[Federal Register Volume 61, Number 2 (Wednesday, January 3, 1996)]
[Notices]
[Pages 174-192]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10103]
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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 December 11, 1995, through December 20,
1995. The last biweekly notice was published on December 20, 1995 (60
FR 65672).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By February 2, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
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.
[[Page 175]]
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: November 1, 1995, as supplemented on
December 1, 1995
Description of amendments request: The proposed amendments would
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 2 and 3,
Technical Specifications (TSs) and supporting TS Bases relating to the
electrical distribution system. The changes are necessary to
accommodate the installation of a new safety-related emergency diesel
generator (EDG) and a non-safety EDG. The non-safety EDG will be used
as an alternate air conditioning source of power in case of a station
blackout. In addition to reflecting the new plant configuration, the
proposed TSs also reflect the upgraded electrical capacities of the
existing EDGs, increased fuel oil storage, and fire protection system
for the new EDG building.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The Engineered Safety Features (ESF) electrical system provides
a reliable source of electrical power to the 4.16 kV ESF busses to
operate the necessary accident mitigation equipment, should offsite
power be lost. The proposed change to Units 1 and 2 Technical
Specifications was prompted by two significant modifications to this
system - the addition of No. 1A Emergency Diesel Generator (EDG) and
the upgrade of the electrical capacity of two of the three existing
Fairbanks Morse EDGs. The addition of No. 1A EDG provides the plant
with an ESF electrical system configuration consisting of two EDGs
dedicated to each unit, thereby eliminating reliance upon a
``swing'' diesel capable of being aligned to either unit. The four-
EDG configuration provides a greater degree of flexibility when an
EDG is being overhauled or tested during refueling outages. The
increased electrical capacity of the existing Fairbanks Morse EDGs
will give the operators greater flexibility in the choice of
discretionary loads for the mitigation of accidents. Both
modifications necessitate changes to the Technical Specifications.
The ESF electrical system, including the four EDGs, is used to
mitigate the consequences of an accident. The design of the new No.
1A EDG is such that incorporation of this EDG into the existing ESF
electrical system does not result in this system becoming an
accident initiator. Furthermore, the modification to upgrade the
capacity of the existing EDGs will enhance the plant operators'
ability to mitigate accidents by allowing greater flexibility in the
choice of discretionary loads, but will not change the configuration
of the ESF electrical system or any support systems such that the
EDGs would become an accident initiator. Therefore, the proposed
change would not increase the probability of an accident previously
evaluated.
The addition of the safety-related No. 1A EDG to the ESF
electrical system will enhance the ability to provide reliable
electric power during all modes of operation and shutdown conditions
of the plant. Number 1A EDG and its support systems are designed
such that failure of a single component will not prevent the
capability to safely shut down the plant and to maintain the plant
in a safe shutdown condition. Furthermore, non-safety-related
systems associated with No. 1A EDG are designed so that their
failure will not result in the loss of function of any safety-
related system. The design of the Fire Protection System in the
Diesel Generator Building meets the Codes and Standards specified in
the mechanical and instrumentation and controls design reports,
previously approved by the
[[Page 176]]
Commission. Inclusion of components from these systems into the
Technical Specifications is consistent with Calvert Cliff's current
licensing basis. The proposed Technical Specifications will
demonstrate the reliability and capability of No. 1A EDG and the
upgraded Fairbanks Morse EDGs to perform their accident mitigation
function. Implementation of the proposed Technical Specifications
will not reduce the ability of the EDGs to perform their safety
functions. The increased volume of fuel oil necessary to support
operation of No. 1A EDG and the upgraded Fairbanks Morse EDGs will
not adversely impact the ability of any systems to perform their
safety functions. The auxiliary systems which required modification
or analysis to support the upgraded ratings of the Fairbanks Morse
EDGs will not adversely impact operation of any other plant systems
necessary to mitigate the consequences of an accident. Based on the
above, the proposed change would not increase the consequences of an
accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed change adds Surveillance Requirements, Limiting
Conditions for Operation, and Action Statements to reflect the
addition of a new EDG to the ESF electrical system, and upgrades the
electrical capacity of the existing Fairbanks Morse EDGs. This
change does not add any new equipment, modify any interfaces with
any existing equipment, or change the equipment's function, or the
method of operating the equipment to be modified. The system will
continue to operate in the same manner as before the capacity
upgrades were implemented. The additional fuel oil required to
support the capacity upgrades will be stored in the existing Seismic
Category I fuel oil storage tanks. The modified EDGs will continue
to serve a function as accident mitigators, and will not become an
initiator of any accident.
The NRC has reviewed the design of the new EDG, its attendant
support systems and the new EDG Building, and concurs with Baltimore
Gas and Electric Company's determination that the design satisfies
the design requirements for a safety-related EDG. Number 1A EDG is a
tandem engine-single generator set, and is physically very different
from the existing single engine-generator Fairbanks Morse EDGs.
However, the 4.16 kV three-phase rated electrical output is the same
as that provided by the Fairbanks Morse EDGs to the other ESF
busses. The excess capacity of No. 1A EDG will allow the operators
greater flexibility in choosing post-accident discretionary loads,
but will not cause any detrimental effects to the ESF busses or the
equipment served by those busses. Operation of No. 1A EDG in
accordance with these proposed Technical Specifications will not
jeopardize the operation of any other plant systems. The design of
the Fire Protection System in the Diesel Generator Building meets
the Codes and Standards specified in the mechanical, and
instrumentation and controls design reports, previously approved by
the Commission. Inclusion of components from these systems into the
Technical Specifications is consistent with Calvert Cliffs current
licensing basis. Furthermore, locating No. 1A EDG and its fuel oil
supply in a separate Category I building provides additional
assurance that this equipment will not become an initiator of any
accident.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The safety function of the EDGs and the ESF electrical system is
to provide a reliable source of electrical power to the safety-
related busses to operate the necessary accident mitigation
equipment, should offsite power be lost. The margin of safety
associated with this safety function is two-fold: (1) a level of
redundancy must be designed into the EDGs and the ESF electrical
system such that the single failure criteria is met; and (2) the
power supplied to the ESF electrical system by the EDGs must be
sufficient to power the necessary accident mitigation equipment,
should offsite power be lost.
The addition of No. 1A EDG provides the plant with an ESF
electrical system configuration consisting of two EDGs dedicated to
each unit, thereby eliminating reliance upon a swing diesel capable
of being aligned to either unit. In the current configuration, the
facility meets the single failure criteria on a ``per site'' basis.
However, as a result of the new four-EDG configuration, each unit
will have redundant diesel generators to supply power to redundant
safety-related equipment required for safe shutdown or accident
mitigation. The revised Fuel Oil System configuration and the
minimum fuel oil volume to be maintained in the fuel oil tanks
supports the safety function of the EDGs, while maintaining the
margin of safety associated with this equipment. Altogether, the new
four-EDG configuration may be considered an increase in the margin
of safety.
Inclusion of Surveillances for the Fire Protection System
components into the Technical Specifications is consistent with
Calvert Cliffs current licensing basis, and ensures that adequate
fire detection and suppression capability is available to identify
and extinguish fires in the Diesel Generator Building, thereby
reducing the potential for damage to No. 1A EDG and its auxiliaries.
The Diesel Generator Building and its Fire Protection System is
designed so that smoke and heat from a fire in that building will
not impact the redundant safety-related Emergency Diesel Generator
in the Auxiliary Building.
At the completion of the modifications to increase the
capacities of the Unit 2 EDGs and to install the new No. 1A EDG, we
will have diesel generators with more available margin than
currently exists. This will provide the operators with more
flexibility during conditions where the diesel generators are
providing onsite power. The higher electrical capacities will result
in an increase in the margin between the EDGs' electrical capacities
and the electrical power required to operate safety-related
equipment required for safe shutdown or accident mitigation.
Therefore, these modifications may be considered an increase in the
margin of safety.
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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: November 30, 1995
Description of amendments request: The proposed amendments would
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2,
Technical Specifications (TSs) to allow the installation of tube
sleeves as an alternative to plugging for repairing steam generator
(SG) tubes. The proposed changes to TS 3/4.4.5, ``Steam Generators,''
and their supporting Bases would permit tube sleeving repair techniques
developed by Westinghouse Electric Corporation and ABB Combustion
Engineering, Inc., to be used as a repair method for the SGs at the
Calvert Cliffs site.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The implementation of the proposed steam generator tube sleeving
has been reviewed for impact on the current CCNPP [Calvert Cliffs
Nuclear Power Plant] licensing basis.
Since the sleeve dimensions, materials, and connecting joints to
the existing tube are designed to the applicable American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, the
proposed sleeving
[[Page 177]]
repair acts as an in-kind substitution for the original steam generator
tubing. The applicable design criteria for the sleeves conform to
the stress limits and margins of safety of Section III of the ASME
Code. Safety factors of 3 for normal operation and 1.5 for accident
conditions were applied to the design. Mechanical testing using the
ASME Code stress allowables has been performed in support of the
design. Based on the results of Westinghouse and ABB-Combustion
Engineering analytical and test programs, the sleeves fulfill their
intended function as leak tight structural members and meet or
exceed all design criteria.
Evaluation of the proposed sleeved tubes indicates no
detrimental effects on the sleeve or sleeve-tube assembly from
reactor system flow, primary or secondary coolant chemistries,
thermal conditions or transients, or pressure conditions or
transients as may be experienced at CCNPP. Corrosion testing of
sleeve-tube assemblies indicate no evidence of sleeve or tube
corrosion considered detrimental under anticipated service
conditions.
The installation of the proposed sleeves is controlled via the
sleeving vendors' proprietary processes and equipment. The ABB
Combustion Engineering process has been in use since 1984, and has
been implemented 24 times for the installation of over 4,200
sleeves. The Westinghouse process has been in use since 1988, and
approximately 12,000 laser welded sleeves have been installed
between 1988 and 1994. The CCNPP steam generator design was reviewed
and found to be compatible with both installation processes and
equipment.
The implementation of the proposed sleeves has no significant
effect on either the configuration of the plant, or the manner in
which it is operated. The hypothetical consequences of failure of
the sleeved tube is bounded by the current steam generator tube
rupture analysis described in Section 14.15 of the Calvert Cliffs
Updated Final Safety Analysis Report.
Therefore, BGE [Baltimore Gas and Electric] has concluded that
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) [The proposed amendment] would not create the possibility of
a new or different kind of accident from any other accident
previously evaluated.
As discussed above, the structural integrity, thermal
characteristics, and material properties of the proposed sleeves are
consistent with the existing plant steam generators. Therefore, the
functions of the steam generators will not be significantly affected
by the installation of the proposed sleeves. In addition, the
proposed sleeves do not interact with any other plant systems. The
continued integrity of the installed sleeve is periodically verified
by the Technical Specification requirements. The implementation of
the proposed sleeves has no significant effect on either the
configuration of the plant, or the manner in which it is operated.
Therefore, BGE concludes that this proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
(3) [The proposed amendment] would not involve a significant
reduction in a margin of safety.
The repair of degraded steam generator tubes via the use of the
proposed sleeves has been confirmed to restore the structural
integrity of the faulted tube under normal operating and postulated
accident conditions. The design safety factors utilized for the
sleeves are consistent with the safety factors in the ASME Boiler
and Pressure Vessel Code used in the original steam generator
design. The repair limit for the proposed sleeves is consistent with
that established for the steam generator tubes. The design of the
sleeve to tube joints is verified by testing to preclude significant
leakage during normal and postulated accident conditions. Use of the
previously identified design criteria and design verification
testing assures that the margin to safety with respect to the
implementation of the proposed sleeves is not significantly
different from the original steam generator tubes.
Therefore, BGE concludes that 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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: December 7, 1995
Description of amendments request: The proposed amendments would
change the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2,
Technical Specifications (TSs) by adding an analysis technique to the
list of approved core operating limits analytical methods.
Specifically, these amendments would add the convolution analysis
technique to the list of approved methodologies in TSs 6.9.1.9.b. The
convolution analysis technique has already been reviewed and approved
by the NRC staff and the supporting safety evaluation was provided to
the licensee by an NRC letter dated May 11, 1995.
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 change has been evaluated against the standards in 10 CFR
50.92 and has been determined to not involve a significant hazards
consideration in that operation of the facility in accordance with
the proposed amendment:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change is to add the convolution analysis technique
previously approved by the NRC to the list of approved methodologies
in Calvert Cliffs' Unit 1 and 2 Technical Specifications. By letter
dated November 1, 1994, Baltimore Gas and Electric Company (BGE)
requested approval to use the ABB/Combustion Engineering (ABB/CE)
convolution technique for determining the values in the Calvert
Cliffs Core Operating Limits Report (COLR) related to the pre-trip
main steam line break event. Approval was given by the NRC in their
letter dated May 11, 1995. The addition of this technique to the
list of approved analytical methods in Technical Specification
6.9.1.9.b is simply intended to identify it as an approved
methodology. Therefore, the change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed change is to add the convolution analysis technique
previously approved by the NRC to the list of approved methodologies
in Calvert Cliffs' Unit 1 and 2 Technical Specifications. By letter
dated November 1, 1994, BGE requested approval to use the ABB/CE
convolution technique for determining the values in the Calvert
Cliffs COLR related to the pre-trip main steam line break event.
Approval was given by the NRC in their letter dated May 11, 1995.
The addition of this technique to the list of approved analytical
methods in Technical Specifications 6.9.1.9.b is simply intended to
identify it as an approved methodology. Therefore, the change would
not create the possibility of a new or different type of accident
from any accident previously evaluated.
3. Would not involve a significant reduction in the margin of
safety.
The proposed change is to add the convolution analysis technique
previously approved by the NRC to the list of approved methodologies
in Calvert Cliffs' Unit 1 and 2 Technical Specifications. By letter
dated November 1, 1994, BGE requested approval to use the ABB/CE
convolution technique for determining the values in the Calvert
Cliffs COLR related to the pre-trip main steam line break event.
Approval was given by the NRC in their letter dated May 11, 1995.
The addition of this technique to the list of approved analytical
methods in Technical Specification 6.9.1.9.b is simply intended to
identify it as an approved methodology. Therefore, operation of the
facility in accordance with the proposed amendment
[[Page 178]]
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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library,
Prince Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley
Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of amendment request: December 7, 1995
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 3.4.5 and 3.4.6.2 and their Bases
to maintain voltage-based steam generator tube repair criteria for the
tube support plate elevations beyond the current cycle of operation.
The proposed amendment would implement a 2.0 volt repair limit to
replace a 1.0 volt repair limit which was approved on an interim basis
for only the current fuel cycle by License Amendment No. 184 [issued
February 3, 1995]. The proposed amendment would also include changes in
addition to those incorporated by License Amendment No. 184 to reflect
the guidance provided in NRC Generic Letter (GL) 95-05, ``Voltage-Based
Repair Criteria for Westinghouse Steam Generator Tubes Affected by
Outside Diameter Stress Corrosion Cracking.''
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?
Tube burst criteria are inherently satisfied during normal
operating conditions due to the proximity of the tube support plate
(TSP). Test data indicates that tube burst cannot occur within the
TSP, even for tubes which have 100% throughwall electric discharge
machining notches, 0.75 inch long, provided that the TSP is adjacent
to the notched area. Since tube-to-TSP proximity precludes tube
burst during normal operating conditions, use of the criteria must
retain tube integrity characteristics which maintain a margin of
safety of 1.43 times the bounding faulted condition, main steamline
break (MSLB) pressure differential. As previously stated, the
Regulatory Guide (RG) 1.121 criterion requiring maintenance of a
safety factor of 1.43 times the MSLB pressure differential on tube
burst is satisfied by 7/8'' diameter tubing with bobbin coil
indications with signal amplitudes less than 8.82 volts, regardless
of the indicated depth measurement.
The upper voltage repair limit (Vurl) will be determined
prior to each outage using the most recently approved NRC database
to determine the tube structural limit (Vsl). The structural
limit is reduced by allowances for nondestructive examination (NDE)
uncertainty (Vnde) and growth (Vgr) to establish
Vurl. Using Generic Letter (GL) 95-05 and growth allowances for
an example, the NDE uncertainty component of 20% and a voltage
growth allowance of 30% per full power year can be utilized to
establish a Vurl of 5.9 volts. The 20% NDE uncertainty
represents a square-root-sum-of-the-squares (SRSS) combination of
probe wear uncertainty and analyst variability. The degradation
growth allowance should be an average growth rate or 30% per
effective full power year, whichever is larger. This growth
allowance is conservative for BVPS-1 [Beaver Valley Power Station,
Unit No. 1] as the percent voltage growth rates have decreased for
each of the last three inspections.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated MSLB
outside of containment but upstream of the main steam isolation
valve (MSIV) represents the most limiting radiological condition
relative to the plugging criteria. In support of implementation of
the revised plugging limit, analyses will be performed to determine
whether the distribution of cracking indications at the tube support
plate intersections during future cycles are projected to be such
that primary-to-secondary leakage would result in postulated site
boundary and control room doses exceeding 10 CFR 100, and 10 CFR 50,
Appendix A, GDC-19 requirements, respectively. A separate
calculation has determined the maximum allowable MSLB leakage limit
in a faulted loop. This limit was calculated using the technical
specification reactor coolant system (RCS) Iodine-131 activity level
of 1.0 microcuries per gram dose equivalent Iodine-131 and the
recommended Iodine-131 transient spiking values consistent with
NUREG-0800. The projected MSLB leakage rate calculation methodology
prescribed in Section 2.b of GL 95-05 will be used to calculate the
end-of-cycle (EOC) leakage. Projected EOC voltage distribution will
be developed using the most recent EOC eddy current results and
considering an appropriate voltage measurement uncertainty. The log-
logistic probability of leakage correlation will be used to
establish the MSLB leakrate used for comparison with the faulted
loop allowable limit. Due to the relatively low voltage levels of
indications at BVPS-1 and low voltage growth rates, it is expected
that the calculated leakage values will not exceed this limit.
Therefore, as implementation of the 2.0 volt voltage-based plugging
criteria at BVPS-1 does not adversely affect steam generator tube
integrity and implementation will be shown to result in acceptable
dose consequences, the proposed amendment does not result in any
increase in the probability or consequences of an accident
previously evaluated in the UFSAR [Updated Final Safety Analysis
Report].
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Implementation of the proposed steam generator tube 2.0 volt
plugging limit does not introduce any significant changes to the
plant design basis. Use of the 2.0 volt plugging limit does not
provide a mechanism which could result in an accident outside of the
region of the tube support plate elevations as no outside diameter
stress corrosion cracking (ODSCC) is occurring outside the thickness
of the tube support plates. Neither a single or multiple tube
rupture event would be expected in a steam generator in which the
plugging limit has been applied (during all plant conditions).
Duquesne Light Company will continue to implement a maximum
primary-to-secondary leakage rate limit of 150 gpd [gallons per day]
per steam generator to help preclude the potential for excessive
leakage during all plant conditions. The RG 1.121 criterion for
establishing operational leakage rate limits that require plant
shutdown are based upon leak-before-break considerations to detect a
free span crack before potential tube rupture during faulted plant
conditions. The 150 gpd limit provides for leakage detection and
plant shutdown in the event of the occurrence of an unexpected
single crack resulting in leakage that is associated with the
longest permissible crack length. RG 1.121 acceptance criteria for
establishing operating leakage limits are based on leak-before-break
considerations such that plant shutdown is initiated if the leakage
associated with the longest permissible crack is exceeded.
The single through-wall crack lengths that result in tube burst
at 1.43 times the MSLB pressure differential and the MSLB pressure
differential alone are approximately 0.57 inch and 0.84 inch,
respectively. A leak rate of 150 gpd will provide for detection of
0.41 inch long cracks at nominal leak rates and 0.62 inch long
cracks at the lower 95% confidence level leak rates. Since tube
burst is precluded during normal operation due to the proximity of
the TSP to the tube and the potential exists for the crevice to
become uncovered during MSLB conditions, the leakage from the
maximum permissible crack must preclude tube burst at MSLB
conditions. Thus, the 150 gpd limit provides for plant shutdown
prior to reaching critical crack lengths for MSLB conditions using
the lower 95% leakrate data. Additionally, this leak-before-break
evaluation assumes that the entire crevice area is uncovered during
blowdown. Partial uncovery will provide benefit to the burst
capacity of the intersection. Analyses have shown that only a small
percentage of the TSPs are deflected greater than the TSP thickness
during a postulated MSLB.
As steam generator tube integrity upon implementation of the 2.0
volt plugging limit
[[Page 179]]
continues to be maintained through inservice inspection and primary-to-
secondary leakage monitoring, the possibility of a new or different
kind of accident from any accident previously evaluated is not
created.
3. Does the change involve a significant reduction in a margin
of safety?
The use of the voltage-based bobbin probe tube support plate
elevation plugging criteria at BVPS-1 maintains steam generator tube
integrity commensurate with the criteria of RG 1.121. This guide
describes a method acceptable to the Commission for meeting GDCs
[General Design Criterion] 14, 15, 30, 31, and 32 by reducing the
probability or the consequences of steam generator tube rupture.
This is accomplished by determining the limiting conditions of
degradation of steam generator tubing, as established by inservice
inspection, for which tubes with unacceptable cracking should be
removed from service. Upon implementation of the proposed criteria,
even under the worst case conditions, the occurrence of ODSCC
[Outside Diameter Stress Corrosion Cracking] at the tube support
plate elevations is not expected to lead to a steam generator
tuberupture event during normal or faulted plant conditions. The EOC
distribution of crack indications at the tube support plate
elevations will be confirmed to result in acceptable primary-to-
secondary leakage during all plant conditions and that radiological
consequences are not adversely impacted.
In addressing the combined effects of loss-of-coolant-accident
(LOCA) + safe shutdown earthquake (SEE) on the steam generator
component (as required by GDC 2), it has been determined that tube
collapse may occur in the steam generators at some plants. This is
the case as the tube support plates may become deformed as a result
of lateral loads at the wedge supports at the periphery of the plate
due to the combined effects of the LOCA rarefaction wave and SSE
loadings. Then, the resulting pressure differential on the deformed
tubes may cause some of the tubes to collapse. There are two issues
associated with steam generator tube collapse. First, the collapse
of steam generator tubing reduces the RCS [reactor coolant system]
flow area through the tubes. The reduction in flow area increases
the resistance to flow of steam from the core during a LOCA which,
in turn, may potentially increase peak clad temperature. Second,
there is a potential that partial through-wall cracks in tubes could
progress to complete through-wall cracks during tube deformation or
collapse.
The results of an analysis using the larger break inputs show
that the LOCA loads were found to be of insufficient magnitude to
result in steam generator tube collapse or significant deformation.
Since the leak-before-break methodology is applicable to BVPS-1
reactor coolant loop piping, the probability of breaks in the
primary loop piping is sufficiently low that they need not be
considered in the structural design of the plant. The limiting LOCA
event becomes either the accumulator line break or the pressurizer
surge line break. Analysis results provided in WCAP-14122, dated
July 1994, demonstrate that no tubes were subject to deformation or
collapse. No tubes have been excluded from application of the
subject voltage-based steam generator plugging criteria.
Addressing RG 1.83 considerations, implementation of the bobbin
probe voltage-based tube plugging criteria of 2.0 volts is
supplemented by: enhanced eddy current inspection guidelines to
provide consistency in voltage normalization, a 100% eddy current
inspection sample size at the tube support plate elevations, and
rotating pancake coil inspection requirements for the larger
indications left inservice to characterize the principal degradation
as ODSCC.
As noted previously, implementation of the tube support plate
intersection voltage-based plugging criteria will decrease the
number of tubes which must be repaired. The installation of steam
generator tube plugs reduces the RCS flow margin. Thus, the
implementation of the 2.0 volt plugging limit will maintain the
margin of flow that would otherwise be reduced in the event of
increased tube plugging.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin with respect to plant safety as defined in the UFSAR or any
BASES of the plant technical specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: John F. Stolz
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: December 15, 1995
Description of amendment request: The proposed amendments would (1)
revise Technical Specifications (TSs) 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3,
3/4.6.1.6, and associated Bases, (2) delete TS 6.9.2.g, and (3) add a
new TS 6.17. The proposed changes would make the TSs consistent with
Option B of recently revised Appendix J of 10 CFR Part 50 and the
implementing guidance of Regulatory Guide 1.163, ``Performance-Based
Containment Leak Test Program,'' dated September 1995. Option B of
Appendix J permits licensees to implement a performance based option
rather than the previous prescriptive requirements now contained in
Appendix J as Option A. The proposed amendments would remove from the
TSs the prescriptive requirements of Option A concerning test
frequencies and test methodology and would also include minor
administrative and editorial changes to add consistency between the
Bases and the TSs and to provide additional clarification.
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?
Containment leakage is not an accident initiator. The proposed
amendment does not add or modify any existing plant equipment.
Therefore there is no increase in the probability of an accident
previously evaluated.
The consequences of an accident previously evaluated are not
significantly increased. The proposed changes do not affect the
assumptions, parameters or result of any Updated Final Safety
Analysis (UFSAR) accident analyses. The containment leakage rate
will continue to be maintained within the limit assumed in the
accident analysis for a Design Basis Accident (DBA). The proposed
changes do not modify the response of the containment during a DBA.
The proposed amendment will continue to ensure that the ability of
the containment structure, including the containment air locks, to
limit leakage from a DBA is demonstrated using test methodologies
and guidance on test frequencies that have been determined to be
acceptable to meet the requirements of 10 CFR 50, Appendix J, Option
B.
The potential increase to overall accident risk due to the
containment leak tightness decreasing between extended testing
intervals and the resulting potential increased radioactivity
release to the environment during a DBA has been determined to be
minimal based on the findings of NUREG 1493 titled ``Performance-
Based Containment Leak-Test Program.'' In addition, due to the
performance based nature of 10 CFR 50 Appendix J, Option B, the
extended test intervals are utilized only when the component(s) have
demonstrated an acceptable performance history. Therefore, a
significant decrease in containment leak tightness between extended
test intervals is not expected as a result of this proposed change.
Based on the above discussion, it is concluded that this change
does 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?
The proposed change does not involve any physical changes to the
plant or changes in
[[Page 180]]
plant operating configuration. The proposed amendment involves changes
to plant programs and administrative requirements used in
determining acceptable containment performance. The performance of
plant systems, including the containment structure, during plant
operation remains unchanged.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The margin of safety is not significantly reduced by this
proposed change. The acceptance criteria for ``as left'' measured
containment leakage rates is not being increased as result of this
proposed amendment. For Beaver Valley Power Station (BVPS) Unit No.
1 only, the ``as found'' maximum allowable overall Type A leakage
rate is being slightly increased. However, the slight increase does
not exceed the value assumed in accident analysis for containment
leakage during a DBA due to changing the acceptance criteria from
less than to less than or equal to. The margin between the
acceptable ``as left'' measured overall Type A containment leakage
rate and the leakage rate assumed in the accident analysis is not
being decreased.
The maximum ``as found'' allowable overall Type A leakage rate
remains unchanged for BVPS Unit No. 2. The margin between the
acceptable ``as left'' measured overall Type A containment leakage
rate and the leakage rate assumed in the accident analysis is also
not being decreased.
The maximum allowable measured combined Type B and C leakage
rate is not being increased above the current limits.
The maximum peak containment pressure following a DBA remains
unchanged. The containment depressurization time following a DBA
remains unchanged. The calculated offsite dose consequences of a DBA
remains unchanged.
The proposed amendment continues to ensure reactor containment
system reliability by periodic testing in compliance with 10 CFR 50,
Appendix J, Option B. The extension of Type A, B and C test
frequencies permitted by 10 CFR 50 Appendix J, Option B, is not
expected to result in a significant decrease in containment leak
tightness between test intervals. Due to the performance based
nature of 10 CFR 50 Appendix J, Option B, the extended test
intervals are utilized only when the component(s) have demonstrated
an acceptable performance history. Therefore, a significant decrease
in containment leak tightness between extended test intervals is not
expected as a result of this proposed change.
The changes which are either administrative or editorial in
nature will not reduce the margin of safety because they have no
impact on any safety analysis assumptions.
Therefore, based on the above discussion, it can be concluded
that 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, Pennsylvania 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 19, 1995, as supplemented by letter
dated December 7, 1995.
Description of amendment request: May 19, 1995, submittal requested
to modify Action Statement for Technical Specification (TS) 3.6.4.2 for
the hydrogen recombiners. It also requested to make the surveillance
requirements for hydrogen recombiners consistent with NUREG-1432,
``Standard Technical Specifications Combustion Engineering Plants.''
The December 7, 1995, letter withdrew the request to change the Action
Statement for TS 3.6.4.2.
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 response is predicated on the following technical bases: (1)
the current licensing basis of record establishes that only one
recombiner system is required to maintain hydrogen concentration
below 4%, (2) the proposed technical specification changes are
conservative when compared with the recommendations of Regulatory
Guide 1.7, (3) short term post LOCA hydrogen generation is less than
1%, (4) long term post LOCA hydrogen generation is less than the
flame propagation limit, which according to Regulatory Guide 1.7
would not result in adverse effects to containment systems, and (5)
a design basis LOCA without long term hydrogen control would produce
pressures below the containment design pressure.... Therefore, the
proposed change will not involve a significant increase in the
probability or consequences of any accident previously evaluated.
The proposed change will not alter the configuration or
operation of any other plant system or component. The change does
not involve any change to the operational design or limits of any
other plant systems or components. Thus, no new failure modes are
introduced or associated with the proposed change. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change will have no adverse impact on the
protective boundaries, safety limits, or margin or safety. There are
no limits or margins of safety being revised for any systems,
components, or protective boundaries.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 7, 1995
Description of amendment request: Amendment to Technical
Specification (TS) 3/4.8.1 ``Electrical Power Systems - AC Sources''
and the associated TS BASES. The proposed amendment would implement
selected changes from NUREG 1432, ``Standard Technical Specifications
Combustion Engineering Plants,'' Generic Letter (GL) 94-01, ``Removal
of Accelerated Testing and Special Reporting Requirements for Emergency
Diesel Generators,'' and GL 93-05, ``Line-Item Technical Specifications
Improvements to Reduce Surveillance Requirements for Testing During
Power Operation.'' The intent of these changes is to increase Emergency
Diesel Generator (EDG) reliability by reducing the stresses on the EDGs
caused by unnecessary testing. This proposed TS amendment will also
relocate the Surveillance Requirements for maintaining the properties
of the fuel oil to TS Section 6, ``Administrative Controls.'' These
requirements will be implemented as part of the Fuel Oil Testing
Program. In addition, the requirement for cleaning the diesel fuel oil
storage tanks with a sodium hypochlorite solution or equivalent will be
changed to also allow an appropriate mechanical method (such as
pressure washing or manual wiping) to be utilized.
Basis for proposed no significant hazards consideration
determination:
[[Page 181]]
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 Standby Diesel Generators do not initiate any accidents,
therefore the proposed changes do not increase the probability of an
accident previously evaluated. The proposed changes to TS 3/4.8.1
and the associated BASES affect the required actions in response to
inoperable offsite and onsite AC sources, Surveillance Requirements
for the EDG, and reporting requirements for EDG failures. The
majority of the proposed changes are based on the recommendations of
NUREG 1432, GL 94-01, and GL 93-05. These proposed changes have been
extensively reviewed by the NRC during the preparation of these
documents and by Waterford 3 SES during the development of this
request for TS amendment. The proposed changes are expected to
result in improvements in EDG performance and reduce EDG aging due
to excessive testing. The proposed changes will permit the
elimination of the unnecessary mechanical stress and wear on the
EDGs while ensuring that the EDGs will perform their design
function. The elimination of mechanical stress and wear will improve
reliability and availability of the EDGs which will have a positive
effect on the ability of the EDGs to perform their design function.
The proposed changes do not affect the availability or the testing
requirements of the offsite circuits.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The proposed changes to TS 3/4.8.1 and the associated Bases do
not introduce any new modes of plant operation or new accident
precursors, involve any physical alterations to plant
configurations, or make any changes to system setpoints which could
initiate a new or different kind of accident. The proposed changes
do not affect the design or performance characteristics of any EDG
or its ability to perform its design function. No new failure modes
have been defined and no new system interactions have been
introduced for any plant system or component. In addition, there
have not been any new limiting failures identified as a result of
the proposed changes. The proposed changes will eliminate
unnecessary EDG testing and will increase EDG reliability and
availability. This will have an overall positive affect on plant
safety. Accidents concerning loss of offsite power and a single
failure (e.g., loss of an EDG) have previously been evaluated. These
changes are intended to improve plant safety, decrease equipment
degradation, and remove an unnecessary burden on personnel resources
by reducing the amount of testing that the TS requires during power
operation.
Relocating the diesel fuel oil testing requirements to the
Waterford 3 Fuel Oil Testing Program outside of the Technical
Specifications is an administrative change only and consequently has
no effect on accident probability, consequences, or margin. Also,
the proposed cleaning method for the diesel fuel oil storage tanks
meets the intent of Regulatory Guide 1.137 and will not result in
the degradation of the fuel oil.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Under the proposed changes to TS 3/4.8.1 and the associated
Bases, the EDGs will remain capable of performing their safety
function. The changes do not affect the design or performance of the
EDGs, but will increase EDG reliability and availability by reducing
the stresses and the effects of aging on the EDG by eliminating
unnecessary testing. This will result in an overall increase in
plant safety. The ability of the EDGs to perform their safety
function will not be degraded. Relocating the diesel fuel oil
testing requirements to the Waterford 3 Fuel Oil Testing Program
outside of the Technical Specifications is an administrative change
only and consequently has no effect on accident probability,
consequences, or margin. Also, the proposed cleaning method for the
diesel fuel oil storage tanks meets the intent of Regulatory Guide
1.137 and will not result in a reduction in the margin of safety.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: (TS 93-09) December 8, 1995
Description of amendment request: The proposed change would revise
the setpoints and time delays for the auxiliary feedwater loss-of-power
and 6.9-kv shutdown board loss-of-voltage and degraded-voltage
instrumentation setpoints in Items 6 and 7 of Technical Specification
Table 3.3-4, respectively.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revision supports the implementation of design
logic and setpoint changes to the loss-of-power relaying. This
relaying is designed to ensure adequate voltage is available to
safety-related loads in order to enhance their operability and
support accident mitigation functions and to provide for auxiliary
feedwater (AFW) pump starts. The design changes alter relay logic
and delete unnecessary relaying, but do not change the diesel
generator (D/G) start and load-shedding actuations that result from
loss-of-power conditions. Therefore, no new actuations or functions
have been created; and because the existing and proposed functions
provide for accident mitigation considerations that are not the
source of an accident, the probability of an accident is not
increased. The deletion of the 6.9-kilovolt shutdown board normal-
feedwater undervoltage relays actually reduces the potential for
inadvertent shutdown board blackouts as a result of short-duration
voltage transients or instrument failures.
The setpoints and time delays for loss-of-power functions have
been modified based on the guidelines developed by the Electrical
Distribution System Clearinghouse as evaluated and determined
through detailed analysis by TVA. This design is documented in TVA
Calculations SQN-EEB-MS-T106-0008, 27DAT, and DS-1-2 and is
available for NRC review at the SQN site. The assigned values are
conservative settings that will ensure adequate voltage is supplied
to safety-related loads for accident mitigation and safety functions
under normal, degraded, and loss-of-offsite power voltage conditions
with appropriate time delays to prevent damage to electrical loads
and minimize premature or unnecessary actuations. The identification
of loss-of-voltage conditions is enhanced by the design changes to
ensure the timely sequencing of loads onto the D/G and the
initiation of AFW pump starts for accident mitigation. Because there
are no reductions in safety functions resulting from the design
logic, setpoint and time-delay changes to the loss-of-power
instrumentation and offsite dose levels for postulated accidents
will not be increased, the consequences of an accident are not
increased.
The applicable mode addition, TS 3.0.4 exclusion deletion, and
response time measurement clarification incorporated in the proposed
change do not affect plant functions. These changes reflect the
requirements that SQN has been maintaining and serve to clarify the
requirements to provide consistency of application and easier
understanding. The AFW footnote addition and bases revision only
clarify operability conditions that are consistent with the plant
design for the AFW pump and loss-of-power instrumentation. Because
there are no changes to plant functions or operations, these
revisions have no impact on accident probabilities or consequences.
[[Page 182]]
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
As described above, the loss-of-power instrumentation ensures
adequate voltage to safety-related loads by initiating D/G starts
and load shedding and provides for AFW pump starting, but is not
considered to be the source of an accident. Although the design
logic, setpoint, and time-delay actuation criteria have changed, the
output functions to various plant systems that actuate for load
shedding and D/G starts remain the same. Therefore, actuation
criteria have been affected, but not safety functions, and the TVA
evaluation has confirmed that the new design enhances the ability to
maintain adequate voltage to support safety functions. Since safety
functions have not changed and the new loss-of-power instrumentation
design continues to support operability of safety-related equipment,
no new or different accident is created.
The applicable mode addition, TS 3.0.4 exclusion deletion, and
response time measurement clarification, as well as the AFW
operability clarifications, do not affect plant functions and will
not create a new accident.
3. Involve a significant reduction in a margin of safety.
The proposed loss-of-power TS changes support design logic,
setpoint, and time-delay requirements that have been verified by TVA
analysis to provide acceptable voltage levels for safety-related
components. In determining the acceptability of these voltage
levels, the minimum voltage for operation as well as detrimental
component heating resulting from sustained degraded-voltage
conditions were considered. This design ensures that safety-related
loads will be available and operable for normal and accident plant
conditions. The applicable mode addition, TS 3.0.4 exclusion
deletion, response time measurement clarification, and AFW
operability clarifications provide enhancements to TS requirements
and do not affect plant functions. Therefore, no safety functions
are reduced by these changes and there is no reduction in the margin
of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: (TS 95-20) December 8, 1995
Description of amendment request: The proposed change would revise
Surveillance Requirements 4.6.2.1.1.d and 4.6.2.1.2.b to extend the
containment spray nozzle air or smoke flow tests from the present 5-
year interval to a 10-year interval, in accordance with Generic Letter
93-05.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The TS change is consistent with the guidance provided in
Generic Letter 93-05. Containment spray (CS) systems' header piping
is stainless steel; therefore, corrosion will be negligible during
the extended surveillance interval. Since the CS systems' headers
are maintained dry, there is no mechanism that could cause blockage
of the spray nozzles. Therefore, the nozzles in the CS systems will
remain operable, during the 10-year surveillance interval, to
mitigate the consequence of an accident previously evaluated.
Additionally, clogging or blockage has not been observed during the
5-year surveillance tests that have been performed in the past at
SQN. Testing the CS systems' nozzles at the proposed reduced
frequency will not increase the probability of occurrence of a
postulated accident or the consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed reduced frequency testing of the CS systems'
nozzles does not change the manner in which these systems are
operated. The reduced testing frequency of the spray nozzles does
not generate any new accident precursors. Therefore, the possibility
of a new or different kind of accident previously evaluated is not
created by the proposed changes in surveillance frequency of the CS
system's nozzles.
3. Involve a significant reduction in a margin of safety.
Reduced testing of the CS systems' nozzles does not change the
way the systems are operated or the systems' operability
requirements. In this application, any additional corrosion of
stainless steel piping will be negligible during the extended
surveillance interval. Since the CS systems are maintained dry,
there is no additional mechanism that could cause blockage of the
nozzles. Therefore, the proposed reduced testing frequency is
adequate to ensure spray nozzle operability. The surveillance
requirements do not affect the margin of safety since the
operability requirements of both the CS systems remains unchanged.
The existing safety analysis remains bounding. Therefore, there is
no reduction in the margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: December 8, 1995 (TS 95-24)
Description of amendment request: The proposed change would modify
various Technical Specification requirements in order to implement the
recent rule change to 10 CFR Part 50, Appendix J. The new Appendix J
rule (Option B) provides a voluntary performance based testing option
for containment leakage rate testing (CLRT). Option B CLRT requirements
are based on system and component performance in lieu of compliance
with the current prescriptive requirements. Option B allows extension
of the integrated leakage rate test (Type A test) frequency based on an
acceptable past history. For Type B and Type C local leak rate test,
Option B allows extension of the test frequency based on plant-specific
experience history of each component and establishes controls to ensure
continued performance during extended testing intervals.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria
[[Page 183]]
established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant
(SQN) in accordance with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment to SQN TSs is in accordance with Option B
to 10 CFR 50, Appendix J. The proposed amendment adds a voluntary
performance based option for containment leak rate testing. The
changes being proposed do not affect the precursor for any accident
or transient analyzed in Chapter 15 of SQN Updated Final Safety
Analysis Report. The proposed change does not increase the total
allowable primary containment leakage rate. The proposed change does
not reflect a revision to the physical design and/or operation of
the plant. Therefore, operation of the facility, in accordance with
the proposed change, does not significantly affect the probability
or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed amendment to SQN TSs is in accordance with the new
performance-based option (Option B) to 10 CFR 50, Appendix J. The
changes being proposed will not change the physical plant or the
modes of operation defined in the facility license. The proposed
changes do not increase the total allowable primary containment
leakage rate. The changes do not involve the addition or
modification of equipment, nor do they alter the design or operation
of plant systems. Therefore, operation of the facility in accordance
with the proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to SQN TSs is in accordance with the new
option to 10 CFR 50, Appendix J. The proposed option is formulated
to adopt performance-based approaches. This option removes the
current prescriptive details from the TS. The proposed changes do
not affect plant safety analyses or change the physical design or
operation of the plant. The proposed change does not increase the
total allowable primary contaiment leakage rate. Therefore,
operation of the facility, in accordance with the proposed change,
does not involve a significant reduction in the margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear
Plant, Unit 2, Hamilton County, Tennessee
Date of amendment request: December 12, 1995 (TS 95-23)
Description of amendment request: The proposed change would
incorporate new requirements associated with steam generator tube
inspections and repair. The new requirements would establish alternate
steam generator tube plugging criteria at the tube support plate
intersections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Testing of model boiler specimens for free-span tubing (no tube
support plate restraint) at room temperature conditions shows burst
pressures in excess of 5,000 pounds per square inch (psi) for
indications of outer diameter stress corrosion cracking with voltage
measurements as high as 19 volts. Burst testing performed on
intersections pulled from SQN with up to a 1.9-volt indication shows
measured burst pressure in excess of 6,600 psi at room temperature.
Burst testing performed on pulled tubes from other plants with up to
7.5-volt indications shows burst pressures in excess of 5,200 psi at
room temperatures. Correcting for the effects of temperature on
material properties and minimum strength levels (as the burst
testing was done at room temperature), tube burst capability
significantly exceeds the safety-factor requirements of NRC
Regulatory Guide (RG) 1.121.
Tube burst criteria are inherently satisfied during normal
operating conditions because of the proximity of the tube support
plate (TSP). Since tube-to-tube support plate proximity precludes
tube burst during normal operating conditions, use of the criteria
must retain tube integrity characteristics that maintain a margin of
safety of 1.43 times the bounding faulted condition steam line break
(SLB) pressure differential. During a postulated SLB, the TSP has
the potential to deflect during blowdown following a main SLB,
thereby uncovering the TSP intersections.
Based on the existing database, the RG 1.121 criterion requiring
maintenance of a safety factor of 1.43 times the SLB pressure
differential on tube burst is satisfied by 7/8-inch-diameter tubing
with bobbin coil indications with signal amplitudes less than 8.82
volts (WCAP-13990), regardless of the indicated depth measurement. A
2.0-volt plugging criterion (resulting in a projected end-of-cycle
[EOC] voltage) compares favorably with the 8.82-volt structural
limit considering the extremely slow apparent voltage growth rates
and few numbers of indications at SQN. Using the established
methodology of RG 1.121, the structural limit is reduced by
allowances for uncertainty and growth to develop a beginning of
cycle (BOC) repair limit that would preclude indications at EOC
conditions that exceed the structural limit. The nondestructive
examination (NDE) uncertainty component is 20.5 percent, and is
based on the Electric Power Research Institute (EPRI) alternate
repair criteria (ARC).
Test data indicates that tube burst cannot occur within the TSP,
even for tubes that have 100 percent throughwall electro-discharge
machining notches, 0.75 inch long, provided that the TSP is adjacent
to the notched area. Because of the few number of indications at
SQN, the EPRI methodology of applying a growth component of 35
percent per effective full power year (EFPY) will be used. Near-term
operating cycles at SQN are expected to be bounded by 1.23 years,
therefore, a 43 percent growth component is appropriate. When these
allowances are added to the BOC alternate plugging criteria (APC) of
2.0 volts in a deterministic bounding EOC voltage of approximately
3.26 volts for Cycle 7, operation can be established. A 5.56-volt
deterministic safety margin exists (8.82 structural limit - 3.26-
volt EOC equal 5.56-volt margin).
For the voltage/burst correlation, the EOC structural limit is
supported by a voltage of 8.82 volts. Using this structural limit of
8.82 volts, a BOC maximum allowable repair limit can be established
using the guidance of RG 1.121. The BOC maximum allowable repair
limit should not permit the existence of EOC indications that exceed
the 8.82-volt structural limit. By adding NDE uncertainty allowances
and an allowance for crack growth to the repair limit, the
structural limit can be validated. Therefore, the maximum allowable
BOC repair limit (RL) based on the structural limit of 8.82 volts
can be represented by the expressions:
RL + (0.205 x RL) + (0.43 x RL) = 8.82 volts, or,
the maximum allowable BOC repair limit can be expressed as,
RL = 8.82-volt structural limit/1.64 = 5.4 volts.
This RL (5.4 volts) is the appropriate limit for APC
implementation to repair bobbin indications greater than 2.0 volts
independent of rotating pancake coil (RPC) confirmation of the
indication. This 5.4-volt upper limit for non-confirmed RPC calls is
consistent with other recently approved APC programs (Farley Nuclear
Plant, Unit 2).
The conservatism of the growth allowance used to develop the
repair limit is shown by the most recent SQN eddy current data. Only
seven tubes in Unit 2 required repair because of outside diameter
stress corrosion cracking (ODSCC) at the TSP intersections.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated main
[[Page 184]]
SLB outside of containment, but upstream of the main steam isolation
valve (MSIV), represents the most limiting radiological condition
relative to the APC. Implementation of the APC will determine
whether the distribution of cracking indications at the TSP
intersections is projected to be such that primary-to-secondary
leakage would result in site boundary doses within a small fraction
of the 10 CFR 100 guidelines. A separate analysis has determined
this allowable SLB leakage limit to be 3.7 gallons per minute (gpm)
in the faulted loop. This limit uses the TS reactor coolant system
(RCS) Iodine-131 activity level of 1.0 microcuries per gram dose
equivalent Iodine-131 and the recommended Iodine-131 transient
spiking values consistent with NUREG-0800. The analysis method is
WCAP-14277, which is consistent with the guidance of the NRC generic
letter (GL) [95-05] and will be used to calculate EOC leakage.
Because of the relatively low number of indications at SQN, it is
expected that the actual leakage values will be far less than this
limit. Additionally, the current Iodine-131 levels at SQN range from
about 25 to 100 times less than the TS limit.
Application of the criteria requires the projection of
postulated SLB leakage, based on the projected EOC voltage
distribution for Cycle 8 operation. Projected EOC voltage
distribution is developed using the most recent EOC eddy current
results and a voltage measurement uncertainty. Data indicates that a
threshold voltage of 2.8 volts would result in throughwall cracks
long enough to leak at SLB condition. The GL requires that all
indications to which the APC are applied must be included in the
leakage projection. Tube pull results from another plant with 7/8-
inch tubing with a substantial voltage growth database have shown
that tube wall degradation of greater than 40 percent throughwall
was readily detectable either by the bobbin or RPC probe. The tube
with maximum throughwall penetration of 56 percent (42 average) had
a voltage of 2.02 volts. The SQN Unit 1 pulled tube had a 1.93-volt
indication with a maximum depth of 91 percent and did not leak at
SLB condition. Based on the SQN pulled tube and industry pulled tube
data supporting a lower threshold for SLB leakage of 2.8 volts,
inclusion of all APC intersections in the leakage model is quite
conservative. The ODSCC occurring at SQN is in its earliest stages
of development. The conservative bounding growth estimations to be
applied to the expected small number of indications for the upcoming
inspection should result in very small levels of predicted SLB
leakage. Historically, SQN has not identified ODSCC as a contributor
to operational leakage.
In order to assess the sensitivity of an indication's BOC
voltage to EOC leakage potential, a Monte Carlo simulation was
performed for a 2.0-volt BOC indication.
The maximum EOC voltage (at 99.8 percent cumulative probability)
was found to be 4.8 volts. The leakage component from an indication
of this magnitude, using the EPRI leakage model, is 0.028 gpm.
Therefore, as implementation of the 2.0-volt APC does not
adversely affect steam generator (S/G) tube integrity and
implementation will be shown to result in acceptable dose
consequences, the proposed amendment does not result in significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Implementation of the proposed S/G tube APC does not introduce
any significant changes to the plant design basis. Use of the
criteria does not provide a mechanism that could result in an
accident outside of the region of the TSP elevations; no ODSCC is
occurring outside the thickness of the TSP. Neither a single or
multiple tube rupture event would be expected in a S/G in which the
plugging criteria is applied (during all plant conditions).
TVA will implement a maximum leakage rate limit of 150 gallon
per day per S/G to help preclude the potential for excessive leakage
during all plant conditions. The SQN TS limits on primary-to-
secondary leakage at operating conditions include a maximum of 0.42
gpm (600 gallons per day [gpd]) for all S/Gs, or, a maximum of 150
gpd for any one S/G. The RG 1.121 criterion for establishing
operational leakage rate limits that require plant shutdown is based
upon leak-before-break considerations to detect a free-span crack
before potential tube rupture during faulted plant conditions. The
150-gpd limit should provide for leakage detection and plant
shutdown in the event of the occurrence of an unexpected single
crack resulting in leakage that is associated with the longest
permissible crack length. RG 1.121 acceptance criteria for
establishing operating leakage limits are based on leak-before-break
considerations such that plant shutdown is initiated if the leakage
associated with the longest permissible crack is exceeded. The
longest permissible crack is the length that provides a factor of
safety of 1.43 against bursting at faulted conditions maximum
pressure differential. A voltage amplitude of 8.82 volts for typical
ODSCC corresponds to meeting this tube burst requirement at a lower
95 percent prediction limit on the burst correlation coupled with
95/95 lower tolerance limit material properties. Alternate crack
morphologies can correspond to 8.82 volts so that a unique crack
length is not defined by the burst pressure versus voltage
correlation. Consequently, typical burst pressure versus through-
wall crack length correlations are used below to define the
``longest permissible crack'' for evaluating operating leakage
limits.
The single through-wall crack lengths that result in tube burst
at 1.43 times the SLB pressure differential and the SLB pressure
differential alone are approximately 0.57 inch and 0.84 inch,
respectively. A leak rate of 150 gpd will provide for detection of
0.4-inch-long cracks at nominal leak rates and 0.6-inch-long cracks
at the lower 95 percent confidence level leak rates. Since tube
burst is precluded during normal operation because of the proximity
of the TSP to the tube and the potential exists for the crevice to
become uncovered during SLB conditions, the leakage from the maximum
permissible crack must preclude tube burst at SLB conditions. Thus,
the 150-gpd limit provides for plant shutdown before reaching
critical crack lengths for SL-conditions. Additionally, this leak-
before-break evaluation assumes that the entire crevice area is
uncovered during blowdown. Partial uncover will provide benefit to
the burst capacity of the intersection.
As S/G tube integrity upon implementation of the 2.0-volt APC
continues to be maintained through in-service inspection and
primary-to-secondary leakage monitoring, the possibility of a new or
different kind of accident from any accident previously evaluated is
not created.
3. Involve a significant reduction in a margin of safety.
The use of the voltage based APC at SQN is demonstrated to
maintain S/G tube integrity commensurate with the criteria of RG
1.121. RG 1.121 describes a method acceptable to the NRC Staff for
meeting General Design Criteria (GDC) 14, 15, 31, and 32 by reducing
the probability or the consequences of S/G tube rupture. This is
accomplished by determining the limiting conditions of degradation
of S/G tubing, as established by in-service inspection, for which
tubes with unacceptable cracking should be removed from service.
Upon implementation of the criteria, even under the worst-case
conditions, the occurrence of ODSCC at the TSP elevations is not
expected to lead to a S/G tube rupture event during normal or
faulted plant conditions. The EOC distribution of crack indications
at the TSP elevations will be confirmed to result in acceptable
primary-to-secondary leakage during all plant conditions and
radiological consequences are not adversely impacted.
In addressing the combined effects of loss-of-coolant accident
(LOCA), plus safe shutdown earthquake (SSE) on the S/G component (as
required by GDC 2), it has been determined that tube collapse may
occur in the S/Gs at some plants. This is the case as the TSP may
become deformed as a result of lateral loads at the wedge supports
at the periphery of the plate because of the combined effects of the
LOCA rarefaction wave and SSE loadings. Then, the resulting pressure
differential on the deformed tubes may cause some of the tubes to
collapse.
There are two issues associated with S/G tube collapse. First,
the collapse of S/G tubing reduces the RCS flow area through the
tubes. The reduction in flow area increases the resistance to flow
of steam from the core during a LOCA, which in turn, may potentially
increase peak clad temperature (PCT). Second, there is a potential
that partial through-wall cracks in tubes could progress to through-
wall cracks during tube deformation or collapse.
Consequently, since the leak-before-break methodology is
applicable to the SQN reactor coolant loop piping, the probability
of breaks in the primary loop piping is sufficiently low that they
need not be considered in the structural design of the plant. The
limiting LOCA event becomes either the accumulator line break or the
pressurizer surge line break. LOCA loads for the primary pipe breaks
were used to bound the conditions at SQN for smaller breaks. The
results of the analysis
[[Page 185]]
using the larger break inputs show that the LOCA loads were found to be
of insufficient magnitude to result in S/G tube collapse or
significant deformation. The LOCA, plus SSE tube collapse evaluation
performed for another plant with Series 51 S/Gs using bounding input
conditions (large-break loadings), is applicable to SQN. Therefore,
at SQN, no tubes will be excluded from using the voltage repair
criteria due to deformation of collapse of S/G tubes following a
LOCA plus an SSE. Additional supporting information relative to NRC
review of J.M. Farley Nuclear Plant was provided in Enclosure 5,
Item 3 of TVA's submittal dated September 7, 1995 (TAC No. M92961).
Addressing RG 1.83 considerations, implementation of the bobbin
probe voltage based interim tube plugging criteria of 2.0 volt is
supplemented by: (1) enhanced eddy current inspection guidelines to
provide consistency in voltage normalization, (2) a 100 percent eddy
current inspection sample size at the TSP elevations, and (3) RPC
inspection requirements for the larger indications left in service
to characterize the principal degradation as ODSCC.
As noted previously, implementation of the TSP elevation
plugging criteria will decrease the number of tubes that must be
repaired. The installation of S/G tube plugs reduces the RCS flow
margin. Thus, implementation of the alternate plugging criteria will
maintain the margin of flow that would otherwise be reduced in the
event of increased tube plugging.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: November 21, 1995
Brief description of amendments: The proposed amendments would
modify the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2
Technical Specifications (TS) to allow the containment personnel
airlock (PAL) doors to remain open during movement of irradiated fuel
and during core alterations.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change allows the PAL doors for containment to
remain open during the movement of irradiated fuel and core
alterations. Whether or not the PAL doors are open does not effect
the movement of fuel, the strict compliance with the procedures
governing refueling operations, or the integrity of fuel assemblies.
The position of the airlock doors cannot, in itself, be the
initiating event in any accident. The probability of a fuel handling
accident is not changed.
The consequences of leaving the airlock doors open during this
accident are bounded by the existing analysis, provided the fuel
handling accident assumptions are maintained (e.g. 100 hours after
reactor shutdown and the water level remains 23 feet above the
fuel). The existing analysis postulates the limiting fuel handling
accident to occur in the Fuel Building with no credit taken for
barrier or filtration. This accident analysis envelopes the proposed
change for a fuel handling accident occurring in the Containment
Building.
Were a fuel handling accident to occur with the PAL doors open,
the impact would be minimal. Pressure is expected to be essentially
equalized across the door with little air flow either into or out of
containment. Based on transport time from the location of the
accident to the PAL, little, if any, radioactive material is
expected to escape containment via the PAL. The amount that might
escape would not necessarily be anymore than might escape as the
door is cycled to evacuate personnel. What does escape will be
filtered by the Primary Plant Ventilation System, the same as if the
accident were to occur in the fuel building. In summary, not only is
the accident clearly bounded by the existing analysis, the actual
increase in release of radioactive material outside the plant will
be insignificant if there is any measurable increase at all.
Based on the above, allowing the PAL doors to remain open during
movement of irradiated fuel and core alterations, has no significant
effect on the probability or consequences of an accident previously
evaluated.
2. Do the proposed changes create the possibility of a new or
different type of accident from any accident previously evaluated?
The change does not add new hardware. The only change in the
operation of the plant is that the PAL doors will remain open during
movement of irradiated fuel and core alterations. Because the
current fuel handling accident analysis considers fuel handling
accidents in either the Fuel Building or the Containment Building,
the current fuel handling accident analysis remains bounding for the
proposed change. Therefore, the proposed change does not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The assumptions used to calculate the offsite dose resulting
from a fuel handling accident in [the] Containment Building are
equivalent to assuming that the PAL remains open for the entire
accident and that no filtration occurs. Since no credit was taken
for any containment barrier or ventilation system filtration, the
dose to the public as calculated in the analysis is not affected by
this change. 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, N.W., Washington, DC 20036
NRC Project Director: William D. Beckner
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: November 21, 1995
Brief description of amendments: The proposed amendment would
revise the core safety limit curves and revised N-16 Overtemperature
reactor trip setpoints as a result of the reload analyses for CPSES
Unit 2, Cycle 3. In addition, the minimum required Reactor Coolant
System (RCS) flow is increased and an administrative enhancement is
included in the footnotes of the RCS flow - low reactor trip function
setpoint 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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
A. Increase in Unit 2 minimum required flow
This revision increases the Unit 2 minimum required RCS flow
rate assumed in
[[Page 186]]
the safety analyses by 3.6%. The actual core flow is unchanged and is
approximately 6.6% higher than the value assumed in previous
accident analyses. The remaining 3.0% flow is sufficient to account
for all uncertainties associated with the core flow measurement.
Since this change only involves analysis methodology and does
not affect the actual core flow, it does not increase the actual
probability or consequences of any postulated accident.
When considered separately, increasing the minimum required RCS
flow is a conservative change. Although there is no impact on the
initiation of any postulated accidents, the potential severity of
the affected accidents is typically less when flow is increased. In
general, the increased ability to remove heat from the fuel will
reduce the peak temperature seen by the fuel and reduce the
potential for undesirable boiling conditions. Thus, the increase in
the assumed RCS flow will not increase the probability or
consequences of an accident previously analyzed.
B. Revision to the Unit 2 Core Safety Limits
Analyses of reactor core safety limits are required as part of
reload calculations for each cycle. TU Electric has performed in-
house analyses of the Unit 2, Cycle 3 core to determine the reactor
core safety limits. The newer methodologies and safety analysis
values result in new operating curves which, in general, permit
plant operation over a similar range of acceptable conditions. This
change means that if a transient were to occur with the plant
operating at the limits of the new curve, a higher temperature and
power level might be attained than if the plant were operating
within the bounds of the old curves. However, since the new curves
were developed using approved methodologies which are wholly
consistent with and do not represent a change in the Technical
Specification bases for safety limits, all applicable postulated
transients will continue to be properly mitigated. As a result,
there will be no significant increase in the consequences, as
determined by accident analyses, of any accident previously
evaluated.
C. Revision to Unit 2 Overtemperature N-16 Reactor Trip
Setpoints, Parameters and Coefficients
As a result of changes discussed, the Overtemperature N-16
reactor trip setpoint has been recalculated. These trip setpoints
help ensure that the core safety limits are maintained and that all
applicable limits of the safety analysis are met.
Based on the calculations performed, the safety analysis value
for Overtemperature N-16 reactor trip setpoint has changed. This
essentially means if a transient were to occur, the actual
temperature and power level could be slightly higher. However, the
analyses performed show that, using the TU Electric methodologies,
all reactor core safety limits are met and all applicable limits of
the safety analysis are met. This parameter has a setpoint which
allows the mitigation of postulated accidents and has no impact on
accident initiation. Therefore, the changes in safety analysis
values do not involve an increase in the probability of an accident
and, based on satisfying the core safety limits and all applicable
safety analysis limits, there is no significant increase in the
consequences of any accident previously evaluated.
In addition, the changes result in setpoint values which
potentially offer safety benefits. The risk of turbine runbacks or
reactor trips due to upper plenum flow anomalies will be minimized
with a higher overtemperature setpoint, thus reducing potential
challenges to the plant safety systems. A final benefit is that the
new methods for considering N-16 setpoints and values will be
consistent with Unit 1, which reduces the potential for personnel
error due to unit differences.
Considering both the safety analysis impact and the benefits
described above, the changes in N-16 setpoints and parameters will
result in slight reduction in the probability of an accident and do
not significantly increase the consequences of an accident
previously evaluated.
D. Deletion of footnotes associated with the RCS flow - low
reactor trip setpoint
In lieu of revising the footnotes to support the Unit 2 Cycle 3
operation, the deletion of the footnote is proposed. Further, for
consistency with Unit 2, the same change is proposed for Unit 1.
This change will not affect current plant practice; however, it will
impose a more restrictive RCS flow - low setpoint than is currently
required. The RCS flow - low reactor trip setpoint is currently
specified in Technical Specification Table 2.2-1, Functional Unit
12.b, to be 90% of the minimum measured RCS flow. The proposed
change would require the setpoint to be 90% of the instrument span
where 100% of instrument span approximately corresponds to the
actual RCS flow. The actual RCS flow is verified to be greater than
the RCS flow assumed in the accident analysis through compliance
with Technical Specification 3.2.5. Thus, through deletion of the
footnotes, the RCS volumetric flow corresponding to the reactor trip
setpoint will be greater than or equal to the volumetric flow
allowed by the current specifications.
In summary, the proposed deletion of the footnotes will have no
impact on current plant operations. A possible relaxation of the RCS
flow - low setpoint which is currently allowed by the Technical
Specifications will be removed without creating the potential for
unnecessary plant trips.
The RCS flow - low reactor trip setpoint can have no effect on
the probability of an accident. Because the reactor will be tripped
at or prior to the conditions assumed in the accident analyses,
there will be no effect on the consequences of an accident
previously identified.
SUMMARY
The changes in the amendment request applies new NRC approved
methodologies, changes in safety analysis values, new core safety
limits and new N-16 setpoint and parameter values to assure that all
applicable safety analysis limits have been met. The potential for
an operational transient to occur has been reduced and there has
been no significant impact on the consequences of any accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes involve the use of revised safety analysis
values and the calculation of new reactor core safety limits and
reactor trip setpoints. As such, the changes play an important role
in the analysis of postulated accidents but none of the changes
effect plant hardware or the operation of plant systems in a way
that could initiate an accident. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
In reviewing and approving the methods used for safety analyses
and calculations, the NRC has approved the safety analysis limits
which establish the margin of safety to be maintained. While the
actual impact on safety is discussed in response to question 1, the
impact on margin of safety is discussed below.
A. Increase in the Unit 2 minimum required flow
In performing the DNB-related analyses, the Reactor Coolant
System flow rate assumed in these analyses is increased by 3.6
percent to insure that all applicable limits of the safety analysis
are met. The Technical Specification 3/4.2.5 limit for this
parameter will be changed to insure that it is maintained within the
normal steady-state envelope of operation assumed in the transient
and accident safety analyses (i.e., ensuring that the RCS flow rate
assumed in the safety analyses remains valid). The Technical
Specification limits are consistent with the initial safety analysis
assumption (plus uncertainties) and have been analytically
demonstrated to be adequate to maintain a minimum DNBR at or above
the safety analysis DNBR limit throughout each analyzed transient.
Because the 95/95 DNBR acceptance criteria is met with the proposed
change and assumptions of the safety analyses are maintained valid
by the Technical Specification limits, there is no change in a
margin of safety.
B. Revision to the Unit 2 Reactor Core Safety Limits
The TU Electric reload analysis methods have been used to
determine new reactor core safety limits. All applicable safety
analysis limits have been met. The methods used are wholly
consistent with Technical Specification BASES 2.1 which is the bases
for the safety limits. In particular, the curves assure that for
Unit 2, Cycle 3, the calculated DNBR is no less than the safety
analysis limit and the average enthalpy at the vessel exit is less
than the enthalpy of saturated liquid.
In conjunction with the reactor core safety limit methodology,
the NRC approved TUE-1 DNB correlation is used for performing DNB-
related analyses. This correlation will be applied to the core
configuration of CPSES Unit 2, Cycle 3 and future core
configurations. The TUE-1 correlation DNBR limit is established such
that there is a 95 percent probability with 95 percent confidence
level that DNB will not occur when the minimum DNBR for the limiting
fuel is greater than or equal to the TUE-1 correlation DNBR limit.
This 95/95 criteria defines the ``margin of safety'' for the DNB-
[[Page 187]]
related analysis and remains valid even though the DNB correlation and
associated correlation limit are changed. Margin is provided in the
DNB-related analysis for known and potential effects such as
hydraulic differences between the two co-resident fuel assembly
designs and the presence of the Reactor Coolant System lower plenum
flow anomaly. The TUE-1 correlation DNBR limit plus margin
constitutes the safety analysis DNBR limit. The accident analyses
are performed to ensure that the safety analysis DNBR limit
acceptance criteria are satisfied. Because the 95/95 DNBR acceptance
criteria remains valid and continues to be satisfied, no change in a
margin of safety occurs.
C. Revision to Unit 2 Overtemperature N-16 Reactor Trip
Setpoints, Parameters and Coefficients
Because the reactor core safety limits for CPSES Unit 2, Cycle 3
are recalculated, the Reactor Trip System instrumentation setpoint
values for the Overtemperature N-16 reactor trip setpoint which
protect the reactor core safety limits must also be recalculated.
The Overtemperature N-16 reactor trip setpoint helps prevent the
core and Reactor Coolant System from exceeding their safety limits
during normal operation and design basis anticipated operational
occurrences. The most relevant design basis analysis in Chapter 15
of the CPSES Final Safety Analysis Report (FSAR) which is affected
by the change in the safety analysis value for the CPSES Unit 2
Overtemperature N-16 reactor trip setpoint is the Uncontrolled Rod
Cluster Control Assembly Bank Withdrawal at Power (FSAR Section
15.4.2). This event has been re-analyzed with the revised safety
analysis value for the Overtemperature N-16 reactor trip setpoint to
demonstrate compliance with event specific acceptance criteria.
Because all event acceptance criteria are satisfied, there is no
degradation in a margin of safety.
The nominal Reactor Trip System instrumentation setpoints values
for the Overtemperature N-16 reactor trip setpoint (Technical
Specification Table 2.2-1) are determined based on a statistical
combination of all of the uncertainties in the channels to arrive at
a total uncertainty. The total uncertainty plus additional margin is
applied in a conservative direction to the safety analysis trip
setpoint value to arrive at the nominal and allowable values
presented in Technical Specification Table 2.2-1. Meeting the
requirements of Technical Specification Table 2.2-1 assures that the
Overtemperature N-16 reactor trip setpoint assumed in the safety
analyses remains valid. The CPSES Unit 2, Cycle 3 Overtemperature N-
16 reactor trip setpoint is different from previous cycles which
provides more operational flexibility to withstand mild transients
without initiating automatic protective actions. Although the
setpoint is different, the Reactor Trip System instrumentation
setpoint values for the Overtemperature N-16 reactor trip setpoint
are consistent with the safety analysis assumption which has been
analytically demonstrated to be adequate to meet the applicable
event acceptance criteria. Thus, there is no reduction in a margin
of safety.
D. Deletion of footnotes associated with the RCS flow - low
reactor trip function
The deletion of the footnotes, and the potential relaxation of
the RCS flow - low setpoint which could be used, will provide
further assurance that, in the event of a partial loss of forced RCS
flow or locked rotor transient, a reactor trip signal would be
initiated prior to the conditions assumed in the accident analyses.
Thus, the accident analyses are unaffected, and there is no
reduction in a margin of safety.
SUMMARY
The proposed changes to the CPSES Technical Specifications
involve using NRC-approved licensing analysis methods developed by
TU Electric to determine the Technical Specification reactor core
safety limits and perform DNB-related analysis for CPSES Unit 2,
Cycle 3. The DNB-related analyses are performed by TU Electric using
a qualified, state-of-the-art departure from nucleate boiling (DNB)
correlation, TUE-1, which has also been approved by the NRC for the
CPSES Unit 2, Cycle 3 core configuration. In performing these
analyses, the minimum required Reactor Coolant System flow rate is
increased by 3.6 percent. Because the core safety limits for CPSES
Unit 2, Cycle 3 are recalculated, the Reactor Trip System
instrumentation setpoints values for the Overtemperature N-16
reactor trip setpoint which protect the core safety limits are also
recalculated.
Using the NRC approved TU Electric methods, the reactor core
safety limits are determined such that all applicable limits of the
safety analyses are met, particularly the 95/95 DNBR limit. The
Technical Specification 3/4.2.5 limits for the DNB Parameters insure
the assumptions in the safety analyses remain valid. Because the
applicable event acceptance criteriacontinue to be met, there is no
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, N.W., Washington, DC 20036
NRC Project Director: William D. Beckner
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: October 17, 1995
Description of amendment request: The proposed amendment would
modify the North Anna Power Station, Units 1 and 2 Technical
Specifications (TS) to allow both of the containment personnel airlock
doors to remain open during refueling operations, delete the license
condition referencing the analyses for limiting doses to the control
room operators, and modify the TS Bases to clarify the emergency power
system requirements relative to mitigation of the consequences of a
Fuel Handling Accident.
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:
There is no significant change in the probability or
consequences of an accident previously evaluated. There are no
system changes which would increase the probability of an accident
occurring. Allowing both personnel airlock doors to remain open
during core alterations or fuel movement inside containment will not
have any impact on the probability of a Fuel Handling Accident
either in containment or in the fuel building. The consequences of a
Fuel Handling Accident have been investigated by performing a
reanalysis with no credit for isolation or filtration by the Fuel
Building or containment ventilation systems. The Exclusion Area
Boundary [EAB] and Low Population Zone [LPZ] doses for a Fuel
Handling Accident without credit for iodine filtration remain well
within (<25%) of the NRC regulatory limits of 10 CFR [Part] 100. The
predicted control room operator doses remain bounded by the limiting
case for control room doses and within the regulatory limits of
General Design Criterion [GDC] 19. In addition, the action to
clarify the responses to NRC question 6.72 [of the original Final
Safety Analysis Report] will not increase the probability or
consequences of the Fuel Handling Accident.
No new accident types or equipment malfunction scenarios are
introduced as a result of the clarification to the Virginia Power
response to [NRC question] 6.72 or as a result of these changes in
analysis methods or the proposed Technical Specifications changes to
allow both personnel airlock doors to remain open during core
alterations or fuel movement inside containment. Therefore, there is
no possibility of an accident of a different type than any
previously evaluated in the North Anna USFAR [Updated Final Safety
Analysis Report].
There is no significnt reduction in the margin of safety. An
evaluation of the Fuel Handling Accident doses at the EAB, the LPZ
and to control room operators has been performedand it has been
concluded that the acceptance criteria defined by GDC-19, 10 CFR
100, and the NRC Standard Review Plan will continue to be met.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request
[[Page 188]]
involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
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 23212.
NRC Project Director: David B. Matthews
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of application for amendments: September 13, 1995, as amended
November 27, 1995
Brief description of amendments: The proposed amendments would
permit the licensee to implement the performance-based option provided
by 10 CFR Part 50, Appendix J, which allows leakage testing intervals
to be based on system and component testing performance.
Date of publication of individual notice in Federal Register:
December 12, 1995 (60 FR 63739)
Expiration date of individual notice: January 11, 1996
Local Public Document Room location: The University of North
Carolina at Wilmington, William Madison Randall Library, 601 S. College
Road, Wilmington, North Carolina 28403-3297
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-529 and
STN 50-530, Palo Verde Nuclear Generating Station, Units 2 and 3,
Maricopa County, Arizona
Date of application for amendments: October 3, 1995
Brief description of amendments: The amendments delete Sections
2.B.(7)(a) and (b) of
Facility Operating License No. NPF-51 (Unit 2) and Sections
2.b.(6)(a) and (b) of
Facility Operating License No. NPF-74 (Unit 3) relating to certain
previous sale and leaseback transactions that were added by Amendment
No. 3 for NPF-51 and Amendment No. 1 for NPF-74.
Date of issuance: December 8, 1995
Effective date: December 8, 1995
Amendment Nos.: Unit 2 - Amendment No. 91; Unit 3 - Amendment No.
74
Facility Operating License Nos. NPF-51 and NPF-74: The amendments
revised the license.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56363) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 8, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of application for amendments: October 23, 1995
Brief description of amendments: The amendments revised the
Technical Specifications to delete the applicability of the primary
coolant water chemistry limits when the primary system is being
chemically decontaminated and the reactor vessel is defueled.
Date of issuance: December 13, 1995
Effective date: December 13, 1995
Amendment Nos.: 180 and 211
Facility Operating License Nos. DPR-71 and DPR-62.
Date of initial notice in Federal Register: The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated December 13, 1995. No significant hazards
consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: September 14, 1995, as
supplemented November 8, 1995.
Brief description of amendments: The amendments allow the use of an
alternate zirconium based fuel cladding,
[[Page 189]]
ZIRLO, and permit limited substitution of fuel rods with ZIRLO filler
rods. In addition, a clarification and an editorial change have been
included.
Date of issuance: December 19, 1995
Effective date: December 19, 1995
Amendment Nos.: 78 and 70
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 25, 1995 (60 FR
54716) The November 8, 1995 letter, provided clarifying information
that did not change the scope of the September 14, 1995, application
and the initial proposed no significant hazards consideration
determination.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 19, 1995No significant
hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: September 15, 1995.
Brief description of amendments: The amendments upgrade the current
custom Technical Specifications (TS) for Dresden and Quad Cities to the
Standard Technical Specifications contained in NUREG-0123, ``Standard
Technical Specification General Electric Plants BWR/4.'' The
application dated September 15, 1995, contains some of the TSUP open
items from previous Dresden and Quad Cities TS amendments issued by the
NRC.
Date of issuance: December 19, 1995Effective date: Immediately, to
be implemented no later than June 30, 1996.
Amendment Nos.: 145, 139, 167 and 163
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 5, 1995 (60 FR
52220) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 19, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: December 8, 1995
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.1.A.5 to revise the wording to allow a single
train of Power-Operated Relief Valves (PORVs)/Block Valves to be closed
and deenergized indefinitely. The proposed change is administrative and
is intended to correct inconsistencies between the intended operation
of the PORVs/Block Valves and the language of the TSs.
Date of issuance: December 8, 1995
Effective date: As of the date of issuance to be implemented
immediately.
Amendment No.: 185
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: NoThe Commission's related
evaluation of the amendment, emergency circumstances and consultation
with the State, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated December 8,
1995.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Ledyard B. Marsh
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: October 31, 1994
Brief description of amendments: The amendments remove the stroke
times for the steam generator power operated relief valves from
Technical Specification Tables 3.6-2a and 3.6-2b.
Date of issuance: December 18, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: Unit 1 - 139 - Unit 2 - 133
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8745) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 18, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas
Date of amendment request: September 4, 1993, as supplemented by
letters dated February 16, 1994, and August 4, 1995
Brief description of amendments: The license amendments revised the
Arkansas Nuclear One Industrial Security Plan.
Date of issuance: December 19, 1995
Effective date: December 19, 1995
Amendment Nos.: 183 and 172
Facility Operating License Nos. DPR-51 and NPF-6. Amendments
revised the licenses.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56368) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 19, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 6, 1993, as supplemented by
letters dated May 12, August 9, and September 18, 1995.
Brief description of amendment: The amendment changes the Appendix
A TSs to allow installation of steam generator tube repair sleeves at
the Waterford Steam Electric Station, Unit 3. The sleeves are designed
and manufactured by Combustion Engineering Incorporated.
Date of issuance: December 14, 1995
Effective date: December 14, 1995, to be implemented within 60 days
Amendment No.: 117
[[Page 190]]
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2868) The May 12, August 9, and September 18, 1995, letters provided
additional 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 December 14, 1995.No significant hazards consideration comments
received: No
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 7, 1993, as supplemented by
letters dated February 8, 1994, and August 9, 1995.
Brief description of amendment: The amendment revised the license
condition on physical security and approves the revision to Physical
Security Plan for the Waterford Steam Electric Station, Unit 3.
Date of issuance: December 19, 1995
Effective date: December 19, 1995
Amendment No.: 118
Facility Operating License No. NPF-38. Amendment revised the
license. The additional information contained in the supplemented
letter dated August 9, 1995, was clarifying in nature and thus, within
the scope of the initial notice and did not affect the staff's proposed
no significant hazards consideration determination.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14887) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 19, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: July 26, 1995
Brief description of amendments: The amendments consist of changes
to the Technical Specifications relating to nuclear instrumentation
system adjustments based on calorimetric measurements at reduced power
levels.Date of issuance: December 12, 1995
Effective date: December 12, 1995
Amendment Nos. 180 and 174Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47617) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 12, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: January 26, 1995, as
supplemented March 9 and May 24, 1995
Brief description of amendment: This amendment increases the
allowable U-235 enrichment of fuel to be stored in the new and spent
fuel storage facilities.
Date of issuance: December 15, 1995
Effective date: December 15, 1995
Amendment No.: 151
Facility Operating License No. DPR-72. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20517) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 15, 1995. No
significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 24, 1995
Brief description of amendment: The amendment revised the Technical
Specifications to reflect the approval for the River Bend Station to
use 10 CFR Part 50, Appendix J, Option B for the containment leak rate
testing.
Date of issuance: December 19, 1995
Effective date: December 19, 1995
Amendment No.: 84
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56368) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 19, 1995.No significant
hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: May 25, 1995 (AEP:NRC:1200B)
Brief description of amendments: The amendments change the
surveillance frequency for the manual actuation function for main steam
line isolation from monthly to quarterly and delete obsolete footnotes
associated with previous surveillance interval extensions from Unit 2
Table 4.3-2.
Date of issuance: December 13, 1995
Effective date: December 13, 1995, with full implementation within
45 days
Amendment Nos.: 204 and 189
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35081)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 13, 1995. No
significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: August 8, 1995
Brief description of amendment: This amendment modifies the
definitions of Transthermal (Condition 4), Hot Shutdown (Condition 5),
and Hot Standby (Condition 6) reactor operating conditions. The
Transthermal and Hot Shutdown Conditions are modified to establish an
applicable range of subcriticality and be consistent with other
Definitions. The wording of Hot Standby is modified to remove reference
to control rod position, consistent with NUREG-1432, Standard Technical
Specifications for Combustion Engineering Plants, Revision 1, dated
April 1995.
[[Page 191]]
Date of issuance: December 15, 1995
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 154
Facility Operating License No. DPR-36: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52931) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 15, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: June 27, 1995
Brief description of amendment: The amendment revises Technical
Specification (TS) 2.2 on chemical and volume control system (CVCS) to
reformat and clarify the requirements and make them more consistent
with the requirements of the Combustion Engineering Standard Technical
Specifications (STS), as presented in NUREG-0212, Revision 2.
Date of issuance: December 12, 1995
Effective date: December 12, 1995
Amendment No.: 171
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39447) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 12, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: February 10, 1995, as
supplemented by letter dated November 10, 1995.
Brief description of amendments: These amendments (1) modify the
Susquehanna Steam Electric Station, Unit 1 and 2 Technical
Specifications to extend the allowable out-of-service times (AOTs) for
maintenance and repair and the surveillance test intervals (STIs)
between channel functional tests for the following groups of
instruments: reactor protection systems instrumentation (TS 3.3.1),
isolation actuation instrumentation (TS 3.3.2), emergency core cooling
system actuation instrumentation (TS 3.3.3), ATWS (anticipated
transient without scram) recirculation pump trip system instrumentation
(TS 3.3.4.1), end-of-cycle recirculation pump trip system
instrumentation (TS 3.3.4.2), reactor core isolation cooling system
(RCIC) actuation instrumentation (TS 3.3.5), control rod block
instrumentation (TS 3.3.6), radiation monitoring instrumentation (TS
3.3.7.1), and feedwater/main turbine trip system actuation
instrumentation (TS 3.3.90); (2) change the required actions and AOTs
for the instruments listed above to make requirements consistent with
supporting analysis in General Electric topical reports and change
additional actions required to prevent extended AOTs from resulting in
extended loss of instrument function; (3) change the required actions
and AOTs for the instruments listed above for instrumentation
associated with the ADS (automatic depressurization system),
recirculation pump trip, and pump suction lineup for HPCI (high
pressure core injection) and RCIC; (4) change applicability
requirements and required actions for the reactor vessel water level-
low, level 3 function that isolates the RHR (residual heat removal)
system shutdown cooling system so that the function is required to be
operable in operational conditions 3,4, and 5 to prevent inadvertent
loss of reactor coolant via the RHR shutdown cooling system; (5) remove
notes in Table 3.3.2-1, 3.3.2-2, and 4.3.1-1 related to maintenance on
leak detection temperature detectors and remove the note toTS 3.3.6 for
Unit 1 related to a previous relief from TS 3.0.4; and (6) reformat,
renumber, and/or reword existing requirements to incorporate the
changes listed above.
Date of issuance: December 18, 1995
Effective date: As of date of issuance and to be implemented within
30 days.
Amendment Nos.: 155 and 126
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16194) The supplemental letter provided corrected TSs and did not
change the original proposed no significant hazards consideration nor
the Federal Register notice.The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated December 18,
1995No significant hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
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: September 26, 1995
Brief description of amendments: The amendments change the
containment air lock door seal leakage rate from ``no detectable seal
leakage'' to ``less than or equal to 0.01 La'' when the gap
between the door seals is pressurized to greater than or equal to 10
psig for a period of not less than 15 minutes.
Date of issuance: December 8, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 118 and 109
Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56370) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 8, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
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: August 7, 1995 (TS 95-17)
Brief description of amendments: The changes relocate the heat flux
hot channel factor penalty from Surveillance Requirement 4.2.2.2.e.1 to
the Core Operating Limits Report and replace the methodology (WCAP-
10216-P-A) listed in Technical Specification 6.9.1.14.a.2 with WCAP-
10216-P-A, Revision 1A.
Date of issuance: December 11, 1995
Effective date: December 11, 1995
Amendment Nos.: 216 and 206
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45186) The Commission's related evaluation of the amendment is
contained in a Safety
[[Page 192]]
Evaluation dated December 11, 1995.No significant hazards consideration
comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402No significant
hazards consideration comments received: None
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: August 15, 1995 (TXX-95215)
Brief description of amendments: These changes relocated the
Shutdown Margin limits from the Technical Specifications (TSs) to the
Core Operating Limits Report (COLR). The changes were consistent with
the intent of Generic Letter 88-16 which provides guidelines for the
removal of cycle-specific parameter limits from the TSs.
Date of issuance: December 15, 1995
Effective date: December 15, 1995
Amendment Nos.: Unit 1 - Amendment No. 44; Unit 2 - Amendment No.
30
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52935) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 15, 1995.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, Callaway
County, Missouri
Date of amendment request: April 26, 1995
Brief description of amendment: The amendment revises Technical
Specification (TS) 3/4.7.6 to reduce the upper limit on the flow rate
through the control room filtration subsystem and adopts ASTM D-3803-
1989 as the laboratory testing standard for control room filtration and
control building pressurization charcoal adsorber. The amendment also
revises the Bases for TS 3/4.7.6 to reflect the changes.
Date of issuance: December 20, 1995
Effective date: December 20, 1995, to be implemented within 30 days
from the date of issuance.
Amendment No.: 106
Facility Operating License No. NPF-30. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27345) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 20, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: June 14, 1995, as supplemented by
letters dated July 13, 1995, and August 22, 1995.I11Brief description
of amendment: The amendment revises Technical Specification (TS) 3.2.3,
``Nuclear Enthalpy Rise Hot Channel Factor,'' TS 6.9.1.9, ``Core
Operating Limits Report,'' and the associated Bases sections. The
revisions incorporate changes associated with the planned
implementation of advanced nuclear and core thermal-hydraulic design
methodologies licensed from Westinghouse Electric Corporation for core
reload design, starting with Cycle 9.
Date of issuance: December 8, 1995
Effective date: December 8, 1995, to be implemented prior to
restart from the eighth refueling outage, which is scheduled to begin
in March 1996.
Amendment No.: 92
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39456) The August 22, 1995, supplemental letter forwarded the
nonproprietary version of Wolf Creek Nuclear Operating Corporation's
safety evaluation and analysis provided in the June 14, 1995, submittal
and did not change the staff's original no significant hazards
determination.The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 8, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: August 22, 1995
Brief description of amendment: The amendment revises the
requirements of Technical Specification (TS) 3.3.1 and TS 3.3.2 and
relocate Tables 3.3-2 and 3.3-5 and applicable Bases, which provide the
response time limits for the reactor trip system (RTS) and the
engineered safety features actuation system (ESFAS) instruments, from
the TS to the Updated Safety Analysis Report (USAR). The licensee has
stated that the next USAR change request will include these changes.
Date of issuance: December 12, 1995
Effective date: December 12, 1995, to be implemented within 60 days
of issuance.
Amendment No.: 93
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49950) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 12, 1995.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
Dated at Rockville, Maryland, this 21st Day of December 1995.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor
Regulation
[Doc. 96-1 Filed 1-2-96; 8:45 am]
BILLING CODE 7590-O1-F