to
groundwater inleakage. In addition, the new safety-related RSS sump
pump system provides additional assurance that groundwater inleakage
would not affect structures or equipment during an extended loss of
offsite power or a design basis accident. Thus, it is concluded that
there is no impact on the probability of occurrence of any
previously evaluated accident.
The change results in the use of the new air-driven sump pumps
to remove groundwater in-leakage from the RSS cubicles. To preclude
the possibility for radiological contamination of the groundwater,
all sources of liquid radiological contamination to the sumps have
been eliminated. The RSS cubicle floor drains leading to Sumps 7A/7B
have been plugged. Drains from equipment determined not to be a
potential source of radiological contamination continue to drain to
Sumps 7A/7B (sources include CCP [component cooling water] and
Service Water relief valves) and are covered with splash guards to
prevent the entrance of contaminated spray. The Hydrogen Recombiner
area floor drains and the drain from the PASS [post accident
sampling system] sample sink, all of which are nonsafety-related,
have been isolated from the indirect waste receptor which drains to
Sump 7B. Sumps 7A and 7B have been cleaned and the existing
nonsafety-related sump pumps replaced to remove any existing
residual contamination. The nonsafety-related pumps (3DAS-P8A/B)
discharge to ESF Building sump 3DAS-SUMP 10. To preclude any
potential siphoning from the potentially contaminated Sump 10 back
to Sumps 7A/7B, the lines of the existing nonsafety-related pumps
have been shortened to discharge above the water level in Sump 10.
The walls of Sumps 7A/7B have been extended to protect from a
Limited Passive Failure and Pipe Break in the RSS cubicles. The
expected flooding height is 6.6 inches [ ]. The sump cubicle height
was extended to 3 ft. above the cubicle floor, well above this
height. The sumps are covered with a vented hood to protect from
pipe break spray and miscellaneous overhead leaks to further assure
the sumps remain isolated from potentially contaminated RSS system
fluids.
The existing SLCRS [supplementary leak collection and release
system] boundary has been extended to the isolation valves located
outside of the ESF building. Additionally, when the sump level is
reduced while using the air driven pump, the pumps are designed to
prevent air from being discharged through the pump discharge outside
of the ESF building.
Thus, use of the new sump pumps would not affect the offsite
doses following a design basis accident.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The current nonsafety-related RSS sump pump system will continue
to provide protection from groundwater inleakage during normal
operation. This will continue to provide assurance there is no
potential for a transient because of equipment or structural failure
due to groundwater inleakage. In addition, the new safety-related
RSS sump pump system provides additional assurance that groundwater
inleakage would not affect structures or equipment during an
extended loss of offsite power or a design basis accident.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The current FSAR credits the waterproof membrane for assuring
that groundwater inleakage is not significant and would have no
impact on safety related structures and components. However,
degradation of the waterproof membrane has been detected and it is
now concluded that groundwater inleakage can be significant in that
it could affect the operability of the RSS pumps. Original design
had only nonsafety-related RSS sump pumps available for pumping the
groundwater from the RSS sumps. These pumps are not powered from the
emergency busses and would not be accessible during a design basis
LOCA. Thus, it is assumed that they would not be available to
mitigate a design basis accident. Two independent safety-related
air-driven sump pumps have been installed to eliminate the potential
for groundwater inleakage that would affect the RSS pumps. The new
system is single failure proof and meets the requirements of
Standard Review Plan 3.4.1.
Use of the new system requires operator action to install pre-
staged air compressors to provide power for the new air-driven sump
pumps. It is estimated that these sump pumps would be needed
approximately ten hours after a design basis accident. Thus, there
is sufficient time for the operators to perform this action. Since
sufficient time is available, the action has been incorporated into
procedures and the environmental conditions allow access to the
area, it is concluded that credit for operator action can be taken.
With credit for the new single failure proof air-driven sump
pumps and operator action to install pre-staged compressors to
provide power for the pumps, the new subsystem provides the required
assurance that the RSS pumps will not be affected by groundwater
inleakage. Thus, it is concluded that the RSS pumps would be
operable for long term accident mitigation and there is no impact on
the margin of safety as defined in the basis of the Emergency Core
Cooling Technical Specifications or any other Technical
Specification.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is determined
that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: Phillip F. McKee.
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 amendment request: December 23, 1997.
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 to revise Technical Specification (TS)
3/
[[Page 19976]]
4.7.11, Table 3.7-1, ``Maximum Allowable Power Range Neutron Flux High
Setpoint With Inoperable Steam Line Safety Valves.'' The power range
(PR) neutron flux high setpoints would be changed based on revised
calculational methodologies for 1, 2, or 3 inoperable MSSVs per steam
generator (SG). The proposed TS change would lower the PR neutron flux
high setpoints when 2 or 3 MSSVs are inoperable per loop such that the
maximum power level allowed would be within the heat removing
capability of the remaining operable MSSVs. Although the method for
calculating the maximum power level allowed when one MSSV per loop is
inoperable has been revised, the results have not and the limit remains
the same. The associated Bases would also be revised.
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 reduction in the power range (PR) neutron flux high setpoint
Technical Specification (TS) values does not initiate an accident.
Technician adjustments to lower the PR neutron flux high setpoints
could cause a reactor trip (RT). However, this action is already a
TS requirement. Thus, reducing the TS setpoint values from their
current values will not change the requirement for a technician to
adjust the setpoints downward when main steam safety valves (MSSVs)
become inoperable, and therefore, will not increase the probability
of a RT.
The reduction of the setpoints assures that the consequences of
an accident when the MSSVs are inoperable are not affected by
assuring that the MSSVs will continue to prevent overpressure of the
main steam leads and steam generators (SGs) and remove adequate heat
for the reactor coolant system.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Reduction of the PR neutron flux high setpoints does not change
the method by which any safety-related system performs the function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
With the reduction in the PR neutron flux high setpoints for
inoperable MSSVs, the MSSVs will still prevent SG pressure from
exceeding 110 percent of SG design pressure in accordance with the
ASME code. The change is conservative. The conclusions for the Final
Safety Analysis Report Update accident analyses are unaffected by
the change, remain valid, and provide margin.
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 requests involve no significant hazards consideration.
Local Public Document Room Location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: February 6, 1998.
Description of amendment request: The proposed changes would revise
the Reactor Protection System (RPS) Normal Supply Electrical Protection
Assembly (EPA) Undervoltage Trip setpoint to reflect a reanalysis of
the most limiting applied load minimum voltage requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed setpoint change evaluated in Section III does not
involve any physical changes to the plant, does not alter the way
these systems function, and will not degrade the performance of the
plant safety systems. The proposed instrument setpoint changes
ensures that plant safety limits are not exceeded for the most
limiting voltage requirements. The type of testing and the
corrective actions required if the subject surveillances fail
remains the same. The proposed changes do not adversely affect the
reliability of these systems or affect the ability of the systems to
meet their design objectives. A historical review of surveillance
test results supports these conclusions.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed setpoint change evaluated in Section III does not
modify the design or operation of the plant, therefore, no new
failure modes are introduced. The proposed instrument setpoint
change ensures that plant safety limits are not exceeded for the
most limiting voltage requirements. No changes are proposed to the
type and method of testing performed. A historical review of
surveillance test results supports these conclusions.
3. Involve a significant reduction in a margin of safety.
The proposed setpoint change evaluated in Section III results in
minimal impact on system reliability in the interval between
surveillance tests. This is based on the redundant design of the
evaluated systems. A review of past surveillance history has shown
no evidence of failures which would significantly impact the
reliability of these systems. Operation of the plant remains
unchanged by this proposed setpoint change. The assumptions in the
Plant Licensing Basis are not adversely impacted. Therefore, the
proposed changes do not result in 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: February 6, 1998.
Description of amendment request: The proposed changes would allow
reactor coolant system pressure tests to be conducted in Cold Shutdown
Mode. Primary containment integrity is not required in this mode,
facilitating containment access for inspections. The proposed changes
also allow some outage activities on other systems to continue during
the pressure testing. The licensee claims the proposed changes are
consistent with the Boiling Water Reactor Standard Technical
Specifications given in NUREG-1433, Revision 1.
Basis for proposed no significant hazards consideration
determination:
[[Page 19977]]
As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The probability of a leak in the reactor coolant pressure
boundary during reactor coolant system pressure testing is not
increased by considering the reactor to be in Cold Shutdown. Since
the pressure tests are performed nearly water solid, at low decay
heat values, and near Cold Shutdown conditions, the stored energy in
the reactor core will be low. Under these conditions, the potential
for failed fuel and a subsequent increase in coolant activity is
minimized. In addition, Special Operations LCO [Limiting Condition
for Operation] 3.12.A requires supporting LCOs for ECCS [emergency
core cooling system]-Cold Condition, Standby Gas Treatment,
Secondary Containment isolation and Standby Gas Treatment initiation
instrumentation, and Auxiliary Electrical Systems to be met to
ensure secondary containment integrity is maintained and capable of
handling any airborne radioactivity or steam leaks that could occur
during the performance of hydrostatic or leak testing. A listing of
secondary containment isolation valves required to maintain
Secondary Containment Integrity is included in plant controlled
procedures. The required pressure testing conditions provide
adequate assurance that the consequences of a steam leak will be
conservatively bounded by the consequences of the postulated main
steam line break outside of primary containment. In the event of a
large primary system leak, the reactor vessel would rapidly
depressurize, allowing the low pressure core cooling systems to
operate. The capability of these systems would be adequate to keep
the core flooded under this low decay heat load condition. Small
system leaks would be detected by leakage inspections before
significant inventory loss occurred. Therefore, the consequences of
an accident previously evaluated are not significantly increased.
2. Create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed changes do not introduce any new accident
initiators or failure mechanisms since the changes do not involve
any changes to structures, systems, or components, do not involve
any change to the operation of systems, and alter procedures only to
the extent that the 212 deg. F limit may be exceeded during reactor
coolant system pressure testing with certain systems inoperable.
There are no alterations to plant systems designed to mitigate the
consequences of accidents. The only difference is that a different
subset of plant systems would be utilized for accident mitigation
than those utilized during the Hot Shutdown Mode. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from those previously evaluated.
3. Involve a significant reduction in the margin of safety.
Since pressure tests are performed nearly water solid, at low
decay heat values, and near Cold Shutdown conditions, the stored
energy in the reactor core will be low. Under these conditions, the
potential for failed fuel and a subsequent increase in coolant
activity is minimized. Since secondary containment integrity will be
maintained, in accordance with the Special Operations LCO, the
secondary containment will be capable of handling any airborne
radioactivity or steam leaks that could occur during the performance
of hydrostatic or leak testing. Therefore, the proposed 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: February 26, 1998.
Description of amendment request: The proposed changes would change
the allowed containment leakage rate to 1.5 percent per day, changes
the assumed standby gas treatment system (SBGT) filter efficiency, and
revises reactor coolant sampling requirements for low Iodine-131
concentrations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The Authority [the licensee] has evaluated the proposed TS
[technical specification] Amendment and determined that it does not
represent a significant hazards consideration. Based on the criteria
for defining a significant hazards consideration established in 10 CFR
50.92, operation of the James A. FitzPatrick Nuclear Power Plant in
accordance with the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The proposed changes do not involve a change to the design or
operation of the plant. The systems affected by this proposed TS
change are not assumed in any safety analyses to initiate any
accident sequence. Therefore, the probability of any accident
previously evaluated is not increased by this proposed TS change.
The change in the allowable containment leakage rate (La)
is consistent with the accident analyses. The assumption of only 90%
SBGT filter efficiency is conservative with actual system
performance and is consistent with Regulatory Guide 1.52. There is
no significant change to the consequences of an accident previously
evaluated because maintaining containment leakage within limits
assumed in the accident analyses ensures that the dose consequences
resulting from an accident are not increased. The calculated doses
with the decreased SBGT system charcoal efficiency for design basis
accidents are marginally increased but still meet, and are well
below, the dose acceptance criteria of 10 CFR 100, the SRP [Standard
Review Plan, NUREG-0800], and GDC [General Design Criterion] 19 of
Appendix A to 10 CFR 50. The proposed TS changes maintain an
equivalent level of reliability and availability for all affected
systems. The ability of the affected systems associated with
maintaining leak rate integrity to perform their intended function
is unaffected by the proposed TS changes. Implementation of these
changes will provide continued assurance that specified parameters
associated with containment integrity will remain within acceptance
limits, and as such, will not significantly increase the
consequences of a previously evaluated accident. The change in the
value of .007 [microcurie]/ml to .002 [microcurie]/ml in section of
4.6.C. ``Coolant Chemistry'' is a minor editorial change, is more
conservative, and will correct the inconsistency between the
technical specification and its basis and as such, will not
significantly increase the consequences of a previously evaluated
accident.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The proposed amendment changes the allowed containment leakage
rate to 1.5%, changes the assumed value for SBGT system charcoal
filter efficiency, and changes a specification in section of 4.6.C.
``Coolant Chemistry'' from the value of .007 [microcurie]/ml to .002
[microcurie]/ml. No new accident modes are created by clarifying the
numerical value of the allowable containment leakage rate
(La) or changing the assumed value for the SBGT system
charcoal filter efficiency. No safety-related equipment or safety
functions are altered, or adversely affected, as a result of these
changes. The proposed changes will not introduce failure mechanisms
beyond those already considered in the current plant safety
analyses. Changing the allowable leakage rate, the assumed value for
the efficiency of the SBGT system charcoal filter, and the
specification in the bases section of 4.6.C. ``Coolant Chemistry''
does not contribute to the possibility of a new or different kind of
accident or malfunction from those previously analyzed.
[[Page 19978]]
(3) Involve a significant reduction in the margin of safety
because:
The proposed amendment changes the allowed containment leakage
rate to 1.5%, changes the assumed value for SBGT system charcoal
filter efficiency, and changes a specification in section of 4.6.C.
``Coolant Chemistry'' from the value of .007 [microcurie]/ml to .002
[microcurie]/ml. The design of the FitzPatrick plant is not changed.
The methodology for test performance is unchanged and Type A, B and
C tests will continue to be performed at [greater than or equal to]
Pa. The value of La specified in proposed
specification 6.20 is consistent with the accident analyses,
therefore, the dose consequences of any analyzed accidents are not
increased as a result of this change. The calculated doses as a
result of the decrease in the assumed efficiency of the SBGT system
charcoal filters for design basis accidents are marginally increased
but still meet, and are well below, the dose acceptance criteria of
10 CFR 100, the SRP, and GDC 19 of Appendix A to 10 CFR 50. The
change in the specification in section 4.6.C. ``Coolant Chemistry''
from .007 [microcurie]/ml to .002 [microcurie]/ml is a minor
editorial change, is more conservative, and will correct the
inconsistency between the technical specification and its basis.
Therefore, the proposed changes provide continued assurance of the
leak tightness of the containment and conservatively assume SBGT
system charcoal filter efficiency for the purpose of dose
calculations for design basis accidents without adversely affecting
the public health and safety and, as such, 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: March 30, 1998.
Description of amendment request: The proposed changes would change
the interval of selected Logic System Functional Tests (LSFT) from
semiannually to once per 24 months. The definition of LSFT is also
revised to be consistent with the Boiling Water Reactor Standard
Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The only significant change proposed by this application
involves the extension of the surveillance test interval for the
LSFTs required by the TS. The other changes involve editorial,
format, and clarification changes, which by their nature are of no
safety significance.
Extending the LSFT interval from semiannually to once per 24
months does not involve plant physical changes, change any TS
setpoints, or introduce any new mode of plant operation. Therefore,
the change does not degrade the performance of any safety system
assumed to function in the accident analysis, and therefore, will
not increase the consequences of an accident.
Extending the LSFT interval from semiannually to 24 months
results in no significant change in the logic system unavailability
due to equipment failure. The reliability of safety systems subject
to the LSFT are dominated by that of the mechanical components, and
the logic system circuit relay coils which are subject to the more
frequent functional test requirements. These factors are confirmed
by the availability record of the affected safety system based on
the past surveillance test history. Furthermore, the longer test
intervals reduce the unavailability due to testing for the
applicable safety system while the plant is operating. For these
reasons, there is not a significant increase in the probability of
an accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not introduce any new accident
initiators or failure mechanisms since the changes do not introduce
any new modes of plant operation, make any physical changes, or
change any TS setpoints. The changes reduce the probability of
accidents initiated by test-induced plant transients by reducing the
number of times the tests must be performed.
3. Involve a significant reduction in a margin of safety.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. In several aspects, the proposed changes
may actually enhance the margin of safety by reducing the potential
for test-induced plant transients, reducing the unavailability due
to test of the applicable safety system, and reducing any potential
incremental logic system component wear. For these reasons, the
changes do not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.
Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear
Generating Station, Unit No. 1, Salem County, New Jersey
Date of amendment request: March 26, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.1.3.3, ``Rod Drop Time,'' to change
the applicability from Mode 3 (hot shutdown) to Modes 1 and 2 (startup
and power operation).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change to the Technical Specification Mode
applicability provides consistency between the testing requirements
as stated in the surveillance requirement of the Technical
Specifications and intended by the initial conditions specified in
the limiting condition for operations. The proposed change does not
introduce any physical changes to the plant or equipment already in
place in the plant, the proposed change ensures that testing of the
rod drop times is performed in a manner that is consistent with the
Technical Specifications and the assumptions made in the Salem
accident analysis.
Therefore, the proposed amendment does not increase the
probability or consequences of any 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 does not introduce a new component or
changes the manner in which the facility is operated, maintained or
tested. Thus no new accident scenarios, failure mechanisms or
limiting single failures are introduced as a result of the proposed
change to the facility.
Therefore the proposed amendment does not create the possibility
of a new or different kind of accident from any previously
evaluated.
[[Page 19979]]
3. The proposed changes does not involve a significant reduction
in a margin of safety.
As stated in question number 2, the proposed change does not
introduce a new component or changes the manner in which the
facility is operated. Operation of the facility in accordance with
the proposed amendment would not involve a significant reduction in
the margin of safety. The Technical Specifications remain the same,
as the input, or initial conditions, of the safety analysis have not
changed. Therefore, there is no reduction in the margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: Robert A. Capra.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2, and 3, Limestone County, Alabama
Date of amendment request: March 3, 1998.
Description of amendment request: The proposed amendment would
change the Browns Ferry Nuclear Plant Unit 2 and Unit 3 Technical
Specification Figure 3.6-1 which contains the reactor vessel pressure-
temperature (PT) limits. The change would extend the validity of the
curves to 32 effective full-power years (EFPY). The current PT curves
are effective up to 12 EFPY. In addition to revised PT curves, several
changes to the notes applicable to the curves are also proposed to be
consistent with the supporting analysis.
The proposed PT curves also would support a planned 5% power
increase for each unit. Approval of the proposed power increase is
pending and is the subject of a separate action before the Commission.
The Tennessee Valley Authority has submitted the proposed change in
current technical specification (CTS) format and in the improved
standard technical specification (ISTS) format. Conversion to the ISTS
format is pending.
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.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10
CFR 50.92(c)(1)) because the proposed changes merely adjust the
reference temperature for the limiting reactor vessel beltline
material to account for accumulated and projected irradiation
effects. The adjusted reference temperature analyses were performed
in accordance with the requirements of Appendix G of 10 CFR part 50
and the guidance contained in Regulatory Guide 1.99, Revision 2. The
changes do not otherwise affect the manner by which the facility is
operated and do not change any facility design feature or equipment.
Since the protection previously provided will continue to be
provided and there is no change to the facility or operating
procedures, there is no effect upon the probability or consequences
of any accident previously analyzed.
B. The changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
(10 CFR 50.92(c)(2)) because no new failure modes are introduced.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
(10 CFR 50.92(c)(2)) because they do not affect the function of any
facility structure, system or component, or affect the manner by
which the facility is operated.
C. The changes do not involve a significant reduction in a
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes
assure that the reactor vessel PT limits will be valid for operation
up to 32 EFPY and that the safety margins specified in Appendix G of
10 CFR part 50 will be maintained.
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: Athens Public Library, 405 E.
South Street, Athens, Alabama 35611.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
ET 10H 400 West Summit Hill Drive, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: February 13, 1998 (TS 97-03).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Technical Specification (TS) by adding a new
limiting condition for operation (LCO) that addresses requirements for
the main feedwater isolation, regulating, and bypass valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
TVA has concluded that operation of SQN Units 1 and 2 in
accordance with the proposed change to the TSs [or operating
license(s)], does not involve a significant hazards consideration.
TVA's conclusion is based on its evaluation, in accordance with 10
CFR 50.91(a)(1), of the three standards set forth in 10 CFR
50.92(c).
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
TVA will not change plant components, functions, or operating
practices by implementing a change that adds a TS requirement for
the main feedwater isolation, regulating, and bypass valves. TVA
will maintain and verify operability of these valves through the
proposed surveillance and actions to ensure the accident mitigation
functions are available when applicable. These valves are not
considered to be the source of an accident and the conservative
addition of a requirement to maintain their safety function will not
increase the probability of an accident. TVA will not increase the
consequences of an accident by implementing this change because this
addition ensures that the isolation of main feedwater is available
to mitigate the consequences of an accident.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
TVA will not alter plant equipment or operating activities in
the implementation of the proposed TS change. The valves used for
the isolation of main feedwater are not a potential source for
accidents and are designed for accident mitigation purposes.
Therefore, TVA 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.
TVA maintains and ensures the availability of the isolation
function for the main feedwater system as assumed in the SQN
accident analysis. TVA proposes this TS change to further assure
this capability and to meet the requirements of 10 CFR 50.36. TVA
will not change the methods of operating the plant or setpoints
associated with safety-related equipment in the implementation of
this request. Therefore, TVA will not reduce the margin of safety by
implementing a TS LCO for the isolation functions of the main
feedwater system.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 19980]]
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 Project Director: Frederick J. Hebdon.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: February 13, 1998 (TS 97-07).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Technical Specification (TS) requirements for
main steam isolation valves (MSIVs) to incorporate MSIV requirements
consistent with the Westinghouse Standard TS (NUREG-1431) and would add
testing requirements for the MSIVs that ensure the valves close on an
actual or simulated automatic actuation signal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
TVA has concluded that operation of SQN Units 1 and 2, in
accordance with the proposed change to the TSs, does not involve a
significant hazards consideration. TVA's conclusion is based on its
evaluation, in accordance with 10 CFR 50.91(a)(1), of the three
standards set forth in 10 CFR 50.92(c).
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes provide enhancements and clarifications of
the requirements for inoperable MSIVs and periodic testing
provisions. These changes do not alter the safety functions of the
MSIVs or the operating practices that govern their application to
plant conditions. The actions for Modes 2 and 3 are revised such
that a longer time could occur before an inoperable MSIV is closed
or the unit is placed in a mode that does not apply. However, this
increase will not significantly impact the ability of the valves to
mitigate an accident or affect the accident generation possibility.
This is based on the low probability of an accident occurring that
would require closure of the MSIVs and reasonable time intervals to
transition to lower modes based on operating experience to reach the
required modes in an orderly manner without challenging unit
systems.
The MSIVs provide accident mitigation functions but do not
contribute to accident generation. The MSIV functions have not been
altered by the proposed changes. Therefore, the proposed changes
will not increase the probability of a previously evaluated
accident. Based on the above discussions, the proposed changes will
not significantly increase the consequences of an accident and in
some instances they will enhance the safety functions.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The primary function of the MSIVs is to support accident
mitigation and are not a significant contributor to events that
could generate accidents. The main impact that could result from an
inoperable MSIV is an inadvertent closure that results in a unit
trip. This event is bounded by the accidents that are currently
evaluated for SQN. Since the proposed change does not alter MSIV
functions and the new surveillance will be performed in modes that
will not challenge unit systems, the possibility of a new or
different kind of accident is not created.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed changes clarify and enhance the current SQN
requirements for the MSIVs with one exception. This exception is the
completion time added to the Modes 2 and 3 action that could be a
negative impact to the margin of safety. This change could allow the
MSIV safety function to be inoperable for a longer period of time.
The overall effect of the proposed changes considering the
additional end-device testing, periodic verification of inoperable
MSIV closure, and removal of the action to allow MSIV closure in
Mode 1, is considered a positive impact to the margin of safety.
Therefore, there is not a significant reduction in the margin of
safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 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 Project Director: Frederick J. Hebdon.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: March 25, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) Sections 6.1.1; 6.2.1.b; 6.5.1.1;
6.5.1.6. a, d, h, and m; 6.5.1.7.c; 6.5.1.8; 6.14.1.2; 6.15.b; 6.2.3.5;
6.5.1.2; 6.5.1.7.a for Unit 1 and 6.1.1; 6.2.1.b; 6.5.1.1; 6.5.1.6. a,
d, h, and m; 6.5.1.7.c; 6.5.1.8; 6.13.b; 6.14.b; 6.2.3.5; 6.5.1.2; and
6.5.1.7.a for Unit 2, changing the title of Station Manager to Site
Vice President, and the titles of the Assistant Station Managers to
Manager-Station Operation and Maintenance and Manager-Station Safety
and Licensing.
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:
Virginia Electric and Power Company has reviewed the proposed
Technical Specifications changes against the criteria of 10 CFR
50.92 and has concluded that the changes do not pose a significant
hazards consideration. Specifically, station operations in
accordance with the proposed Technical Specifications changes will
not:
a. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes are administrative in nature. The overall
responsibility for safe operation and review of plant operations is
not being changed. There are no changes to the operation of any
plant system or its design as a result of these changes. Therefore,
neither the probability of occurrence nor the consequences of an
accident or malfunction of equipment important to safety previously
evaluated in the safety analysis report are increased.
b. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes are administrative in nature. The overall
responsibility for safe operation and review of plant operations is
not being changed. There are no changes to the operation of any
plant system or its design that could create any new modes of
operation or accident precursors. Therefore, it is concluded that no
new or different kind of accident or malfunction from any previously
evaluated has been created.
c. The proposed changes do not result in a significant reduction
in margin of safety as defined in the basis for any Technical
Specifications.
The proposed changes are administrative in nature. The overall
responsibility for safe operation and review is not being changed.
There are no changes to the operation of any plant system or its
design as a result of these changes. Safety systems are maintained
operable as required by Technical Specifications. Therefore, the
margin of safety is not changed.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
[[Page 19981]]
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: P. T. Kuo, Acting Project Director.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 5, 1997.
Description of amendment request: The proposed Operating License
change and changes to the technical specifications (TS) would permit
the use of a temporary alternate supply line (jumper) to provide
service water (SW) to the component cooling heat exchangers. The
temporary jumper will permit maintenance to be performed on the
existing supply line.
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:
Virginia Electric and Power Company has reviewed the proposed
changes against the criteria of 10 CFR 50.92 and has concluded that
the changes do not pose a significant safety hazards consideration
as defined therein. The proposed Operating License and Technical
Specifications and Bases changes are necessary to allow the use of a
temporary, seismic, non-missile protected jumper to provide service
water (SW) to the Component Cooling Heat Exchangers (CCHXs) while
maintenance work is performed on the existing SW supply line to the
CCHXs. Since there is only one SW supply line to the CCHXs, an
alternate SW supply must be provided whenever the line is removed
from service. The temporary jumper provides this function.
The use of the temporary jumper has been thoroughly evaluated,
and appropriate constraints and compensatory measures (including a
Contingency Action Plan) have been developed to ensure that the
temporary jumper is reliable, safe, and suitable for its intended
purpose. A complete and immediate loss of SW supply to the operating
CCHXs is not considered credible, given the project constraints and
the unlikely probability of a generated missile. Existing station
abnormal procedures already address a loss of component cooling, and
the use of alternate cooling for a loss of decay heat removal, in
the unlikely event that they are required. Furthermore, appropriate
mitigative measures have been identified to address potential
flooding concerns. The minor administrative changes merely correct a
table format inconsistency and update Basis section references.
Consequently, the operation of Surry Power Station with the
proposed amendment and license condition will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The SW and CC Systems will function as designed under the Unit
operating constraints specified by this project (i.e., Unit 2 in
operation and Unit 1 in a refueling outage), and the potential for a
loss of component cooling is already addressed by Station Abnormal
Procedures. Therefore, there is no increase in the probability of an
accident previously evaluated. The possibility of flooding due to
failure of the temporary SW supply jumper in the Turbine Building
basement has been evaluated and dispositioned by the implementation
of appropriate precautions and compensatory measures to preclude
damage to the temporary jumper and to respond to a postulated
flooding event. A flood watch will be present around-the-clock with
authority and procedural guidance to isolate the jumper, if
required. Furthermore, the CCHXs serve no design basis accident
mitigating function. Therefore, the consequences of an accident
previously evaluated are not increased.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The SW and CC Systems' design functions and basic configurations
are not being altered as a result of using a temporary SW supply
jumper. The temporary jumper is designed to be safety-related and
seismic with all of the design attributes of the normal SW supply line,
except for the automatic isolation function and complete missile
protection. The design functions of the SW and CC systems are unchanged
as a result of the proposed changes due to (1) required plant
conditions, (2) compensatory measures, (3) a Contingency Action Plan
for restoration of the normal SW supply if required, and (4) strict
administrative control of the temporary SW valve to preclude flooding
or to isolate non-essential SW within the design basis assumed time
limits. Unit 1 will be in a plant condition which will provide adequate
time to restore the normal SW supply, if required. Therefore, since the
SW and CC systems will basically function as designed and will be
operated in their basic configuration, the possibility of a new or
different type of accident than previously evaluated in the UFSAR
[Updated Final Safety Analysis Report] is not created.
3. Involve a significant reduction in a margin of safety.
The margin of safety as defined in the Technical Specifications is
not reduced since an operable SW flowpath to the required number of
CCHXs is provided, and Unit operating constraints, compensatory
measures and contingencies will be implemented as required to ensure
the integrity and the capability of the SW flowpath. The use of the
temporary jumper will be limited to the time period when missile
producing weather is not expected, and Unit 1 meets specified unit
conditions. Therefore, the temporary SW jumper, under the imposed
project constraints and compensatory measures, provides the same
reliability as the normal SW supply line. Furthermore, the
Probabilistic Safety Assessment for Surry Power Station has been
reviewed relative to potential flooding when the temporary SW jumper is
in use. It has been determined that due to the SW restoration project's
compensatory and contingency measures, as well as the constraints
imposed by the Maintenance Rule online risk matrix, the impact on core
damage frequency due to flooding is negligible.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: P. T. Kuo, Acting.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed no Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
[[Page 19982]]
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: March 6, 1998.
Description of amendment request: The proposed amendment would (1)
update the Technical Specification heatup and cooldown rate curves and
extend their reactor fluence limit from the current 20 effective full
power years (EFPY) to a new value of 35 EFPY, (2) incorporate into
Technical Specifications the use of a Pressure and Temperature Limits
Report (PTLR), and (3) change the power-operated relief valves (PORVs)
temperature requirement for operability.
Date of individual notice in the Federal Register: March 27, 1998
(63 FR 14972).
Expiration date of individual notice: April 27, 1998.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power
Station, Unit 2, Grundy County, Illinois
Date of application for amendment: March 19, 1998, as supplemented
by letters dated March 28, 1998, and April 3, 1998.
Brief description of amendment: The amendment changes the Technical
Specifications (TS) by revising the Dresden, Unit 2, Minimum Critical
Power Ratio (MCPR) in TS Section 2.1.B and footnotes in TS Section
5.3.A, to allow the use of Siemens Power Corporation ATRIUM-9B fuel for
all operating Modes at Dresden, Unit 2, Cycle 16.
Date of issuance: April 10, 1998.
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 168.
Facility Operating License No. DPR-19: The amendment revised the
TS. Public comments requested as to proposed no significant hazards
consideration: Yes (63 FR 14735 dated March 26, 1998). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by April 27, 1998, but indicated that if the Commission makes a
final no significant hazards consideration determination any such
hearing would take place after issuance of the amendment.
The March 28, 1998, and April 3, 1998, letters provided additional
clarifying information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated April 10, 1998.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant Middlesex County, Connecticut
Date of application for amendment: Two applications, both dated May
30, 1997.
Brief description of amendment: Changes Administrative Controls
Section of the Technical Specifications to implement new Certified Fuel
Handler position and to implement revised management responsibilities
and titles that reflect the permanently shut down status of the plant.
In addition, minor typographical errors were corrected.
Date of issuance: March 27, 1998.
Effective date: Date of issuance, but to be implemented within 60
days of issuance.
Amendment No.: 192.
Operating License No. DPR-61: Amendment revised the Technical
Specifications.
Date of initial notice in Federal Register: July 16, 1997 (62 FR
38132 and 62 FR 38133). The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated March 27, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of application for amendment: March 26, 1997.
Brief description of amendment: The amendment revises the
Containment Systems technical specifications (TS) to incorporate a note
to allow opening an operable airlock door to perform repairs on
inoperable airlock components when the other airlock door is
inoperable. This amendment is in partial response to Consumers Energy's
March 26, 1997, application. The Consumers Energy request also proposed
revising the requirements contained in TS sections 3.6 and 4.5 to
closely emulate the format and content of NUREG-1432, ``Standard
Technical Specifications, Combustion Engineering Plants,'' (STS). That
portion of the Consumers Energy request remains under staff review and
will be addressed in a separate evaluation.
Date of issuance: April 8, 1998.
Effective date: April 8, 1998.
Amendment No.: 179.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
[[Page 19983]]
Date of initial notice in Federal Register: December 17, 1997 (62
FR 66136).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: January 28, 1998 (NRC-98-0008),
as supplemented on March 10, 1998 (NRC-98-0036).
Brief description of amendment: The amendment revises the technical
specifications (TSs) by modifying the ``#'' footnote to Table 1.2 and
the ``*'' footnote to surveillance requirements 4.9.1.2 and 4.9.1.3 to
permit the Reactor Mode Switch to be placed in the Run or Startup/Hot
Standby positions to test switch interlock functions provided that all
control rods are verified to remain fully inserted in core cells
containing one or more fuel assemblies.
Date of issuance: March 31, 1998.
Effective date: March 31, 1998, with full implementation within 90
days.
Amendment No.: 116.
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9599) The March 10, 1998, supplement requested a change in the
implementation period and was not outside the scope of the initial
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated March 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Duke Energy Corporation (DEC), Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: December 17, 1997.
Brief description of amendments: The amendments revise Technical
Specification Section 6.9.1.9 to reference updated or recently approved
topical reports, which contain methodologies used to calculate cycle-
specific limits contained in the Core Operating Limits Report. For
several reports DEC indicated staff approval, but neglected to provide
an ``A'' designation for the report number. Upon agreement by DEC, the
staff has made these appropriate editorial corrections. These topical
reports have all been previously approved by the staff under licensing
actions separate from the current amendment request.
Date of issuance: April 8, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1--178; Unit 2--160.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4311).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: February 9, 1998.
Brief description of amendment: The amendment approves the use of
the repair roll technology (reroll) for the upper tubesheet region of
the ANO-1 steam generators. The reroll technology is an alternative to
the either sleeving or plugging steam generator tubes found during
inservice inspections to have defects that exceed the stated repair
criteria. The reroll methodology works by creating a new mechanical
tube to tubesheet structural joint below the tube defect indication.
Date of issuance: April 10, 1998.
Effective date: April 10, 1998.
Amendment No.: 190.
Facility Operating License No. DPR-51: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1998 (63
FR 9268).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 10, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: April 1, 1998, as supplemented by letter
dated April 8, 1998.
Brief description of amendment: The amendment allows approximately
440 steam generator tubes with confirmed volumetric indications within
the upper tube sheet to remain in service during Cycle 15. The
amendment revises TS 4.18.5.b to incorporate five criteria which need
to be satisfied to allow steam generator tubes to remain in service
during Cycle 15 with indications of outer diameter intergranular attack
(ODIGA) in the upper tube sheet region of the steam generators.
Date of issuance: April 10, 1998.
Effective date: April 10, 1998.
Amendment No.: 191.
Facility Operating License No. DPR-51: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated April
10, 1998.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 18, 1996, as supplemented by
letter dated January 21, 1998.
Brief description of amendment: The amendment changes Technical
Specification (TS) Surveillance Requirement 4.4.8.3.1.b to test the
Shutdown Cooling System suction line relief valves in accordance with
TS 4.0.5. Editorial changes to 4.4.8.3.1 and 4.4.8.3.1.a have also been
made.
Date of issuance: April 1, 1998.
Effective date: April 1, 1998.
Amendment No.: 140.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 11, 1998 (63
FR 6985).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
[[Page 19984]]
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 24, 1997, as supplemented by letter
dated January 21, 1998.
Brief description of amendment: The amendment modifies Technical
Specifications (TS) 3.1.1.1, 3.1.1.2, 3.10.1 and Figure 3.1-1 by
removing cycle dependent boron concentration and boration flow rate
from the Action Statements and removing the ``RWSP at 1720 ppm'' curve
from the figure. A change to TS Bases 3/4.1.1.1 and 3/4.1.1.2 has been
included to support this change.
Date of issuance: April 8, 1998.
Effective date: April 8, 1998, to be implemented within 60 days.
Amendment No.: 141.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33123).
The January 21, 1998, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Entergy Operations, Inc., 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: October 28, 1997, as
supplemented by letter dated January 9, 1998.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to implement the containment leak rate testing
provisions of 10 CFR Part 50, Appendix J, Option B.
Date of issuance: April 6, 1998.
Effective date: April 6, 1998.
Amendment No: 135.
Facility Operating License No. NPF-29: Amendment revises the TSs.
Date of initial notice in Federal Register: December 3, 1997 ( 62
FR 63976).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 6, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
IES Utilities Inc, Central Iowa Power Cooperative, and Corn Belt Power
Cooperative, Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of application for amendment: February 3, 1998.
Brief description of amendment: The amendment changes the
operability requirement for the Standby Liquid Control system to Run/
Power Operations and Startup.
Date of issuance: March 31, 1998.
Effective date: March 31, 1998.
Amendment No.: 221.
Facility Operating License No. DPR-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 26, 1998 (63
FR 9874).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, Iowa 52401.
IES Utilities Inc, Central Iowa Power Cooperative, and Corn Belt Power
Cooperative, Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of application for amendment: October 3, 1997 as supplemented
on December 10, 1997.
Brief description of amendment: The amendment revises the Operating
License to allow the start of core offload as soon as 60 hours after
shutdown.
Date of issuance: April 2, 1998.
Effective date: April 2, 1998.
Amendment No.: 222.
Facility Operating License No. DPR-49: Amendment revised the
Operating License.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4314).
The December 10, 1997 submittal provided clarifying information
that did not change the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, Iowa 52401.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: October 20, 1997, as
supplemented February 10, and March 10, 1998.
Brief description of amendment: The amendment replaces in their
entirety the existing Technical Specifications incorporated in Facility
Operating License No. DPR-36 as Appendix A. Maine Yankee developed the
revised Technical Specifications, titled Permanently Defueled Technical
Specifications, to reflect the permanently shutdown and defueled status
of the plant. Changes were made to the definitions, limiting conditions
for operation, surveillance, and administrative control sections.
Date of issuance: March 30, 1998.
Effective date: March 30, 1998.
Amendment No.: 161.
Facility Operating License No. DPR-36: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63978). The February 10, and March 10, 1998, submittals added
additional programs to the Section 5.5 Procedures and Section 5.6
Programs and Manuals did not change the proposed no significant hazards
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 30, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: October 31, 1997, as
supplemented by letter dated February 3, 1998.
Brief description of amendment: This amendment changes Technical
Specifications to support design changes to upgrade the analog-based
average power range monitor system with General Electric's Nuclear
Measurement Analysis and Control Power Range Neutron Monitor System,
including an Oscillation Power Range Monitor function.
Date of issuance: March 31, 1998.
Effective date: As of the date of issuance to be implemented upon
[[Page 19985]]
completion and acceptance of design modifications resulting from the
installation of the Nuclear Measurement Analysis and Control Power
Range Neutron Monitor System.
Amendment No.: 80.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68310).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: September 26, 1997, as supplemented by
letter dated March 12, 1998.
Description of amendment request: The amendment revises Technical
Specification 3.7.6, ``Control Room Emergency Makeup Air and
Filtration,'' and its associated Bases to separate the requirements for
the control room air conditioning subsystem from the requirements for
control room makeup air and filtration subsystem based on system
function. The amendment also increases the allowed outage time for the
Control Room Air Conditioning Subsystem.
Date of issuance: April 9, 1998.
Effective date: As of the date of issuance, with full
implementation within 60 days.
Amendment No.: 56.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54874). The March 12, 1998, supplemental letter did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: December 1, 1997.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) by adding a 2.0 second plus or minus 0.1 second
time delay to the 4.16 kV Emergency Bus Undervoltage Loss of Power,
Level One, trip setpoint and allowable values in TS Table 3.3-4.
Date of issuance: April 1, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 214.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 14,1998 (63 FR
2280).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room locations: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: November 11, 1997.
Brief description of amendment: The amendment allows NNECO to
credit soluble boron for maintaining k-effective at less than or equal
to 0.95 within the spent fuel pool rack matrix following a seismic
event of a magnitude greater than or equal to an operating basis
earthquake.
Date of issuance: April 9, 1998.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 158.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63980).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room locations: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of application for amendment: December 9, 1996, as
supplemented on June 12, 1997, and March 13, 1998.
Brief description of amendment: This amendment revised the
Technical Specification to incorporate the requirements of appendix I
of 10 CFR Part 50, into the Radiological Effluent Technical
Specification (RETS) and to relocate the controls and limitations on
RETS and radiological environmental monitoring (Currently in the
Technical Specifications) to the Offsite Dose Calculation Manual and
the Process Control Program. The amendment also revised the Technical
Specifications to implement Generic Letter 89-10 (GL 89-10) and to
incorporate the requirements of the revised 10 CFR Part 20.
Date of issuance: April 8, 1998.
Effective date: As of the date of issuance and shall be implemented
no later than 30 days from the date of issuance.
Amendment No.: 32.
Facility Operating License No. DPR-7: Amendment revised the TS.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18174).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Humboldt County Library, 131
3rd Street, Eureka, California 95501.
Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania
Date of application for amendment: August 26, 1997, as supplemented
by letters dated December 4, 1997, February 2, March 19, and April 2,
1998.
Brief description of amendment: This amendment changes the
Susquehanna Unit 1 Technical Specifications to support the use of the
Siemens Power Corporation ATRIUM-10 fuel design in the upcoming Cycle
11 refueling outage.
Date of issuance: April 6, 1998.
Effective date: As of date of issuance, to be implemented within 30
days.
[[Page 19986]]
Amendment No.: 174.
Facility Operating License No. NPF-14: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68314).
The December 4, 1997, February 2, March 19, and April 2, 1998,
submittals provided clarifying information that did not change the
initial no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 6, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Philadelphia Electric Company, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of application for amendment: September 2, 1997.
Brief description of amendment: These amendments revise LGS, Units
1 and 2, TS Section 4.0.5, and Bases Sections B 4.0.5 and B 3/4.4.8
regarding the surveillance requirements associated with Inservice
Testing and Inservice Inspection Programs of the American Society of
Mechanical Engineers Code Class 1, 2, and 3 components.
Date of issuance: March 31, 1998.
Effective date: March 31, 1998.
Amendment Nos.: 125 and 89.
Facility Operating License No. NPF-39: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 11, 1998 (63
FR 6990).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: July 23, 1997, as supplemented by
letters dated September 30, October 27, and December 18, 1997, and
February 12, 1998.
Brief Description of amendments: The amendments revise the
Pressure-Temperature Limit Heatup, Cooldown, and Hydrostatic Testing
curves for Farley Units 1 and 2 and relocate the curves from the
Technical Specifications to a Pressure and Temperature Limits Report
for each unit.
Date of issuance: April 9, 1998.
Effective date: As of the date of issuance to be implemented for
Unit 1 prior to entering Mode 4 for Cycle 16 refueling outage (fall
1998); for Unit 2 prior to entering Mode 4 for Cycle 13 refueling
outage (spring 1998).
Amendment Nos.: Unit 1-136; Unit 2-128.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: September 10, 1997 (62
FR 47699); January 14, 1998 (63 FR 2281); February 23, 1998 (63 FR
9020).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: October 24, 1997 (TXX-97228).
Brief description of amendments: The amendments revise core safety
limit curves and Overtemperature N-16 reactor trip setpoints based on
analyses of the core configuration and expected operation for CPSES
Unit 1, Cycle 7. The changes apply equally to CPSES Units 1 and 2
licenses since the Technical Specifications are combined.
Date of issuance: March 27, 1998.
Effective date: March 27, 1998, to be implemented within 30 days.
Amendment Nos.: Unit 1--Amendment No. 57; Unit 2--Amendment No. 43.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications/operating licenses.
Date of initial notice in Federal Register: November 19, 1997 (62
FR 61847).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 27, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: August 8, 1997, as supplemented
by letter dated January 16, 1998.
Brief description of amendment: The amendment revises the Callaway
Plant, Unit 1 Technical Specification Table 3.3-3 Functional Units
4.b.2 and 5.a.2 to make the number of main steam and feedwater
isolation system (MSFIS) channels consistent with the solid state
protection system, adds a clarifying note and changes Table 4.3-2
Functional Units 4.b.2 and 5.a.2 slave relay quarterly test to a
monthly staggered actuation logic test.
Date of issuance: March 25, 1998.
Effective date: March 25, 1998, to be implemented within 30 days
from the date of issuance.
Amendment No.: 123.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 17, 1997 (62
FR 66143) The January 16, 1998, supplemental letter provided additional
clarifying information that did not change the staff's original no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 25, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: October 17, 1997, as
supplemented by letters dated March 3, 1998, and March 17, 1998.
Brief description of amendment: The amendment revises the technical
specifications to modify the heatup and cooldown curves and the maximum
allowable power operated relief valve setpoint curves for cold
overpressure protection.
Date of issuance: April 2, 1998.
Effective date: April 2, 1998, to be implemented within 30 days
from the date of issuance.
Amendment No.: 124.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 14, 1998 (63 FR
2282).
The March 3, 1998, and March 17, 1998, supplemental letters
provided
[[Page 19987]]
additional clarifying information and did not change the initial no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated April 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: October 11, 1996.
Brief description of amendment: The amendment revises the Technical
Specifications regarding the amount of foam concentrate required to
support operability of the reactor recirculation motor generator set
foam fire suppression system.
Date of Issuance: March 31, 1998.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 156.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54877).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated March 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: June 9, 1997.
Brief description of amendment: The amendment revises Technical
Specification Section 6.0 to add and revise reference to NRC-approved
methodologies which will be used to validate or generate the cycle-
specific thermal hydraulic stability based operating limits in the
Vermont Yankee Core Operating Limits Report.
Date of Issuance: April 7, 1998.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 157.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 13, 1997 (62 FR
43377).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated April 7, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: August 22, 1996.
Brief description of amendment: The amendment revises the Technical
Specifications to change the action statement for the high range stack
noble gas monitor based on the guidance of Generic Letter 83-36, NUREG-
0737 Technical Specifications.
Date of Issuance: April 8, 1998.
Effective date: April 8, 1998, to be implemented within 30 days.
Amendment No.: 158.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30647).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated April 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: January 28, 1998.
Brief description of amendment: The amendment revises Technical
Specification Secs. 6.3 and 6.12 to reflect a merger for the positions
of Superintendent Radiation Protection and Superintendent Chemistry
into one new position, Manager Chemistry/Radiation Protection.
Date of issuance: March 30, 1998.
Effective date: March 30, 1998.
Amendment No.: 115.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9614).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 30, 1998.
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
[[Page 19988]]
opportunity to provide for public comment on its no significant hazards
consideration determination. In such case, the license amendment has
been issued without opportunity for comment. If there has been some
time for public comment but less than 30 days, the Commission may
provide an opportunity for public comment. If comments have been
requested, it is so stated. In either event, the State has been
consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By May 22, 1998, the licensee
may file a request for a hearing with respect to issuance of the
amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public 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).
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: April 2, 1998 (NRC-98-0062).
[[Page 19989]]
Description of amendment request: The amendment revised the action
specified in Technical Specification Table 3.3.7.5-1 if one channel of
drywell oxygen monitoring is inoperable.
Date of issuance: April 3, 1998.
Effective date: April 3, 1998, with full implementation by April 6,
1998.
Amendment No.: 117.
Facility Operating License No. NPF-43: Amendment revises the
License and the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, and final determination
of no significant hazards consideration are contained in a Safety
Evaluation dated April 3, 1998.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter.
Dated at Rockville, Maryland, this 15th day of April 1998.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-10470 Filed 4-21-98; 8:45 am]
BILLING CODE 7590-01-P