# Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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URL: https://www.frixlaw.com/law-library/documents/fr%3A97-11910

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** July 30, 1997
- **Citation:** 62 FR 40843

## Text

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.

[[Page 40844]]

This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 3, 1997, through July 18, 1997. The
last biweekly notice was published on July 16, 1997.

Notice of Consideration of Issuance of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By August 29, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with

[[Page 40845]]

the Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Docketing and Services Branch, or
may be delivered to the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington DC, by the above date. A copy
of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendments request: May 23, 1997
Description of amendments request: The proposed amendment would
revise Technical Specification 3/4.4.4 to allow the installation of
ABB/CE welded sleeves, in accordance with ABB/CE Topical Report CEN-
630-P, ``Repair of 3/4 Inch Outer Diameter Steam Generator Tubes Using
Leak Tight Sleeves,'' Revision 1, in the Palo Verde Units 1, 2 and 3
steam generators.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below: 1. The proposed change does
not involve a significant increase in the probability or consequences
of an accident previously evaluated.
The proposed amendment to permit the use of steam generator tube
sleeves as an alternative to tube plugging is a safe and effective
repair procedure that does not result in removing a tube from
service. Mechanical strength, corrosion resistance, installation
methods, and inservice inspection techniques of sleeves have been
shown to meet NRC acceptance criteria.
Analytical verifications were performed using design and
operating transient parameters selected to envelope loads imposed
during normal operating and accident conditions. Fatigue and stress
analysis of sleeved tube assemblies were completed in accordance
with the requirements of Section III of the ASME Code. The results
of qualification testing, analysis and plant operating experience at
other facilities demonstrates that the sleeving process is an
acceptable means of maintaining steam generator tube integrity. The
sleeve configuration has been designed and analyzed in accordance
with the structural margins specified in Regulatory Guide 1.121 (RG
1.121). Furthermore, the installed sleeve will be monitored through
periodic inspections on a sample basis with eddy current techniques.
A sleeve-specific plugging margin, per the recommendations of
Regulatory Guide 1.121, has been specified with appropriate
allowances for NDE uncertainty and defect growth rate. Therefore,
since the sleeve provides the same protection against a tube rupture
as the original tube, the use of sleeves does not involve a
significant increase in the probability of an accident previously
evaluated.
Recently, industry experience with forced shutdown events
associated with tube failures at sleeve junctions was assessed by
APS and ABB-CE. The root cause of these events has been attributed
to the lack of proper post-installation stress relief and/or the
imposition of high stresses due to tube growth restrictions at
locked tube supports. The material and design of the PVNGS steam
generator supports minimizes the potential for locked supports. The
tube supports are of eggcrate design and are constructed of ferric
stainless steel. The large flow area in the eggcrate design provides
better irrigation and reduces the potential for steam blanketing,
therefore, the tube-to-tube support crevices are less likely to be
blocked by crud, boiler water deposits and corrosion products. Since
the support material is type 409 ferric stainless steel, it is not
susceptible to magnetite corrosion which has resulted in denting and
lockup at plants with carbon steel supports. These conclusions have
been substantiated via tube pull activities conducted in PVNGS Unit
2. Although ABB/CE does not require post-weld heat treatment in all
applications, APS will require that a post-weld stress relief be
conducted for sleeve installations. Therefore, with proper sleeve
installation the proposed change will not involve a significant
increase in the probability of an accident previously evaluated.
The consequences of accidents previously analyzed are not
increased as a result of sleeving activities. The hypothetical
failure of the sleeve would be bounded by the current steam
generator tube rupture analysis contained in the PVNGS UFSAR. Due to
the slight reduction in diameter caused by the sleeve wall
thickness, it is expected that the primary release rates would be
less than assumed for the steam generator tube rupture analysis,
and, therefore, would result in lower primary fluid mass release to
the secondary system. Additionally, further conservatism is
introduced if the break were postulated to occur at a location on
the tube higher than the location where a sleeve is installed. The
overall effect would be reduced steam generator tube rupture release
rates. The minimal reduction in flow area associated with a tube
sleeve has no significant affect on steam generator performance with
respect to heat transfer or system flow resistance and pressure
drop. The installation of sleeves rather than plugging also
maintains a greater heat transfer surface in the steam generator. In
any case, the impacts are bounded by evaluations which demonstrate
the acceptability of tube plugging, which totally removes the tube
from service.
Therefore, in comparison to plugging, tube sleeving is
considered a significant improvement with respect to steam generator
performance. Therefore, based on the above, the proposed amendment
does not significantly increase the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
A sleeved steam generator tube performs the same function in the
same passive manner as an unsleeved steam generator tube. Tube
sleeves are designed and qualified to the stress and pressure limits
of Section III of the ASME Code and Regulatory Guide 1.121.
The installation of the sleeve, including weld and welder
qualification and nondestructive examination (NDE), meets or exceeds
the requirements of ASME Section XI. Three types of NDE are
conducted. Ultrasonic Testing (UT) is performed to verify the
adequacy of the tube to sleeve weld assuring proper fusion. Eddy
Current testing (ECT) is performed following each installation to
establish baseline data for each sleeve in order to monitor future
degradation of the primary to secondary pressure boundary. Visual
inspections will be performed to verify or ascertain the mechanical
and structural condition of a weld. Critical conditions which are
checked include weld width and completeness, and the absence of
visibly noticeable indications such as cracks, pits, and burn
through.
ABB Combustion Engineering, Inc., Report CEN-630-P, Revision 01,
``Repair of 3/4'' O.D. Steam Generator Tubes Using Leak Tight
Sleeves'' dated November, 1996, demonstrates that the repair of
degraded steam generator tubes using tube sleeves will result in
tube bundle integrity consistent with the original design basis.
Extensive analyses and testing have been performed on the sleeve and
sleeve to tube joints to demonstrate that the design criteria are
met. The proposed amendments have no significant effect on the
configuration of the plant, and the change does not affect the way
in which the plant is operated. Therefore, reactor operation with
sleeves installed in the steam generator tubes does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.

[[Page 40846]]

Evaluation of the sleeved tubes indicates no detrimental effects
on the sleeve-tube assembly resulting from reactor coolant system
flow, coolant chemistries, or thermal and pressure conditions.
Structural analyses have been performed for sleeves which span the
tube at the top of the tube sheet and which span the flow
distribution plate or eggcrate support. Mechanical testing has been
performed to support the analyses. Corrosion testing of typical
sleeve-tube assemblies has been completed and reveals no evidence of
sleeve or tube corrosion considered detrimental under anticipated
service conditions.
Steam generator tube integrity is maintained under the same
limits for sleeved tubes as for unsleeved tubes, ie., Section III of
the ASME Code and Regulatory Guide 1.121. The portions of the
installed sleeve assembly which represents the reactor coolant
pressure boundary can be monitored for the initiation and
progression of sleeve/tube wall degradation, thus satisfying the
requirements of Regulatory Guide 1.83. The degradation limit at
which a sleeve/tube boundary is considered inoperable has been
analyzed in accordance with Regulatory Guide 1.121 and is specified
in the proposed amendment. Eddy current detectability of flaws has
been verified by ABB Combustion Engineering. Additionally, the
Technical Specifications continue to require monitoring and
restriction of primary- to- secondary system leakage through the
steam generators. The minimal reduction in RCS flow due to sleeving
results in an insignificant impact on RCS operation during normal or
accident conditions and is bounded by tube plugging evaluations.
Based upon the testing and analyses performed, the installation
of tube sleeves will not result in a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: William H. Bateman

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 amendments request: 3April 30, 1997
Description of amendments request: The proposed amendments would
revise Surveillance Requirements (SRs) 4.7.2.b.2 and 4.7.2.c in the
Technical Specifications for the Brunswick Steam Electric Plant, Units
1 and 2. These SRs require periodic testing of the control room
emergency ventilation system charcoal filters. The proposed amendments
would revise the temperature and relative humidity conditions under
which the testing is performed. The revised conditions were selected to
approximate operating or accident conditions. Testing at the revised
conditions is more conservative than testing at the currently required
conditions. Additionally, the proposed amendments would relax the
acceptance criterion for filtration efficiency from 95% to a value
corresponding to a filtration efficiency of 90%. The 90% value is the
filtration efficiency assumed in the current bounding calculations for
control room dose under accident conditions.
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 amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendments revise Surveillance Requirements
4.7.2.b.2 and 4.7.2.c to require testing of the control room
emergency ventilation system (CREVS) charcoal in accordance with
ASTM D3803-1989, ``Standard Test Method for Nuclear-Grade Activated
Carbon.'' Currently, Surveillance Requirements 4.7.2.b.2 and 4.7.2.c
to [sic] require testing in accordance with the criteria of
Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1,
1976. The purpose of the CREVS is to mitigate an accident. It is not
associated with any initiating events and, therefore, cannot affect
the probability of any accident.
ASTM D3803-1989 is an industry accepted standard for charcoal
filter testing. The conditions employed by this standard were
selected to approximate operating or accident conditions of a
nuclear reactor which would severely reduce the performance of
activated carbons. The ASTM D3803-1989 testing is more stringent
than that required by the criteria of Regulatory Position C.6.a of
Regulatory Guide 1.52, Revision 1, 1976. Specifically, the testing
temperature of ASTM D3803-1989 is 30.0 [plus or minus] 0.2 deg.C
versus 80 deg.C for the Regulatory Guide 1.52 testing. Also, ASTM
D3803-1989 requires a relative humidity of 93 to 96% versus [greater
than or equal to] 70% for the Regulatory Guide 1.52 testing. Both
these parameters result in the ASTM D3803-1989 test being a more
conservative test [than] that required by the criteria of Regulatory
Position C.6.a of Regulatory Guide 1.52, Revision 1, 1976.
The proposed changes to Surveillance Requirements 4.7.2.b.2 and
4.7.2.c require that charcoal samples tested in accordance with the
methodology of ASTM D3803-1989 meet the acceptance criteria of , instead of Note
is an administrative change and thus will not
change the probability or consequences of an accident.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The changes to the definition of Channel Calibration and
correction of the other miscellaneous errors in the TS and TS Bases
will not create the possibility of a new or different kind of
accident, because the changes will not affect the design or
operation of any structure, system, or component in the plant.
3) Involve a significant reduction in the margin of safety
because:
a. The definition of Channel Calibration is being changed to be
like the definition in NUREG 1434, Standard Technical Specifications
General Electric Plants, BWR/6, Revision 1. The primary changes
involve requiring only an inplace qualitative assessment of
thermocouple and RTD sensors. These sensors are not adjustable and
not susceptible to setpoint drift. Thus the appropriate check of the
sensors is a qualitative assessment only. The inplace qualitative
assessment assures operability of the sensors. Therefore there is no
reduction in the margin of safety.
b. The remaining miscellaneous changes are corrections due to
errors in the TS. The corrections will make the associated TS
consistent with the design and licensing basis of LaSalle or correct
typographical errors. Therefore, there is no reduction in the margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra

Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 17, 1996, as supplemented by
letters dated June 3, and July 7, 1997.
Description of amendment request: The proposed change request
modifies Waterford Steam Electric Station, Unit 3, Technical
Specifications (TSs) 3/4.7.1.3, ``CONDENSATE STORAGE POOL,'' by
increasing the minimum Condensate Storage Pool (CSP) level from 82
percent to 91 percent in Modes 1, 2, and 3. The July 7, 1997,
supplement proposes to expand the applicability of TS 3.7.1.3 to
include Mode 4 operational requirements and maintains the 91 percent
minimum CSP level previously requested for Modes 1, 2, and 3. The staff
previously issued No Significant Hazard Considerations notice on March
26, 1997 (62 FR 14461).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?Response: No.
Increasing the minimum required Condensate Storage Pool (CSP)
level to 91 percent will insure that the minimum required 170,000
gallons of water is available to supply the Emergency Feedwater
System and that 3,500 gallons of water is available for use by the
Component Cooling Water Makeup System in Modes 1, 2, and 3.
Maintaining a minimum required CSP level of 11 percent will insure
that 3,500 gallons of water is available for use by the Component
Cooling Water Makeup System in Mode 4. Maintaining the minimum
required water volume will not increase the probability of any
accident previously evaluated. Additionally, it will not affect the
consequences of any accident. Maintaining a minimum required CSP
level will ensure that the system remains within the bounds of the
accident analysis. Therefore, the proposed change will not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No.
Increasing the minimum water volume of the CSP from 82 percent
to 91 percent in Modes 1, 2, and 3 does not create a possibility for
a new or different kind of accident. Maintaining a minimum water
volume of the CSP at 11 percent in Mode 4 does not create a
possibility for a new or different kind of accident. The CSP will be
operated in the same manner as previously evaluated. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
Operation in accordance with this proposed change will ensure
that the minimum contained water volume of the CSP will remain
adequate under all conditions. This will improve the present 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, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: James W. Clifford, Acting Director

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: May 29, 1997
Description of amendment request: The proposed amendments will
improve consistency throughout the Technical Specifications and their
related Bases by removing outdated material, incorporating minor
changes in text, making editorial corrections, and resolving other
inconsistencies identified by the licensee's plant operations staff.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments consist of administrative changes to the
Technical Specifications (TS) for St. Lucie Units 1 and 2. The
amendments will implement minor changes in text to rectify
reference, typographic, spelling, and/or consistency-in-format
errors; update the TS Bases; and/or otherwise improve consistency
within the TS for each unit. The proposed amendments do not involve
changes to the configuration or method of operation of
plantequipment that is used to mitigate the consequences of an
accident, nor do the changes otherwise affect the initial conditions
or conservatisms assumed in any of the plant accident analyses.
Therefore, operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.

[[Page 40850]]

The proposed administrative revisions will not change the
physical plant or the modes of plant operation defined in the
Facility License for each unit. 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 amendments would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendments are administrative in nature and do not
change the basis for any technical specification that is related to
the establishment of, or the preservation of, a nuclear safety
margin. Therefore, operation of the facility in accordance with the
proposed amendments would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420
NRC Project Director: Frederick J. Hebdon

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island
Nuclear Station, Unit No. 2 (TMI-2), Dauphin County, Pennsylvania

Date of amendment request: December 2, 1996
Description of amendment request: The proposed amendment would
relocate the audit frequency requirements from the plant Technical
Specifications to the Quality Assurance Plan. In addition, the maximum
interval between certain types of audits will be extended. This change
would make the TMI-2 technical specifications consistent with the
Technical Specifications for Three Mile Island, Unit 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
10 CFR 50.92 provides the criteria which the Commission uses to
perform a No Significant Hazards Consideration. 10 CFR 50.92 states
that an amendment to a facility license involves No Significant
Hazards if 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 proposed change to the technical specifications is
administrative and does not involve any physical changes to the
facility. No changes are made to operating limits or parameters, nor
to any surveillance activities. Based on this, GPU Nuclear has
concluded that the proposed change does not:
1. Involve a significant increase in the probability of
occurrence of the consequences of an accident previously evaluated.
The proposed amendment is administrative and does not affect the
function of any system or component. Therefore this change does not
increase the probability of occurrence or the consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change is administrative and no new failure modes
or potential accident scenarios are created.
3. Involve a change in the margin of safety.
This change is administrative in nature and does not affect any
safety settings, equipment, or operational parameters.
Based on the above analysis it is concluded that the proposed
changes involve no significant safety hazards considerations as
defined by 10 CFR 50.92.
The NRC staff has reviewed the analysis of the licensee and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW, Washington, D.C. 20037
NRC Project Acting Director: Marvin M. Mendonca

Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas

Date of amendment request: July 16, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification Table 2.2-1 and 3/4.2.5 to allow the
reactor coolant system total flow to be determined using cold leg elbow
tap differential pressure measurements.
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:
Pursuant to 10[]CFR[]50.92 each application for amendment to an
operating license must be reviewed to determine if the proposed
change involves a Significant Hazards Consideration. The amendment,
as defined below, describing the Technical Specification change
associated with the change has been reviewed and determined to not
involve Significant Hazards Considerations. The basis for this
determination follows.
Proposed Change: The current Technical Specification Table 2.2-1
(page 2-4) ``Reactor Trip System Instrumentation Trip Setpoints,''
provides the Trip Setpoint and Allowable Value for the RCS [reactor
coolant system] Flow-Low trip. The Allowable Value will be changed
to reflect the increased uncertainty associated with the correlation
of the elbow taps to a previous baseline calorimetric. In addition,
Technical Specification 3.2.5 (page 3/4.2-11), ``Power Distribution
Limits, DNB Parameters'', will be changed to allow the RCS total
flow to be measured by the elbow tap [delta]p method. These changes
will include the modification of surveillance requirement 4.2.5.3,
which currently requires performance of a precision heat balance
every 18 months, to allow use of the elbow tap [delta]p method for
RCS flow measurement. Appropriate Technical Specification Bases
sections will also be revised to reflect use of the elbow tap
[delta]p method for flow measurement and to provide clarification.
The revised Technical Specifications are in Appendix C.
Background: The 18-month total RCS flow surveillance is
typically satisfied by a secondary power calorimetric-based RCS flow
measurement. In recent cycles, South Texas Project has experienced
apparent decreases in flow rates which have been attributed to
variations in hot leg streaming effects. These effects directly
impact the hot leg temperatures used in the precision calorimetric,
resulting in the calculation of low RCS flow rates. The apparent
flow reduction has become more pronounced in fuel cycles which have
implemented aggressive low leakage loading patterns. Evidence that
the flow reduction was apparent, but not actual, was provided by
elbow tap measurements. The results of this evaluation, including a
detailed description of the hot leg streaming phenomenon, are
documented in Westinghouse report SAE/FSE-TGX/THX-0152, ``RCS Flow
Verification Using Elbow Taps.''
South Texas Project intends to begin using an alternate method
of measuring RCS flow using the elbow tap [delta]p measurements. For
this alternate method, the RCS elbow tap measurements are correlated
to precision

[[Page 40851]]

calorimetric measurements performed during earlier cycles which
decreased the effects of hot leg streaming.
The purpose of this evaluation is to assess the impact of using
the elbow tap [delta]p measurements as an alternate method for
performing the 18-month RCS flow surveillance on the licensing basis
and demonstrate that it will not adversely affect the subsequent
safe operation of the plant. This evaluation supports the conclusion
that implementation of the elbow tap [delta]p measurement as an
alternate method of determining RCS total flow rate does not
represent a significant hazards consideration as defined in
10[]CFR[]50.92.
Evaluation: Use of the elbow tap [delta]p method to determine
RCS total flow requires that the [delta]p measurements for the
present cycle be correlated to the precision calorimetric flow
measurement which was performed during the baseline cycle(s). A
calculation has been performed to determine the uncertainty in the
RCS total flow using this method. This calculation includes the
uncertainty associated with the RCS flow baseline calorimetric
measurement, as well as uncertainties associated with [delta]p
transmitters and indication via QDPS [qualified display processing
system] or the plant process computer. The uncertainty calculation
performed for this method of flow measurement is consistent with the
methodology recommended by the Nuclear Regulatory Commission (NUREG/
CR-3659, PNL-4973, 2/85). The only significant difference is the
assumption of correlation to a previously performed RCS flow
calorimetric. However, this has been accounted for by the addition
of instrument uncertainties previously considered to be zeroed out
by the assumption of normalization to a calorimetric performed each
cycle. Based on these calculations, the uncertainty on the RCS flow
measurement using the elbow tap method is 2.6% flow which results in
a minimum RCS total flow of 391,500 gpm and must be measured via
indication with QDPS or the plant process computer at approximately
100% power.
The specific calculations performed were for Precision RCS Flow
Calorimetrics for the specified baseline cycles, Indicated RCS Flow
(either QDPS or the plant process computer), and the Reactor Coolant
Flow - Low reactor trip. The calculations for Indicated RCS Flow and
Reactor Coolant Flow - Low reactor trip reflect correlation of the
elbow taps to baseline precision RCS Flow Calorimetrics. As
discussed above, additional instrument uncertainties were included
for this correlation.
The uncertainty associated with the RCS Flow - Low trip
increased slightly. It was determined that due to the availability
of margin in the uncertainty calculation, no change was necessary to
either the Trip Setpoint (91.8% flow) or to the current Safety
Analysis Limit (87% flow) to accommodate this increase. The
Allowable Value is to be modified to allow for the increased
instrument uncertainties associated with the [delta]p to flow
correlation.
Since the flow uncertainty did not increase over the currently
analyzed value, no additional evaluations of the reactor core safety
limits must be performed. In addition, it was determined that the
current Minimum Measured Flow (MMF) assumed in the safety analyses
(389,200 gpm) bounds the required MMF calculated for the elbow tap
method (391,500 gpm).
Based on these evaluations, the proposed change would not
invalidate the conclusions presented in the UFSAR [Updated Final
Safety Analysis Report].
1. Does the proposed modification involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Sufficient margin exists to account for all reasonable
instrument uncertainties; therefore, no changes to installed
equipment or hardware in the plant are required, thus the
probability of an accident occurring remains unchanged.
The initial conditions for all accident scenarios modeled are
the same and the conditions at the time of trip, as modeled in the
various safety analyses, are the same. Therefore, the consequences
of an accident will be the same as those previously analyzed.
2. Does the proposed modification create the possibility of a
new or different kind of accident from any accident previously
evaluated?
The proposed change revises the method for RCS flow measurement,
and therefore does not introduce any new accident indicators or
failure mechanisms.
No new accident scenarios have been identified. Operation of the
plant will be consistent with that previously modeled, i.e., the
time of reactor trip in the various safety analyses is the same,
thus plant response will be the same and will not introduce any
different accident scenarios that have not been evaluated.
3. Does the proposed modification involve a significant
reduction in a margin of safety[?]
There are no changes to the Safety Analysis assumptions.
Therefore, the margin of safety will remain the same.
The proposed change does not impact the results from any
accidents analyzed in the safety analysis.
Conclusion: Based on the preceding information, it has been
determined that this proposed change to allow an alternate RCS total
flow measurement based on elbow tap [delta]p measurements does not
involve a Significant Hazards Consideration as defined by 10 CFR
50.92(c).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: James W. Clifford, Acting

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 1, Oswego County, New York

Date of amendment request: July 2, 1997
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 3/4.2.3 regarding reactor coolant
chemistry in accordance with a report by Electrical Power Research
Institute, Inc. (EPRI) TR-103515-R1, ``BWR Water Chemistry Guidelines,
1996 Revision,'' also known as Boiling Water Reactor Vessel and
Internals Project (BWRVIP)-29. Specifically, the amendment would define
new conductivity limits in TS 3.2.3a (when reactor coolant is 200
degrees F or more and reactor thermal power is no more that 10%), and
in TS 3.2.3b (when reactor thermal power exceeds 10%). The new
conductivity limits would be 1 micro-mho/cm, which is less than the
existing limits of 2 micro-mho/cm and 5 micro-mho/cm. The chloride ion
limit in TS 3.2.3a, 0.1 ppm, would remain at this value but would be
designated as 100 ppb. The chloride ion limit in TS 3.2.3b would be
changed from 0.2 ppm to 20 ppb. Sulfate ion limits would be added to TS
3.2.3a and TS 3.2.3b at 100 ppb and 20 ppb, respectively. In TS 3.2.3c,
the maximum conductivity limit would be changed from 10 micro-mho/cm to
5 micro mho/cm when reactor coolant temperature is 200 degrees F or
more; the maximum chloride ion concentration limit would be changed
from 0.5 ppm to 100 ppb (when reactor thermal power exceeds 10%) and
200 ppb (when reactor coolant temperature is 200 degrees F or more and
reactor thermal power is no more than 10%); and the maximum sulfate ion
concentration of 100 ppb (when reactor thermal power exceeds 10%) and
200 ppb (when reactor coolant temperature is 200 degrees F or more and
reactor thermal power is no more than 10%) would be added. The
requirement to place the reactor in the cold shutdown condition as
currently specified in TS 3.2.3d (when TSs 3.2.2a, b, and c are not
met) and TS 3.2.3e (when the continuous conductivity monitor is
inoperable for more than 7 days) would be changed to require that the
reactor coolant temperature be reduced to below 200 degrees F. TS 4.2.3
would be revised to add that the samples taken and analyzed for
conductivity and chloride ion content are also to be analyzed for
sulfate ion content. TS Bases 3/4.2.3 would also be changed to

[[Page 40852]]

reflect that the purpose of TS 3/4.2.3 is to limit crack growth rates
to values consistent with Unit 1 core shroud analyses in accordance
with an NRC letter dated May 8, 1997.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The changes to the conductivity and chloride ion action levels
and the addition of sulfate ion levels as an action level in reactor
water chemistry are being made to make the TS and its Bases
consistent with the values used in the core shroud vertical weld
cracking evaluations. These new values reflect the BWR water
chemistry guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29)
and are equal to or more restrictive than the present TS values. No
physical modification of the plant is involved and no changes to the
methods in which plant systems are operated are required. None of
the precursors of previously evaluated accidents are affected and
therefore, the probability of an accident previously evaluated is
not increased. These changes to the coolant chemistry TS are more
restrictive limits and no new failure modes are introduced.
Therefore, these changes will not involve a significant increase in
the consequences of an accident previously evaluated.
2. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The changes to the conductivity and chloride ion action levels
and the addition of sulfate ion levels as an action level in reactor
water chemistry are being made to make the TS and its Bases
consistent with the values use in the core shroud vertical weld
cracking evaluations. The new values reflect the BWR water chemistry
guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29) and are
equal to or more restrictive than the present TS values. No physical
modification of the plant is involved and no changes to the methods
in which plant systems are operated are required. The change does
not introduce any new failure modes or conditions that may create a
new or different accident. Therefore, this change does not create
the possibility of a new or different kind of accident [from any
accident] previously evaluated.
3. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not involve a significant reduction in
a margin of safety.
The changes to the conductivity and chloride ion action levels
and the addition of sulfate ion levels as an action level in reactor
water chemistry are being made to make the TS and its Bases
consistent with the values used in the core shroud vertical weld
cracking evaluations. These new values reflect the BWR water
chemistry guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29)
and are equal to or more restrictive than the present TS values. No
physical modification of the plant is involved and no changes to the
methods in which plant systems are operated are required. This
change does not adversely affect any physical barrier to the release
of radiation to plant personnel or the public. Therefore, the change
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Alex Dromerick, Acting

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut

Date of amendment request: June 19, 1997
Description of amendment request: Technical Specification Table
2.2-1 Notes 1 and 3 define the values for the constants used in the
Overtemperature Delta-T and Overpower Delta-T reactor trip system
instrumentation setpoint calculators. The proposed amendment would make
changes to the notes as well as the associated Bases section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve an SHC because the
revision would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to Technical Specification Table 2.2-1
Notes 1 and 3 for the addition of the inequalities ensure that the
constants used for [Overtemperature Delta-T] and [Overpower Delta-T]
will be set conservatively with respect to the assumptions in the
accident analysis. The effect on the turbine
runback function has been evaluated with respect to the Loss of
External Electrical Load And/Or Turbine Trip analysis and it has
been determined that this change does not increase the probability
of this transient. The change was also reviewed to determine if it
produced an increase in the probability of an unnecessary or
spurious reactor trip and it was determined that it did not. This
change does not increase the probability of any previously evaluated
accident.
The consequences of previously evaluated accidents, including
Uncontrolled Rod Cluster Assembly Bank Withdrawal At Power, Rod
Cluster Control Assembly Misalignment, Uncontrolled Boron Dilution,
Loss of External Electrical Load And/Or Turbine Trip, Excessive Heat
Removal Due To Feedwater System Malfunctions, Excessive Load
Increase Incident, Accidental Depressurization Of The Reactor
Coolant System, Accidental Depressurization Of The Main Steam
System, Loss of Reactor Coolant From Small Ruptured Pipes Or From
Cracks In Large Pipes Which Actuate ECCS [emergency core cooling
system], or Major Secondary System Pipe Ruptures have not changed.
The administrative changes have no impact on the design or
operation of Millstone Unit 3.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to Technical Specification Table 2.2-1
Notes 1 and 3 do not alter the design, construction, operation,
maintenance or method of testing of equipment. The proposed changes
alter the Technical Specification description of [an]
[Overtemperature Delta-T] and [Overpower Delta-T] setpoint functions
and requires only slight changes to the actual setpoints in the
field. The [Overtemperature Delta-T] and [Overpower Delta-T]
functions serve to mitigate the effects of accidents by opening the
Reactor Trip breakers or reduce power by ``running back'' turbine
electrical load. The change does not create any new interfaces to
plant control or protection systems and therefore, no new mechanism
for accident initiation has been introduced. The proposed change
does not introduce the possibility of an accident of a different
type than previously evaluated.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to Technical Specification Table 2.2-1
Notes 1 and 3 do not affect the integrity of any physical fission
protective boundaries, increase the delays in actuation of safety
systems beyond that assumed in the safety analysis or reduce the

[[Page 40853]]

margin of safety of any system. These changes ensure that actuation
of Overtemperature [Delta-T] and Overpower [Delta-T] reactor trips
will occur conservatively with respect to the assumptions of the
accident analysis.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Deputy Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut

Date of amendment request: June 19, 1997
Description of amendment request: Technical Specification 3/4.7.1.3
requires sufficient water to be available for the auxiliary feedwater
(AFW) system to maintain the reactor coolant system at hot standby for
10 hours before cooling down to hot shutdown in the next 6 hours. The
proposed amendment would increase the required volume of water when the
condensate storage tank is used, make editorial changes, and expand the
description in the appropriate Bases section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve an SHC because the
revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed change to Technical Specification Surveillance
4.7.1.3.2 will account for the unusable Condensate Storage Tank
(CST) inventory by increasing the required combined CST and
Demineralized Water Storage Tank (DWST) inventory to 384,000
gallons. The increased required water volume is consistent with the
design of the CST and will provide assurance that sufficient water
is available to maintain the reactor coolant system at Hot Standby
for 10 hours before cooling down to Hot Shutdown in the next 6
hours.
The proposed changes to reword Technical Specification 3/
4.7.1.3, expand the description in Bases Section B3/4.7.1.3 and
modify the description in Bases Section B3/4.7.1.2 are to update and
clarify the requirements.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to Technical Specification 3/4.7.1.3 do not
change the use of DWST or CST during normal or accident evaluations.
The proposed changes to reword Technical Specification 3/
4.7.1.3, Bases Section B3/4.7.1.3 and Bases Section B3/4.7.1.2 are
to update and clarify the requirements.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to Technical Specification Surveillance
4.7.1.3.2 will increase the required inventory for the combined CST
and DWST to account for an additional 50,000 gallons of unusable
inventory due to the CST discharge line location, other physical
characteristics, and measurement uncertainty. The proposed change to
the surveillance requirement will increase the required volume of
the combined CST and DWST inventory to 384,000 gallons. The proposed
change ensures that sufficient water is available to maintain the
Reactor Coolant System at Hot Standby conditions for 10 hours with
steam discharge to the atmosphere, concurrent with a total loss-of-
offsite power, and with an additional 6-hour cool down period to
reduce reactor coolant temperature to 350 [degrees] F.
The proposed changes to Technical Specification 3/4.7.1.3 and
Bases Section 3/4.7.1.3 are to clarify the requirements. The
proposed changes to the Bases Section 3/4.7.1.2 update and expands
the description of the design bases accidents for which AFW System
is credited for accident mitigation. This additional information is
consistent with the current AFW System design bases.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Deputy Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut

Date of amendment request: June 30, 1997
Description of amendment request: Technical Specification
Surveillance Requirements 4.7.1.5.1 and 4.7.1.5.2 require the periodic
testing of the main steam isolation valves (MSIVs) to demonstrate
operability. The proposed amendment would (1) clarify when the MSIVs
are partial stroked or full closure tested, (2) add a note to the Mode
4 applicability of Technical Specification 3.7.1.5 to require that the
MSIVs be closed and deactivated at less than 320 degrees F, (3) make
editorial changes, and (4) make changes to the associated Bases
sections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to Technical Specifications Surveillances
4.7.1.5.1 and 4.7.1.5.2 are to clarify the testing of the MSIVs by
rewording and separating the requirements into three surveillances.

[[Page 40854]]

Currently, Technical Specifications Surveillance 4.7.1.5.1 requires
``verifying full closure within 10 seconds ... in MODES 1, 2, and 3
when tested pursuant to Specification 4.0.5.'' The current
surveillance requirement to full stroke test the MSIVs is not
performed during power operation as the Millstone Unit 3 Inservice
Pump and Valve Test Program pursuant to Specification 4.0.5, has
received relief from the quarterly full stroke surveillance testing
requirement. The basis for the relief is that full stroking the
MSIVs to the closed position during power operation would result in
an unbalanced steam flow condition producing an abnormal power
distribution in the reactor core, possibly causing a reactor trip.
The MSIVs are equipped with provisions for inservice testing by
partial stroking. The partial stroking is accomplished by opening a
solenoid valve to admit steam pressure into the lower piston
chamber. After a time delay the solenoid valve for the upper piston
chamber opens. After 10 percent travel the position indicating
device vents both piston chambers and the valve fully opens to the
back seat due to pressure acting on the valve plug. The accepted
alternate testing method is to partially stroke test the MSIVs
during power operation and full stroke test the valves during
shutdowns.
The proposed changes to Technical Specifications Surveillance
4.7.1.5.2 will identify a Mode 3 requirement to perform a 10 second
full closure test of the MSIVs in Mode 3 or 4. Surveillance
4.7.1.5.3 will identify a Mode 4 requirement to perform a 120 second
full closure test of the MSIVs in Mode 4 when the RCS [reactor
coolant system] temperature is greater than or equal to 320 degrees
F. The 320 degrees F restriction on testing the valves is consistent
with recommendations from the valve manufacturer. Additionally, a
footnote is added to the LCO [limiting condition for operation] and
the surveillance to identify that the MSIVs are required to be
closed and deactivated when the RCS temperature is less than 320
degrees F.
The proposed changes are consistent with equipment design and
the surveillance testing of the MSIVs provides the necessary
assurance that the valves will function consistent with accident
analyses.
The other proposed changes to reword the Applicability and
Action statements of Technical Specification 3.7.1.5 and Bases
Section B3/4.7.1.5 are considered administrative changes.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to the surveillance testing of the MSIVs
does not change the operation of the valves as assumed for accident
analyses. The MSIVs are currently equipped with provisions for
partial stroking.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to Technical Specifications Surveillances
4.7.1.5.1 and 4.7.1.5.2 are to clarify the testing of the MSIVs by
rewording and separating the requirements into three surveillances.
Surveillance 4.7.1.5.1 will identify a Mode 1 and 2 requirement to
partial stroke test the MSIVs in Mode 1 and 2 unless a successful 10
second full stroke test was performed during the surveillance
period. Surveillance 4.7.1.5.2 will identify a Mode 3 requirement to
perform a 10 second full closure test of the MSIVs in Mode 3 or 4.
Surveillance 4.7.1.5.3 will identify a Mode 4 requirement to perform
a 120 second full closure test of the MSIVs in Mode 4 when the RCS
temperature is greater than or equal to 320 degrees F. The 320
degrees F restriction on testing the valves is consistent with
recommendations from the valve manufacturer. Additionally, a
footnote is added to the LCO and the surveillance to identify that
the MSIVs are required to be closed and deactivated when the RCS
temperature is less than 320 degrees F. The footnote will eliminate
the potential to declare the MSIVs operable in the upper range of
Mode 4 and then allow the MSIVs to remain open during a cooldown
into the lower range of Mode 4 where they may not be able to meet
their required stroke time. The full closure test times are
consistent with the current MSIV surveillances and the partial
stroke testing is consistent with the Millstone Unit 3 Inservice
Pump and Valve Test Program.
The other proposed changes to reword the Applicability and
Action statements of Technical Specification 3.7.1.5 and Bases
Section B3/4.7.1.5 are considered administrative changes.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut

Date of amendment request: June 30, 1997
Description of amendment request: Technical Specifications 4.6.1.1,
3/4.6.1.2, and 3/4.6.1.3 require the testing of the containment to
verify leakage limits at a specified test pressure. The proposed
amendment would (1) modify the list of valves that can be opened in
Modes 1 through 4, (2) add a footnote on procedure controls, (3) remove
a footnote on Type A testing, and (4) make editorial changes to the
Technical Specifications and associated Bases sections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to Technical Specification Surveillance
4.6.1.1.a include the adding ``or procedure control***'' and adding
footnote ``***''. The changes are requested since the Residual Heat
Removal System (RHR) valves, 3RHS*MV8701A/B and 3RHS*MV8702A/B, are
opened during cooldown and heatup in Mode 4. Allowing these
containment isolation valves to be opened is consistent with
Technical Specification 3.4.1.3, Reactor Coolant System - Hot
Shutdown, which allows the RHR system to be used in Mode 4. The
proposed changes to open the RHR system containment isolation
valves, under procedure control in Mode 4, do not change the way the
RHR system is operated or change the operator's response to an
accident in Mode 4.
The proposed changes to Technical Specification Surveillance
4.6.1.1.a Footnote **, include the modification of the valves listed
in the footnote. Valves 3FPW-V661, 3FPW-V666, 3SAS-V875, 3SAS-V50,
3CCP-V886, 3CCP-V887 and 3CVS-V13 are being deleted and are local
manual containment isolation valves. Deleting these valves from the
list of valves that are allowed to be opened under administrative
control does not modify plant response to or mitigation strategy for
any accident. The valves being added, 3MSS*V885, 3MSS*V886, and
3MSS*V887, are in the steam lines to the steam-driven auxiliary
feedwater pump. These valves are opened to warm the steam lines
prior to testing the steam-driven auxiliary feedwater pump. These
valves were recently reclassified as containment isolation valves,
which resulted in the need to add them to the list of valves allowed
to be opened under administrative control. The administrative
controls include the appropriate considerations that when

[[Page 40855]]

required, containment integrity will be established consistent with
the assumptions in the design basis analyses.
The proposed change to Technical Specification Surveillance
4.6.1.2.a will delete footnote ``*'' which referred to an exemption
granted by the NRC to permit the Type A test to be delayed until
RFO6 [refueling outage 6]. However, the current extended shutdown
has significantly delayed RFO6 and NNECO intends to perform the Type
A test during this midcycle shutdown. The deletion of the footnote
does not alter the operation of any system or the containment or
containment airlocks, as assumed for accident analyses.
Additionally, Technical Specifications 4.6.1.1, 3/4.6.1.2 and 3/
4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/4.6.1.3 are
reworded to provide clarity and consistency. These proposed changes
do not alter the operation of any system or the containment or
containment airlocks during accident analyses. Therefore, the
proposed revision does not involve a significant increase in the
probability or consequence of an accident previously evaluated.
1. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to Technical Specifications 4.6.1.1, 3/
4.6.1.2 and 3/4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/
4.6.1.3 do not alter the operation of any system or the containment
or containment airlocks, during normal operation or as assumed in
accident analyses.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to Technical Specifications 4.6.1.1, 3/
4.6.1.2 and 3/4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/
4.6.1.3 do not alter the design, maintenance or function of any
system or the containment or the containment airlocks. Additionally,
the proposed changes do not alter the testing of any system or the
containment or containment airlocks, or alter any assumption used in
the accident analyses.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Deputy Director: Phillip F. McKee

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California

Date of amendment requests: May 14, 1997
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 by revising Technical Specification (TS)
6.9.1.8.b.5 to replace reference WCAP-10266-P-A with WCAP-12945-P for
best estimate loss-of coolant accident (LOCA) analysis. The amendment
would also revise TS Bases 3/4.2.2 and 3/4.2.3 to change the emergency
core cooling system (ECCS) acceptance criteria limit to state that
there is a high level of probability that the ECCS acceptance criteria
limits are not exceeded. This is consistent with the best estimate LOCA
methodology.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change to use of the Best Estimate Loss of Coolant
Accident (LOCA) analysis methodology does not involve physical
alteration of any plant equipment or change in operating practice at
Diablo Canyon Power Plant (DCPP). Therefore, there will be no
increase in the probability of a LOCA. The consequences of a LOCA
are not being increased.
The plant conditions assumed in the analysis are bounded by the
design conditions for all equipment in the plant. That is, it is
shown that the emergency core cooling system is designed so that its
calculated cooling performance conforms to the criteria contained in
10 CFR 50.46, paragraph b, and it meets the five criteria listed in
Section D. of this evaluation. No other accident is potentially
affected by this change.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change would not result in any physical alteration
to any plant system, and there would not be a change in the method
by which any safety related system performs its function. The
parameters assumed in the analysis are within the design limits of
existing plant equipment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
It has been shown that the analytic technique used in the
analysis realistically describes the expected behavior of the DCPP
Units 1 and 2 reactor system during a postulated LOCA. Uncertainties
have been accounted for as required by 10 CFR 50.46. A sufficient
number of LOCAs with different break sizes, different locations, and
other variations in properties have been analyzed to provide
assurance that the most severe postulated LOCAs were calculated. It
has been shown by the analysis that there is a high level of
probability that all criteria contained in 10 CFR 50.46, paragraph
b, are met.
Therefore the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California

Date of amendment requests: May 15, 1997
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant (DCPP), Unit Nos. 1 and 2 to revise the surveillance
frequencies from at least once every 18 months to at least once per
refueling interval (nominally 24 months) including (1) reactor coolant
system total flow rate, (2) instrumentation for radiation monitoring,
(3) instrumentation and controls for remote shutdown, (4)
instrumentation for

[[Page 40856]]

accident monitoring, and (5) several miscellaneous TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS surveillance interval increases do not alter the
intent or method by which the inspections, tests, or verifications
are conducted, do not alter the way any structure, system, or
component functions, and do not change the manner in which the plant
is operated. The surveillance, maintenance, and operating histories
indicate that the equipment will continue to perform satisfactorily
with longer surveillance intervals. Few surveillance and maintenance
problems were identified. No problems have recurred, or are expected
to recur, following identification of root causes and implementation
of corrective actions.
There was one time-related degradation mechanism identified that
could significantly degrade the performance of the evaluated
equipment during normal plant operation. Accumulation of corrosion
products and debris in the containment fan cooler unit (CFCU)
monitoring system drain lines could affect the use of the CFCU
drains as a backup to the containment gaseous monitor for RCS leak
detection. Primarily because CFCU drain line cleaning has been
instituted to reduce deposit buildup, and also because the CFCU
monitoring systems are used as backup and they are redundant by a
factor of five, it was evaluated that this time-related mechanism
will not significantly degrade the leak detection performance of the
CFCUs.
All other potential time-related degradation mechanisms have
insignificant effects in the period of interest (24 months plus 25
percent allowance, or a maximum of 30 months). Instrument drift and
uncertainty analyses show that, while slight increases in instrument
drift can occur over a longer period, such increases are minimal and
remain within specified instrument accuracy and calibration
allowable values. In cases (pressurizer water level and RVLIS) where
greater than expected instrument drift has been found, design and
procedural changes have been implemented to improve the calibration
process and instrument performance. Based on the past performance of
the equipment, the probability or consequences of accidents
previously evaluated would not be significantly affected by the
proposed surveillance interval increases.
The changes to commitments related to Bulletin 90-01 are
supported by the conclusions above, and otherwise do not alter the
intent or method by which the associated functions are tested, do
not alter the way any structure, system, or component functions, and
do not change the manner in which the plant is operated.
The administrative changes to the Bases sections and to remove a
duplicate line do not alter the frequency, intent, or method by
which the associated functions are tested, do not alter the way any
structure, system, or component functions, and do not change the
manner in which the plant is operated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The surveillance and maintenance histories indicate that the
equipment will continue to effectively perform its design function
over the longer operating cycles. Additionally, the increased
surveillance intervals do not result in any physical modifications,
affect safety function performance or the manner in which the plant
is operated, or alter the intent or method by which surveillance
tests are performed. No problems have reoccurred following
identification of root causes and implementation of corrective
actions. Almost all identified potential time-related degradations,
including instrument drift, have insignificant effects in the period
of interest.
The deposit buildup in the CFCU drain lines is time-related.
This was evaluated to not to be significant to the leak detection
function because the CFCUs have a redundancy factor of five (any one
of the five CFCUs can be used for the leak detection function) and
because the CFCU drain lines will be cleaned each refueling outage.
The proposed surveillance interval increases would not affect the
type or possibility of accidents.
The changes to commitments related to Bulletin 90-01 are
supported by the conclusions above, and otherwise do not result in
any physical modifications, affect safety function performance or
the manner in which the plant is operated, or alter the intent or
method by which surveillance tests are performed.
The administrative change to the Bases sections and to remove a
duplicate line do not result in any physical modifications, affect
safety function performance, or alter the frequency, intent, or
method by which surveillance tests are performed.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Evaluation of historical surveillance and maintenance data
indicates that there have been few problems experienced with the
evaluated equipment. There are no indications that potential
problems would be cycle-length dependent, with the exception of the
CFCU leak detection function, or that potential degradation would be
significant for the period of interest and, therefore, increasing
the surveillance interval will have negligible impact on safety. The
accumulation of corrosion products and debris in the CFCU drain
lines is cycle-length dependent, but has been evaluated to have
insignificant effect on its leak detection function. There is no
safety analysis impact since these changes will have no effect on
any safety limit, protection system setpoint, or limiting condition
for operation, and there are no hardware changes that would impact
existing safety analysis acceptance criteria. Safety margins are not
significantly impacted by surveillance intervals or by the slight
increases in instrument drift that may occur during the extended
interval.
The changes to commitments related to Bulletin 90-01 are
supported by the conclusions above, and otherwise will have no
effect on any safety limit, protection system setpoint, or limiting
condition for operation, and there are no hardware changes that
would impact existing safety analysis acceptance criteria.
The administrative change to the Bases sections and to remove a
duplicate line will have no effect on any safety limit, protection
system setpoint, or limiting condition for operation, and there are
no hardware changes that would impact existing safety analysis
acceptance criteria.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay
Power Plant, Unit 3, Humboldt County, California

Date of amendment request: December 9, 1996, as supplemented on
June 12, 1997.
Description of amendment request: The proposed amendment would
revise the Humboldt Bay Power Plant (HBPP), Unit 3 Technical
Specifications (TSs) to incorporate the requirements of 10 CFR Part 50,
Appendix I, into the Radiological Effluent Technical Specifications
(RETS) and to relocate the controls and limitations on RETS and
radiological monitoring from the technical specifications to the
Offsite Dose Calculation Manual (ODCM) and the Process Control Program
(PCP). Additional minor administrative changes are proposed to make the
TSs on High Radiation Areas consistent with

[[Page 40857]]

the revised requirements in the new 10 CFR Part 20.
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?
Operation of the facility in accordance with the proposed
amendment would not involve any increase in the probability or
consequences of an accident previously evaluated. This change places
new requirements in the Administrative Controls section of the
Technical Specifications to establish programs for the control of
radiological effluents and the conduct of radiological environmental
monitoring in the ODCM. The new Administrative Control requirements
for radiological effluents to be placed in the ODCM incorporate 10
CFR 50, Appendix I, limitations on dose to individual members of the
public that are much more restrictive than the current Technical
Specification limitations. The proposed changes do not involve
modifications to existing plant equipment, the addition of new
equipment, or operation of the plant in a different manner than
previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability of 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?
Operation on the facility in accordance with the proposed
amendment will not create any new or different kind of accident from
any accident previously evaluated. As stated above, new programmatic
controls on radiological effluents and radiological environmental
monitoring are established in the Administrative Controls section of
the Technical Specifications. Additionally, this change is
administrative in nature; procedural details for radiological
effluents and radiological environmental monitoring are being
relocated to the ODCM and PCP consistent with the guidance provided
[by the NRC] in Generic Letter 89-01. The proposed changes do not
involve alterations to plant operating philosophy or methods, or in
changes to installed plant systems, structures, or components.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Operation of the facility in accordance with the proposed
amendment would not involve any reduction in the margin of safety.
These changes do not involve a significant reduction in the margin
of safety. These changes do not involve a significant reduction in
the margin of safety. The changes will provide control over
radiological effluent releases, solid waste management, and
radiological environmental monitoring activities. Also, these
changes will increase the margin of safety for members of the public
by imposing additional controls to ensure that dose to members of
the public resulting from radioactive effluent releases will be
maintained ALARA.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis of the licensee and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Humboldt County Library, 636 F
Street, Eureka, California 95501
Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas
and Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: Seymour H. Weiss

Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of
Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch
Nuclear Plant, Unit 1, Appling County, Georgia

Date of amendment request: May 9, 1997
Description of amendment request: The proposed amendment would
revise the Safety Limit Minimum Critical Power Ratio (SLMCPR) in
Technical Specification (TS) 2.1.1.2 to reflect results of a cycle-
specific calculation performed for Unit 1 Operating Cycle 18 (expected
to commence November 1997).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed technical specification changes do not involve a
significant increase in the probability of an accident previously
evaluated.
The derivation of the revised SLMCPR for Plant Hatch Unit 1
Cycle 18 for incorporation into the TS, and its use to determine
cycle-specific thermal limits, have been performed using NRC
approved methods. Additionally, interim implementing procedures that
incorporate cycle-specific parameters have been used which result in
a more restrictive value for SLMCPR. These calculations do not
change the method of operating the plantand have no effect on the
probability of an accident initiating event or transient.
The basis of the MCPR Safety Limit is to ensure no mechanistic
fuel damage is calculated to occur if the limit is not violated. The
new SLMCPR preserves the existing margin to transition boiling and
the probability of fuel damage is not increased. Therefore, the
proposed changes do not involve an increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes result only from a revised method of
analysis for the Unit 1 Cycle 18 core reload. These changes do not
involve any new method for operating the facility and do not involve
any facility modifications. No new initiating events or transients
result from these changes. Therefore, the proposed TS changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS bases will remain the
same. The new SLMCPR is calculated using NRC approved methods which
are in accordance with the current fuel design and licensing
criteria. Additionally, interim implementing procedures, which
incorporate cycle-specific parameters, have been used. The SLMCPR
remains high enough to ensure that greater than 99.9% of all fuel
rods in the core are expected to avoid transition boiling if the
limit is not violated, thereby preserving the fuel cladding
integrity.
Therefore, the proposed TS changes do not involve a reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366,
Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,
Georgia

Date of amendment request: May 9, 1997
Description of amendment request: The proposed amendments would
revise the operability requirements for the Rod Block Monitor system of
Technical Specification (TS) Table 3.3.2.1-1. The amendments would also

[[Page 40858]]

delete the requirements of TS Section 5.6.5 to report Rod Block Monitor
operability requirements in the cycle-specific Core Operating Limits
Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:

Southern Nuclear Operating Company has evaluated the proposed
changes to the Plant Hatch Units 1 and 2 Technical Specifications
in accordance with the criteria specified in 10 CFR 50.92 and has
determined that they do not involve a significant hazards
consideration because:

1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
since they are more restrictive than the existing requirements for
operation of the plant. These changes provide assurance that the Rod
Block Monitor system will remain operable when necessary to prevent
or mitigate the consequences of an anticipated operational
occurrence that could threaten the integrity of the fuel cladding
integrity. Since changes in RBM [Rod Block Monitor] operability
requirements do not involve any physical or functional modifications
in any plant system, structure or component, there will be no
increase in the probability or consequences of any accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated because
they do not involve any changes in the plant configuration or in the
operation of any system, structure or component.
3. The proposed changes do not reduce a margin of safety in the
plant because they impose more restrictive operability requirements
on the Rod Block Monitor system than those imposed by the existing
specifications. The changes are more restrictive in that they delete
the conditions under which the RBM is allowed to be bypassed at core
thermal power equal to or greater than 29% of rated power. These
more restrictive requirements ensure the RBM will not only prevent
fuel rods from under going transition boiling, they also prevent
fuel rods from exceeding 1% plastic strain (thereby avoiding fuel
cladding damage) during an RWE [rod withdrawal error] event.
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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns
Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama

Date of amendment request: June 19, 1997 (TS 391T)
Description of amendment request: The proposed amendment extends
the allowed outage time for emergency diesel generators from 7 to 14
days on a one-time basis. This extension should permit completion of
extensive recommended maintenance within a single outage interval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the ssue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The EDGs [emergency diesel generators] are designed as backup AC
[alternating current] power sources in the event of loss of off-site
power. The proposed AOT [allowed outage time] does not change the
conditions, operating configurations, or minimum amount of operating
equipment assumed in the safety analysis for accident mitigation. No
changes are proposed in the manner in which the EDGs provide plant
protection or which create new modes of plant operation. Also, the
TS [technical specification] change will improve the overall EDG
availability by allowing the consolidation of planned maintenance
outages and, hence, reducing the time period that each EDG will be
in an outage. Therefore, the proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not introduce any new modes of plant
operation or make physical changes to plant systems. Therefore, the
proposed one-time extension of the allowable AOT for EDGs does not
create the possibility of a new or different accident.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
BFN's [Browns Ferry Nuclear Plant's] emergency AC system is
designed with sufficient redundancy such that an EDG may be removed
from service for maintenance or testing. The remaining EDGs are
capable of carrying sufficient electrical loads to satisfy the UFSAR
[updated final safety analysis report] requirements for accident
mitigation or unit safe shutdown.
Since the 12-year EDG PM [preventive maintenance] work activity
and vendor recommended PMs are required tasks which must be
performed, the proposed TS would reduce EDG unavailability since
multiple outages with resultant longer EDG outage times would not be
necessary to accomplish the planned maintenance activities.
The proposed change does not impact the redundancy or
availability requirements of off-site power supplies or change the
ability of the plant to cope with station blackout events. The TS
change improves overall EDG availability. For these reasons, the
proposed amendment does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: June 24, 1997
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 3/4.3.2.1, ``Safety
Features Actuation System Instrumentation,'' TS Section 3/4.6.1.7,
``Containment Ventilation System,'' TS Section 3/4.6.3.1, ``Containment
Isolation Valves,'' and TS Section 3/4.9.4, ``Refueling Operations -
Containment Penetrations,'' and the associated TS Bases. Valve position
requirements would be added, and certain containment radiation monitor
requirements, valve isolation verification requirements, and
containment radiation monitor optional uses would be deleted.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation ofthe Davis-Besse Nuclear Power

[[Page 40859]]

Station (DBNPS), Unit No. 1, in accordance with this change would:
1a Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions, or assumptions are affected by the proposed changes.
The proposed changes to the Technical Specifications and their
Bases ensure that during Modes 1 through 4 the Containment (CTMT)
purge and exhaust isolation valves are closed with control power
removed. Having these valves closed will not increase the
probability of an accident because these valves are not accident
initiators. They are used to mitigate the consequences of an
accident. The proposed changes require these valves to be maintained
in a closed position as required by design basis accident analysis.
The removal of the Safety Features Actuation System (SFAS)
Radiation Monitors (RE's) and their associated SFAS Level 1
actuations does not affect any accident initiator, condition, or
assumption.
During Modes 1 and 2 and partially in Mode 3, for design basis
accidents which require CTMT isolation, the high/high-high CTMT
pressure or low/low-low Reactor Coolant System (RCS) signals provide
CTMT isolation and isolation and actuation of those components
presently actuated by an SFAS Level 1 High Radiation signal. During
Mode 3, when the RCS pressure is below 1800 psig, the low RCS
pressure trip may be manually bypassed, and when the RCS pressure is
below 600 psig, the low-low pressure trip may be manually bypassed.
During the short period of time that these bypasses are activated in
Mode 3, CTMT isolation is only automatically initiated by the CTMT
high/high-high pressure trips. Manual SFAS actuation is also
available, including Modes 1 through 4. Removing the SFAS RE's does
not affect the operation of the SFAS Levels 2-4 actuation since
these are based only on containment pressure and RCS pressure.
Therefore, the assumption of CTMT isolation following design basis
accidents is maintained.
The SFAS is not required in Mode 5. During Mode 6, the SFAS RE's
and their associated SFAS Level 1 actuation are not credited during
a fuel handling accident inside CTMT. The analysis for a fuel
handling accident inside CTMT assumes that there is no isolation of
CTMT. The probability of a fuel handling accident is not affected by
these changes.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
change the source term, CTMT isolation, or allowable releases.
The proposed changes to the Technical Specifications and their
Bases ensure that during Modes 1 through 4, the CTMT purge and
exhaust isolation valves are closed with control power removed.
Having these valves closed and their control power removed
ensures that the valves are in and will remain in, the proper
position for CTMT isolation during and following design basis
accidents. Also, during Modes 1 and 2 and partially in Mode 3, SFAS
actuation on high/high-high CTMT pressure or low/low-low RCS
pressure provides for diverse CTMT isolation. As noted above, during
Mode 3, when the RCS pressure is below 1800 psig, the low RCS
pressure trip may be manually bypassed, and when the RCS pressure is
below 600 psig, the low-low pressure trip may be manually bypassed.
During the short period of time that these bypasses are activated in
Mode 3, CTMT isolation is only automatically initiated by the CTMT
high/high-high pressure trips. In addition, manual SFAS actuation is
also available, including during Modes 1 through 4. Therefore,
removal of the SFAS RE's and their actuation signal does not prevent
CTMT isolation.
The SFAS RE's and automatic isolation of the CTMT purge and
exhaust isolation valves during a fuel handling accident is not
required because the CTMT purge and exhaust isolation system,
including the associated noble gas monitor, with operator action,
can provide the necessary actions to mitigate a fuel handling
accident inside CTMT, assuming the purge and exhaust valves are
open. Therefore, removing the SFAS RE's and their actuation signal
will not increase the consequences of an accident because CTMT
closure is ensured. Further, it is noted that CTMT isolation is not
assumed in the accident analysis for the fuel handling accident.
The Containment Radiation-High trip feature is not credited for
any DBNPS Updated Safety Analysis Report (USAR) accident analysis,
therefore the proposed removal of this feature will not impact
radiological consequences of such accidents.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes.
As stated above, the CTMT purge and exhaust isolation valves,
the SFAS RE's, and SFAS actuation are not accident initiators.
Maintaining the CTMT purge and exhaust isolation valves closed and
control power removed ensures that the design basis assumption of
CTMT isolation is maintained. Also, since SFAS Levels 2-4 actuation,
as applicable, on high/high-high CTMT pressure or low/low-low RCS
pressure or by manual actuation provides the required diversity of
sensing parameters and isolation of CTMT, the SFAS RE's and their
associated automatic isolation of the CTMT purge and exhaust
isolation valves is not required during Modes 1 through 4.
Therefore, no new or different kind of accident will be introduced.
3. Not involve a significant reduction in a margin of safety
because the proposed changes maintain a redundant and diverse CTMT
isolation capability following design basis accidents. Under TS 3/
4.3.2, diversity in achieving CTMT isolation by means of a high/
high-high CTMT pressure or low/low-low RCS pressure SFAS actuation
will be maintained during Modes 1 through 3 (except during brief
periods of bypass in Mode 3), and the redundancy of the SFAS sensor
instrumentation channels and actuation channels themselves will be
maintained. During Modes 1 through 4 the manual actuation capability
of SFAS will be maintained. During Modes 1 through 4, control room
indication of normal and accident range radiation monitoring will be
maintained in accordance with TS 3/4.3.3.1 and 3/4.4.6.1.
Under TS 3/4.6.1.7, requiring the CTMT purge and exhaust
isolation valves to be closed with control power removed, and
requiring an open CTMT purge and exhaust isolation valve to be
closed with control power removed within 24 hours is more
restrictive than the current Technical Specifications or ``The
Improved Standard Technical Specifications for Babcock and Wilcox
Plants,'' NUREG-1430, Revision 1. Under TS 3/4.9.4, the existing
requirements already allow for the SFAS-initiated closure of the
CTMT purge and exhaust isolation valves to be unavailable and the
CTMT purge and exhaust system noble gas monitor used as an
alternative means of achieving CTMT isolation. Further, it is noted
that CTMT isolation is not credited in the accident analysis for the
fuel handling accident. Therefore, these proposed changes do not
significantly reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri

Date of application request: April 24, 1997, as supplemented by
letters dated June 6, 1997, and June 27, 1997.

Description of amendment request: The amendment would revise
Section 6.0 of the Technical Specifications to change the title
``Senior Vice President, Nuclear'' to ``Vice President and Chief
Nuclear Officer.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This proposed change does not involve any hardware or design
changes, plant procedures, or administrative changes, other than a
revision of title designation in documentation. Within the Union
Electric

[[Page 40860]]

organizational structure, the departments reporting to the former
Senior Vice-President, Nuclear now report to the Vice President and
Chief Nuclear Officer. The position of Vice-President and Chief
Nuclear Officer now reports to the President & Chief Executive
Officer of Union Electric, which is the same management level of
reporting as the previous title, Senior Vice-President, Nuclear.
This change has no impact on the probability or consequences of
accidents previously evaluated in the Final Safety Report (FSAR).
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This proposed change does not involve any hardware or design
changes, plant procedures, or administrative changes, other than a
revision of title designation in documentation. Within the Union
Electric organizational structure, the departments reporting to the
former Senior Vice-President, Nuclear now report to the Vice
President and Chief Nuclear Officer. The position of Vice-President
and Chief Nuclear Officer now reports to the President & Chief
Executive Officer of Union Electric, which is the same management
level of reporting as the previous title, Senior Vice-President,
Nuclear. No new or different kind of accident is introduced by this
purely administrative change to revise documentation to reflect
current organizational titles.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The safety margins of the Technical Specifications are based on
the actual plant design and are unaffected by this purely
administrative change. This change merely updates the Technical
Specifications to reflect the current organizational title for
senior management of the Callaway Plant, and within the
organizational structure of Union Electric. This change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman

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: May 14, 1997
Description of amendment request: The proposed change will provide
clarification to the testing and inspection requirements that each of
the turbine control valves be cycled and movement verified through at
least one complete cycle from the running position and revise the
current wording in Surveillance Requirement 4.7.1.7.2.a for both units
to clarify the testing and inspection methodology of the turbine
control valves. Additionally, Technical Specification Bases Section 3/
4.7.1.7 will be revised to clarify the testing requirements for the
turbine governor control valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of the North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No new or unique accident precursors are introduced by these
changes in surveillance requirements. The clarification for the
turbine control valve testing and inspections do not change
the design, operation, or failure modes of the valves and other
components in the turbine overspeed protection system.
The verification of the operability of the turbine control
valves will continue to provide adequate assurance that the turbine
overspeed protection system will operate as designed, if needed.
Therefore, these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previous[ly] evaluated.
Since the implementation of the proposed change to the
surveillance requirements is to clarify the wording only, operation
of the facilities with these proposed Technical Specifications does
not create the possibility for any new or different kind of accident
which has not already been evaluated in the Updated Final Safety
Analysis Report (UFSAR).
The proposed wording changes to the Technical Specifications
will not result in any physical alteration to any plant system, nor
would there be a change in the method by which any safety-related
system performs its function. The design and operation of the
turbine overspeed protection and turbine control systems are not
being changed. The proposed change merely represents a clarification
to more specifically state current test requirements and test
practice.
These changes do not change the design, operation, or failure
modes of the valves and other components of the turbine overspeed
protection system. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes would not reduce the margin of safety as
defined in the basis for any Technical Specifications. The design
and operation of the turbine overspeed protection and turbine
control systems are not being changed and the operability of the
turbine control valves are being demonstrated in the same manner. In
addition, the results of the accident analyses which are documented
in the UFSAR continue to bound operation under the proposed changes,
so that there is no safety margin reduction. Therefore, the proposed
change does not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: Gordon E. Edison, Acting

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas

Date of amendment request: July 3, 1997
Description of amendment request: This license amendment request
revises Technical Specification Section 5.3.1, Fuel Assemblies, to
allow the use of an alternate zirconium based fuel cladding material,
ZIRLO. Wolf Creek Nuclear Operating Corporation (WCNOC) is planning to
insert Westinghouse fuel assemblies containing ZIRLO fuel rod cladding
during the ninth refueling outage, which is currently scheduled to
begin in October 1997. This request proposes to incorporate additional
information, associated with the requested change, into Technical
Specification 6.9.1.9, ``CORE OPERATING LIMITS REPORT (COLR).'' This
revised submittal supersedes the staff's proposed no significant
hazards consideration determination evaluation for the requested
changes that were published on April 23, 1997 (62 FR 19839).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the

[[Page 40861]]

issue of no significant ha

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A97-11910. Public record. Not legal advice.