# Biweekly Notice

> Briefs, arguments, decisions, and more.

URL: https://www.frixlaw.com/law-library/documents/fr%3AX95-20816

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** August 16, 1995
- **Citation:** 60 FR 42597

## Text

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving
No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 21, 1995, through August 4, 1995. The
last biweekly notice was published on Wednesday, August 2, 1995 (60 FR
39430).

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

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By September 15, 1995, 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

[[Page 42598]]
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland

Date of amendments request: March 15, 1995, as supplemented on June
29, 1995.
Description of amendments request: The proposed amendments would
revise the Calvert Cliffs Nuclear Power Plant, Units Nos. 1 and 2,
Technical Specifications (TSs) Section 6, ``Administrative Controls,''
to be consistent with the guidance provided in NUREG-1432, ``Standard
Technical Specifications, Combustion Engineering Plants.'' The proposed
changes will relocate several requirements to other documents and
programs consistent with NUREG-1432 and other NRC guidance addressing
the administrative section of the TSs such as the ``Final Policy
Statement on Technical Specification Improvements for Nuclear Power
Reactors,'' published in the Federal Register on July 22, 1993 (58 FR
39132).
The Commission indicated that compliance with the Final Policy
Statement satisfies Section 182a of the Act. In particular, the
Commission indicated that certain items could be relocated from the TSs
to licensee-controlled documents, consistent with the standard
enunciated in Portland General Electric Co. (Trojan Nuclear Plant),
ALAB-531, 9 NRC 263, 273 (1979). In that case, the Atomic Safety and
Licensing Appeal Board indicated that ``technical specifications are to
be reserved for those matters as to which the imposition of rigid
conditions or limitations upon reactor operation is deemed necessary to
obviate the possibility of an abnormal situation or event giving rise
to an immediate threat to the public health and safety.'' The policy
statement encouraged licensees to adopt the applicable improved STSs
and provided some guidance for the conversion from the present plant-
specific TSs to the improved Standard TSs.
The proposed changes will provide significant human factors
improvement to the TSs by accomplishing the following: (1) relocating
existing requirements to licensee controlled documents consistent with
the policy statement; (2) eliminating requirements which duplicate
regulations; (3) relocating similar requirements within the same
section; (4) editorial changes; and (5) adding requirements consistent
with NUREG-1432.
In addition, the licensee proposes dual rolls for the Shift
Technical Advisor (STA) and the establishment of a TS Bases Control
Program. Allowing the STA to perform dual rolls is not permitted by the
current TSs, but the current NRC guidance allows the STA to perform a
dual roll. The proposed new TS Bases Control Program will define the
appropriate methods and reviews required to implement a TS Bases change
which is also consistent with the current NRC guidance. Two other
proposed changes, not specifically covered by the above groupings,
include a reduction in reporting requirements and utilizing a more
effective option for estimating doses.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Relocating existing requirements to Baltimore Gas and Electric
Company (BGE)-controlled documents, eliminating requirements which
duplicate regulations, locating similar requirements within the same
sections and making necessary editorial corrections to incorporate
the proposed changes provide Technical Specifications which are
easier to use. Because existing requirements are relocated to
established BGE programs where changes to those programs are
controlled by regulatory requirements, there is no reduction in
commitment and adequate control is still maintained. Likewise, the
elimination of requirements which duplicate regulations enhances the
usability of the Technical Specifications without reducing
commitments. Locating similar requirements within the same sections
and making necessary editorial corrections to incorporate the
proposed changes neither add nor delete requirements, but merely
clarify and improve the readability and understanding of the
Technical Specifications. Since the requirements remain the same,
these changes only affect the method of presentation and would not
affect possible initiating events for accidents previously evaluated
or any system functional requirement. Therefore, the proposed
changes would not involve a significant increase in the probability
or consequences of an accident previously evaluated.

[[Page 42599]]

Since the Shift Technical Adviser (STA) is not considered an
initiator to any previously evaluated accident nor considered in the
accident's response, the use of a dual role STA would not increase
the probability or consequences of any previously evaluated
accident.
The Technical Specification Bases Control Program provides
controls which ensure appropriate reviews of changes to the Bases.
Because NRC approval is still needed for changes to the Bases which
affect the Technical Specifications, the proposed Program would not
affect the probability or consequences of a previously evaluated
accident.
Eliminating the requirement for submitting two reports which
place unwarranted administrative burden on both Baltimore Gas and
Electric Company and the NRC has no affect on the probability or
consequences of an accident previously evaluated. Therefore, the
proposed changes would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Replacing the film badge with the electronic personal dosimeter
provides a more effective, efficient, state-of-the art option for
estimating dose and would not impact accidents previously evaluated.
Therefore, the proposed change would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
As discussed previously, relocating existing requirements to
BGE-controlled documents, eliminating requirements which duplicate
regulations, locating similar requirements within the same sections
and making necessary editorial corrections to incorporate the
proposed changes will not affect any plant system or structure, nor
will it affect any system functional or operability requirements.
Consequently, no new failure modes are introduced as a result of the
proposed changes. Therefore, these types of changes would not create
the possibility of a new or different type of accident from any
accident previously evaluated.
Because the STA does not perform equipment design or equipment
manipulation, the use of a dual role STA would not create the
possibility of a new or different type of accident from any accident
previously evaluated. Since the Technical Specification Bases
Control Program represents an administrative function performed
under existing regulatory controls, it too would not create the
possibility of a new or different type of accident from any
previously evaluated.
The addition of new programs which incorporate existing
Technical Specification requirements and commitments will have no
effect on the design or operation of the plant and would not create
the possibility of a new or different type of accident from any
previously evaluated.
A reporting function such as report submittals would not change
the configuration or operation of the plant. Consequently, the
elimination of the requirement to submit the Startup Report and the
Special Report dealing with iodine activity levels, would not create
the possibility of a new or different type of accident from any
accident previously evaluated.
Since the operation or configuration of the plant is not changed
by the type of personal dosimeter, this change would not create the
possibility of a new or different type of accident from any accident
previously evaluated.
Therefore, the proposed changes would not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
Relocating existing requirements to BGE-controlled documents,
eliminating requirements which duplicate regulations, locating
similar requirements within the same sections and making necessary
editorial corrections to incorporate the proposed changes would not
affect the Updated Final Safety Analysis Report design bases,
accident analysis assumptions or any margin of safety described in
the Technical Specification Bases. In addition, these proposed
changes do not affect effluent release limits, monitoring equipment
or practices. Therefore, these proposed changes would not involve a
significant reduction in a margin of safety.
The use of an STA should provide an additional margin of safety
in the accident response function of licensed operators beyond that
considered in the accident analysis. Since the STA is required to
have the same training and educational qualifications in either the
individual or dual role, the use of a dual role STA should have
minimal impact. Consequently, the proposed change would not involve
a significant reduction in a margin of safety. The Technical
Specification Bases Control Program is an administrative change
controlling how Technical Specification basis information is
reviewed and incorporated. Therefore, this change would not involve
a significant reduction in a margin of safety.
The addition of new programs which incorporate existing
Technical Specification requirements and commitments will have no
effect on the design or operation of the plant and would not result
in a significant reduction in the margin of safety.
Activities described in the Startup Report will continue to be
performed and corrective action taken when required. Similarly,
iodine activity levels will continue to be monitored and actions
taken, including the issuance of a Licensee Event Report when
conditions warrant. Considering the above, elimination of the two
reporting requirements would have no impact on the margin of safety.
Plant operating parameters are not affected by the type of
personnel monitoring device used and as a consequence, would not
impact a margin of safety. Since the replacement dosimeter provides
a more effective mechanism for estimating dose, there is no
degradation in personal safety levels. Consequently, the proposed
change would not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois

Date of application for amendment requests: September 17, 1993, as
supplemented July 28, 1995
Description of amendment requests: As a result of findings by a
Diagnostic Evaluation Team inspection performed by the NRC staff at the
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company
(ComEd, the licensee) made a decision that both the Dresden Nuclear
Power Station and sister site Quad Cities Nuclear Power Station needed
attention focused on the existing custom Technical Specifications (TS)
used.
The licensee made the decision to initiate a Technical
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities.
The licensee evaluated the current TS for both Dresden and Quad Cities
against the Standard Technical Specifications (STS) contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' The licensee's evaluation identified numerous potential
improvements such as clarifying requirements, changing TS to make them
more understandable and to eliminate interpretation, and deleting
requirements that are no longer considered current with industry
practice. As a result of the evaluation, ComEd has elected to upgrade
both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.
The TSUP for Dresden and Quad Cities is not a complete adoption of
the STS. The TSUP focuses on (1) integrating additional information
such as equipment operability requirements during shutdown conditions,
(2) clarifying requirements such as limiting conditions for operation
and action

[[Page 42600]]
statements utilizing STS terminology, (3) deleting superseded
requirements and modifications to the TS based on the licensee's
responses to Generic Letters (GL), and (4) relocating specific items to
more appropriate TS locations.
The September 17, 1993, and July 28, 1995, applications proposed to
upgrade only Section 3/4.5 (Emergency Core Cooling Systems) of the
Dresden and Quad Cities TS.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis.
Implementation of these changes will provide increased reliability
of equipment assumed to operate in the current safety analysis, or
provide continued assurance that specified parameters remain within
their acceptance limits, and as such, will not significantly
increase the probability or consequences of a previously evaluated
accident.
Some of the proposed changes represent minor curtailments of the
current requirements which are based on generic guidance or
previously approved provisions for other stations. The proposed
amendment for Dresden and Quad Cities Station's Technical
Specification Section 3/4.5 are based on STS guidelines or later
operating BWR plants' NRC accepted changes. Any deviations from STS
requirements do not significantly increase the probability or
consequences of any previously evaluated accidents for Dresden or
Quad Cities Stations. The proposed amendment is consistent with the
current safety analyses and has been previously determined to
represent sufficient requirements for the assurance and reliability
of equipment assumed to operate in the safety analysis, or provide
continued assurance that specified parameters remain within their
acceptance limits. As such, these changes will not significantly
increase the probability or consequences of a previously evaluated
accident.
The associated systems that make up the Emergency Core Cooling
Systems are not assumed in any safety analysis to initiate any
accident sequence for Dresden or Quad Cities Stations; therefore,
the probability of any accident previously evaluated is not
increased by the proposed amendment. In addition, the proposed
surveillance requirements for the proposed amendments to these
systems are generally more prescriptive than the current
requirements specified within the Technical Specifications. The
additional surveillance requirements improve the reliability and
availability of all affected systems and therefore, reduce the
consequences of any accident previously evaluated as the probability
of the systems outlined within Section 3/4.5 of the proposed
Technical Specifications performing their intended function is
increased by the additional surveillances.
Create the possibility of a new or different kind of accident
from any previously evaluated because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, the addition of
requirements which are based on the current safety analysis, and
some minor curtailments of the current requirements which are based
on generic guidance or previously approved provisions for other
stations. These changes do not involve revisions to the design of
the station. Some of the changes may involve revision in the
operation of the station; however, these provide additional
restrictions which are in accordance with the current safety
analysis, or are to provide for additional testing or surveillances
which will not introduce new failure mechanisms beyond those already
considered in the current safety analyses.
The proposed amendment for Dresden and Quad Cities Station's
Technical Specification Section 3/4.5 is based on STS guidelines or
later operating BWR plants' NRC accepted changes. The proposed
amendment has been reviewed for acceptability at the Dresden and
Quad Cities Nuclear Power Stations considering similarity of system
or component design versus the STS or later operating BWRs. Any
deviations from STS requirements do not create the possibility of a
new or different kind of accident previously evaluated for Dresden
or Quad Cities Stations. No new modes of operation are introduced by
the proposed changes. Surveillance requirements are changed to
reflect improvements in technique, frequency of performance or
operating experience at later plants. Proposed changes to action
statements in many places add requirements that are not in the
present technical specifications. The proposed changes maintain at
least the present level of operability. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
The associated systems that make up the Emergency Core Cooling
Systems are not assumed in any safety analysis to initiate any
accident sequence for Dresden or Quad Cities Stations. In addition,
the proposed surveillance requirements for affected systems
associated with the Emergency Core Cooling Systems are generally
more prescriptive than the current requirements specified within the
Technical Specifications; therefore, the proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
Involve a significant reduction in the margin of safety because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, the addition of
requirements which are based on the current safety analysis, and
some minor curtailments of the current requirements which are based
on generic guidance or previously approved provisions for other
stations. Some of the latter individual items may introduce minor
reductions in the margin of safety when compared to the current
requirements. However, other individual changes are the adoption of
new requirements which will provide significant enhancement of the
reliability of the equipment assumed to operate in the safety
analysis, or provide enhanced assurance that specified parameters
remain with their acceptance limits. These enhancements compensate
for the individual minor reductions, such that taken together, the
proposed changes will not significantly reduce the margin of safety.
The proposed amendment to Technical Specification Section 3/4.5
implements present requirements, or the intent of present
requirements in accordance with the guidelines set forth in the STS.
Any deviations from STS requirements do not significantly reduce the
margin of safety for Dresden or Quad Cities Stations. The proposed
changes are intended to improve readability, usability, and the
understanding of technical specification requirements while
maintaining acceptable levels of safe operation. The proposed
changes have been evaluated and found to be acceptable for use at
Dresden or Quad Cities based on system design, safety analysis
requirements and operational performance. Since the proposed changes
are based on NRC accepted provisions at other operating plants that
are applicable at Dresden or Quad Cities and maintain necessary
levels of system or component reliability, the proposed changes do
not involve a significant reduction in the margin of safety.
The proposed amendment for Dresden and Quad Cities Stations will
not reduce the availability of systems associated with the Emergency
Core Cooling Systems when required to mitigate accident conditions;
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 10 CFR 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: for Dresden, Morris Public
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra

[[Page 42601]]

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: June 17, 1993, as supplemented July 5,
1995
Description of amendment request: The initial proposed amendment
request dated June 17, 1993, was previously noticed in the Federal
Register on July 21, 1993 (58 FR 39048). The proposed amendment would
revise Technical Specification 5.3.1, ``Fuel Assemblies'' to provide
flexibility in the repair of fuel assemblies containing damaged and
leaking fuel rods by reconstituting the assemblies in accordance with
the guidance in Generic Letter (GL) 90-02, Supplement 1, ``Alternative
Requirements For Fuel Assemblies In The Design Features Section Of
Technical Specifications,'' issued on July 31, 1992. The application is
also generally consistent with the format and content of the improved
Standard Technical Specifications for Westinghouse plants provided in
NUREG-1431.
Additional information was submitted on July 5, 1995, that added TS
changes to increase the fuel enrichment limit from 4.0 to 5.0 weight
percent U-235 that were not previously included the initial June 17,
1993, amendment application. This additional information is being
noticed to provide for public comment and opportunity for hearing.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee provided its analysis of the issue of no significant hazards
consideration (58 FR 39048). The NRC staff's analysis of the July 5,
1995, supplement against the standards of 10 CFR 50.92(c) is presented
below.
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
There is no increase in the probability or consequences of an
accident in the new fuel vault since the only accident that would be
affected by this change would be a criticality accident and it has been
shown that the worst-case keff under optimum moderation conditions
continues to be less than or equal to 0.98.
There is no increase in the probability of a fuel drop accident in
the Spent Fuel Storage Pool since the mass of an assembly will not be
significantly affected by the increase in fuel enrichment. The
likelihood of other accidents, previously evaluated and described in
Section 9.1.2 of the Final Safety Analysis Report (FSAR), is also not
affected by the proposed changes. Since the increase in fuel enrichment
will allow for extended fuel cycles, it could be postulated that there
may be a decrease in fuel movement and the probability of an accident
may likewise be decreased. There is also no increase in the
consequences of a fuel drop accident in the Spent Fuel Pool since the
fission product inventory of individual fuel assemblies will not change
significantly as a result of increased initial enrichment. In addition,
no change to safety-related systems is being made.
Therefore, the consequences of a fuel rupture accident remain
unchanged. In addition, it has been shown that keff is less than
or equal to 0.95, under all conditions. Therefore, the consequences of
a criticality accident in the Spent Fuel Pool remain unchanged as well.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident since fuel handling accidents (fuel drop and
misplacement) are not new or different kinds of accidents. Fuel
handling accidents are already discussed in the FSAR for fuel with
enrichments up to 4.0 weight % and additional analyses have been
performed for fuel with enrichment up to 5.00 weight %.
3.
The proposed changes do not involve a significant reduction in the
margin of safety.
The proposed change does not involve a significant reduction in the
margin of safety since, in all cases, a spent fuel pool keff less
than or equal to 0.95 is being maintained. Criticality analyses have
also been performed that show that the new fuel storage vault will
remain subcritical under a variety of moderation conditions, from fully
flooded to optimum moderation. As discussed above, the Spent Fuel Pool
will remain sufficiently subcritical during any fuel misplacement
accident.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the supplemental amendment submittal involves no
significant hazards consideration.
Local Public Document Room location:: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina

Date of amendment request: July 26, 1995
Description of amendment request: The proposed amendments would
provide a one-time extension of the allowable outage time from 72 hours
to 7 days. This extension is necessary to implement a modification to
the degraded grid protection system and the external grid trouble
protection system.
Basis for proposed no significant hazards
considerationdetermination: 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:

Duke Power Company (Duke) has made the determination that this
amendment request involves a No Significant Hazards Consideration
by applying the standards established by NRC regulations in 10 CFR
50.92. This ensures 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:

Each accident analysis addressed within the Oconee Final Safety
Analysis Report (FSAR) has been examined with respect to the change
proposed within this amendment request. The design basis of the
auxiliary electrical systems is to supply the required engineered
safeguards (ES) loads of one unit and the safe shutdown loads of the
other two units. The systems are arranged so that no single failure
will jeopardize plant safety.
The probability of any Design Basis Accident (DBA) is not
significantly increased by this change. In addition, the
consequences of the accidents are within the bounds of the FSAR
analyses. The reliability of the emergency power system is not
significantly affected by a one time extension of allowable outage
time for the overhead power path. The underground power path is
adequate to assure operability of the Oconee ES loads. Finally, the
enhancement of the Degraded [Grid] Protection System will eliminate
a concern which was expressed by the EDSFI audit team.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
Inoperability of the yellow bus is functionally equivalent to
inoperability of the Keowee Main Step-up Transformer in that it
renders the overhead emergency power path inoperable. The Keowee
Main Step-up Transformer is allowed to be inoperable for a period
not to exceed 28 days. This Technical Specification requirement for
the

[[Page 42602]]
Keowee Main Step-up Transformer has been reviewed and approved by the
NRC. Therefore, operation of ONS [Oconee Nuclear Station] in
accordance with this Technical Specification amendment will not
create any failure modes not bounded by previously evaluated
accidents. Consequently, this change will not create the possibility
of a new or different kind of accident from any kind of accident
previously evaluated.
(3) Involve a significant reduction in a margin of safety:
The design basis of auxiliary electrical systems is to supply
the required ES loads of one Unit and safe shutdown loads of the
other two units. The underground power path is adequate to ensure
operability of the ES loads during the outage of the yellow bus. The
reliability of the emergency power system is not significantly
affected by a one time extension of allowable outage time for the
overhead power path. Therefore, there will be no significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania

Date of amendment request: July 10, 1995
Description of amendment request: The proposed amendment would
modify the technical specifications to minimize the potential for boron
deletion of the reactor coolant system (RCS) during startup of an
isolated loop. The changes would permit RCS loop isolation only during
Modes 5 and 6. RCS loop isolation valves would be required open with
power removed from each isolation valve operator during Modes 1, 2, 3,
and 4. Primary grade water would be isolated from the RCS during Modes
4, 5, and 6, except during planned boron dilution or makeup activities.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment would modify the method used to prevent
an inadvertent boron dilution event during hot shutdown, cold
shutdown and during refueling. An uncontrolled boron dilution
transient cannot occur during this mode of operation. Inadvertent
boron dilution is prevented by administrative controls which isolate
the primary grade water system isolation valves from the Chemical
and Volume Control System, except during planned boron dilution or
makeup activities. Thus unborated water can not be injected into the
reactor coolant system, making an unplanned boron dilution at these
conditions highly improbable, since the source of unborated water to
the charging pumps is isolated. This precludes the primary means for
an inadvertent boron dilution event in this mode of operation.
The primary grade water system isolation valves may be opened
when directed by the control room during this mode of operation only
for a planned boron dilution or makeup activity. The primary grade
water system isolation valves will be verified to be locked, sealed
or otherwise secured in the closed position after the planned boron
dilution or makeup activity is completed. During planned boron
dilution events, operator attention will be focused on the boron
dilution process and any inappropriate blender operation will be
readily identified.
The operator has prompt and definite indication of any boron
dilution from the audible count rate instrumentation supplied by the
source range nuclear instrumentation. High count rate is alarmed in
the reactor containment and the control room. In addition a high
source range flux level is alarmed in the control room. The count
rate increase is proportional to the subcritical multiplication
factor.
The proposed amendment would also modify the method used to
prevent an adverse reactor transient during startup of an isolated
reactor coolant loop. Procedures require that the isolated loop
water boron concentration be verified prior to opening loop
isolation valves. Procedures also require an isolated loop to be
drained and refilled from water supplied from the Refueling Water
Storage Tank (RWST) or Reactor Coolant System (RCS) prior to opening
either the hot or cold leg isolation valves. Using water from the
RWST or RCS ensures 1) that the boron concentration of the isolated
loop is sufficient to prevent a dilution of the active reactor
coolant loops and reducing the shutdown margin to below those values
used in safety analyses when the isolated loop is returned to
service, and 2) that no single failure could cause an isolated loop
to be filled with unborated water.
Thus procedures and interlocks prevent inadvertent opening of
loop isolation valves and require that the startup of an isolated
loop be performed in a controlled manner that virtually eliminates
any sudden positive reactivity addition from boron dilution. Thus
the core cannot be adversely affected by the startup of an isolated
loop and fuel design limits are not exceeded. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not create the possibility of a new or
different kind of accident. No new systems, structures or components
are being proposed. Acceptable alternative administrative controls
are being proposed to address inadvertent boron dilution and the
startup of inactive reactor coolant loops.
The primary source of unborated water will be isolated from
injecting by the charging pumps into the reactor coolant system
during hot shutdown, cold shutdown, and refueling, except for
planned boron dilution events and makeup activities. The proposed
administrative controls prevent the possible accident previously
evaluated, i.e., an inadvertent boron dilution event.
A currently installed interlock to recirculate reactor coolant
in an isolated loop is proposed to be deleted. In its place, each
reactor coolant isolated loop will be drained and refilled with
water supplied from the RWST just before the loop is returned to
service. This administrative control will prevent any inadvertent
reactivity transient when returning the loop to service. Thus, the
proposed administrative controls will prevent the type of accident
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes will continue to ensure that adequate
protection is provided against an inadvertent boron dilution and the
adverse effects from the startup of an isolated reactor coolant
loop. General Design Criteria 10 requirements will not be exceeded
with respect to demonstrating specified acceptable fuel design
limits. The required indications and functions are still maintained
in accordance with current technical specification requirements and
the shutdown margin is unaffected. Therefore, the proposed change
will not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. Library, 663 Franklin
Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz

[[Page 42603]]

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley
Power Station, Unit 1, Shippingport, Pennsylvania

Date of amendment request: July 11, 1995
Description of amendment request: The proposed amendment would
revise the required area of the Reactor Coolant System (RCS)
overpressure protection system vent from 3.14 square inches to 2.07
square inches. This vent is provided to relieve a potential RCS
overpressure condition if the power-operated relief valves (PORVs) are
not operable. The proposed vent area is equal to the relief area of a
PORV. A single PORV is capable of providing sufficient relief capacity
to mitigate potential low temperature overpressurization events.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change is considered to be editorial since it
replaces the 3.14 square inch vent size stated in overpressure
protection system (OPPS) Specifications 3.4.9.3, 3.1.2.1.b, and
3.1.2.3 and Bases 3/4.1.2 and 3/4.4.9 with a 2.07 square inch vent
size. This ensures the vent size stated in the technical
specifications is consistent with the actual size of an installed
PORV. These changes maintain consistency with the analyses
assumptions and the operation of the OPPS in accordance with
applicable analyses and the UFSAR [Updated Final Safety Analyses
Report]. Therefore, we have concluded that these changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated in the UFSAR.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not involve any physical changes to the
OPPS or their setpoints. These changes do not change any function
previously provided by the OPPS. These changes do not affect any
failure modes defined for any plant system or component important to
safety nor has any new limiting single failure been identified as a
result of these changes. Therefore, these changes will not create
the possibility of a new or different kind of accident from any
accident previously evaluated in the UFSAR.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes will not affect the operation of or the
reliability of the OPPS. These changes do not affect the manner in
which the plant is operated or involve a change to equipment or
features that affect the operational characteristics of the plant.
Therefore, operation of the plant in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania

Date of amendment request: July 20, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.8.1.1 to incorporate guidance
provided in NRC Generic Letter (GL) 84-15, ``Proposed Staff Actions to
Improve and Maintain Diesel Generator Reliability,'' and GL 93-05,
``Line-Item Technical Specification Improvements To Reduce Surveillance
Requirements For Testing During Power Operation,'' which includes (1)
revised requirements for testing the operable emergency diesel
generators (EDGs) for various combinations of inoperable offsite
circuits and EDGs and (2) revised surveillance requirements for the
EDGs. The revised surveillance requirements include specifying
generator voltage, frequency limits, and diesel starting time. In
addition, several editorial changes would be made to TS 3/4.8.1.1 which
would be consistent with the guidance provided in the NRC's Improved
Standard Technical Specifications (NUREG-1431).
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The probability of occurrence of a previously evaluated accident
is not increased because the allowable outage times for the offsite
circuits and diesel generators remain unchanged. The consequences of
an accident previously evaluated is not increased because reducing
the diesel generator test frequency and permitting additional test
evolutions are intended to minimize diesel wear and mechanical
stress. By eliminating excessive testing, which can lead to
premature diesel failures and minimizing diesel wear and mechanical
stress, the diesel generator reliability is increased. The
consequences of an accident previously evaluated is also not
increased because the addition of the parameters for generator
voltage, frequency, and diesel starting time to the surveillance
requirement will provide additional assurance that the diesel
generators are performing as assumed in the safety analysis. This
proposed change does not affect the availability or reliability of
the offsite circuits.
Therefore, this change will not increase the probability or
consequences of an accident previously evaluated due to the
continued availability and reliability of the A.C. electrical power
sources.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not alter the method of operating the
plant. The changes do not introduce any new failure modes and are
intended to increase the diesel generator reliability and provide
additional assurance that the diesels are performing as assumed in
the safety analysis. The revision to the various action statements
and surveillance requirements provide assurance that the diesel
generators will be able to power their respective safety systems if
required. The proposed changes do not impact the performance of any
safety system.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The margin of safety is not reduced because the A.C. electrical
power sources will continue to provide sufficient capacity,
capability, redundancy, and reliability to ensure availability of
necessary power to engineered safety feature (ESF) systems. The ESF
systems will continue to function, as assumed in the safety
analyses, to ensure that the fuel, reactor coolant system and
containment design limits are not exceeded. The elimination of
excessive testing on the diesel generators are permitting additional
test evolutions, which result in less diesel wear and mechanical
stress, are intended to increase diesel reliability. The increased
reliability of the diesels adds to the ability of the A.C.
electrical power source to provide power to ESF systems. The
proposed additions to the surveillance requirements will provide
additional assurance of the ability of the A.C. electrical power
sources to provide power to ESF systems.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the

[[Page 42604]]
amendment request involves no significant hazards consideration.
Local Public Document Room Location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley
PowerStation, Unit 2, Shippingport, Pennsylvania

Date of amendment request: July 24, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.4.11, ``Relief Valves,'' and
associated Bases to make Unit 2 TS 3/4.4.11 consistent with Unit 1 TS
3/4.4.11, which was revised by Unit 1 License Amendment No. 187 issued
on May 15, 1995. The proposed amendment would also generally reflect
the guidance provided in NRC Generic Letter 90-06 and in the NRC's
Improved Standard Technical Specifications (NUREG-1431).
Basis for proposed no significant hazards
considerationdetermination: 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?
Implementation of these changes will increase the availability
of the power-operated relief valves (PORVs) and their associated
block valves. The increased availability is obtained through
maintaining power to the block valves which are closed to control
PORV seat leakage. Maintaining power to the block valve provides the
flexibility of reopening the valves to control reactor coolant
system pressure. The proposed change modifies Specification 3.4.11
actions, a surveillance requirement, and Bases to generally reflect
the requirements of Generic Letter (GL) 90-06, and the guidance
provided in NUREG-1431, ``Improved Standard Technical
Specifications'' (ISTS) and is consistent with the changes the NRC
approved for Unit No. 1. A revised stress analysis has been
completed that takes credit for the speed at which the block valve
opens when manually reducing reactor coolant system pressure. The
block valve relatively slow opening speed reduces the peak pressure
surge and results in acceptable downstream piping stress values. The
PORV downstream piping has been evaluated assuming manual vent path
operation with cold loop seal slug flow and it has been determined
that the piping supports can accept these design transient loads.
The proposed change to the action statement to close the block valve
to isolate a PORV and maintain power to the block valve does not
significantly increase the probability of a small break loss of
coolant accident. No PORV function has been deleted and the PORV and
block valve continue to be capable of being manually closed at any
time. As a result of the change to action ``a,'' an exception to the
stroking requirements is no longer required, therefore, reference to
action ``a'' in Surveillance Requirement 4.4.11.2 has been deleted.
Closing the block valve for a PORV that is not capable of being
manually cycled and removing power to the block valve assures that
the valve will not be inadvertently opened when the condition of the
PORV is uncertain.
The changes remain consistent with the analysis assumptions
regarding the operation of the PORVs and block valves and provides
increased assurance of their availability in mitigating the
consequences of a steam generator tube rupture (SGTR) accident. The
requirements of GL 90-06 are substantially addressed in the ISTS
which have been incorporated here except for specific design
differences. Minor editorial changes involving capitalization have
been incorporated to maintain the format and content and do not
affect any of the requirements, the accident analyses, or the
operation of the plant. Therefore, we have concluded that these
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated in the UFSAR
[Updated Final Safety Analysis Report].
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes to the action statements for the PORVs and
the associated block valves will improve the availability of these
valves for normal operation and for mitigation of a SGTR accident.
The proposed changes do not involve any physical changes to the
PORVs or their setpoints. These changes do not delete any design
basis accident function previously provided by the PORV vent path
nor has the probability of inadvertent opening been increased.
Accordingly, no new limiting single failure has been identified as a
result of these changes. Therefore, these changes will not create
the possibility of a new or different kind of accident from any
accident previously evaluated in the UFSAR.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes have been incorporated to provide the
capability to manually stroke the vent path using the block valve to
control the pressure surge as a PORV opens. The resultant downstream
piping forces were found acceptable, therefore, power can be
maintained to the block valve when the block valve has been closed
to isolate a PORV because of excessive seat leakage. This will allow
operation of the PORVs in a manner similar to the guidance provided
in GL 90-06 to improve PORV availability. These changes will improve
the operator use of an isolated PORV since it is now analyzed to be
manually cycled with the block valve closed and power maintained so
the operator can use the PORV if required to mitigate the effects of
a SGTR accident. This is consistent with the intent of the ISTS and
does not affect the UFSAR, therefore, operation of the plant in
accordance with the proposed amendment would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 1500l.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas

Date of amendment request: April 4, 1995
Description of amendment request: The proposed amendment revises
the minimum water level that is required to be maintained over
irradiated fuel assemblies during latching and unlatching of control
element assemblies.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The fuel handling accident analysis assumes that a fuel assembly
is dropped during fuel handling. During the latching and unlatching
of the CEAs, the upper guide structure is in place and the CEDM
extension shaft assemblies are disconnected from their CEA for
subsequent removal with the vessel upper guide structure. The
dropping of a CEA from the maximum height of six inches will not
damage that particular fuel assembly or any surrounding fuel
assemblies since this movement is confined to within the upper guide
structure and the guide tubes of the associated fuel assembly during
this activity. This less than six inches of movement does not have
the potential to result in a fuel handling accident; therefore, an
increase in the probability of this accident does not occur. The
requirement to have at least 23 feet of water over the top of the
irradiated fuel assemblies during fuel and CEA movement ensures
that, should a fuel handling accident occur, the resulting offsite
dose consequences are mitigated. The six inch movement of the CEA
during CEA decoupling does not constitute fuel or CEA

[[Page 42605]]
movement which would result in a fuel handling accident. As such,
Technical Specifications are unchanged with respect to the
mitigating requirements for a fuel handling accident.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change does not change the design, configuration,
or method of operation of the plant; therefore, it does not create
the possibility of a new or different kind of accident. Because no
new equipment is being introduced, and no equipment is being
operated in a manner inconsistent with its design, the possibility
of equipment malfunction is not increased. The proposed change adds
an exception to the applicability section and is bounded by the
existing fuel handling accident analysis.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
There is no reduction in margin of safety in that 23 feet of
water is still maintained over the irradiated fuel assemblies
anytime there is a potential for a fuel handling accident. Adding
the exception of the latching and unlatching of the CEAs to the
applicability section does not involve a change in the accident
analysis for fuel handling which remains bounding.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: July 21, 1995
Description of amendment request: The proposed change requests that
the current expiration date for license NPF-29 be changed to reflect
the issuance date of the new license granted Grand Gulf on November 1,
1984. The change consists of extending the expiration date to 40 years
from the date of issuance of license NPF-29 (November 1, 1984 to
November 1, 2024).
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
a. No significant increase in the probability or consequences of
an accident previously evaluated results from this change.
The proposed change does not affect the design or operation of
any plant system. The effect of 40 years of full power operations
has previously been evaluated and documented in the Updated Final
Safety Analysis Report (UFSAR). The design life of structures,
systems and components is controlled by existing plant problems
[sic., programs] and processes that are not affected by this change.
The proposed change will simply allow Grand Gulf to achieve its
original planned 40 years of service. Equipment associated with
initiating event frequencies or accident mitigation must continue to
meet all applicable maintenance and operability requirements
regardless of license duration (It is also interesting to note that
the license duration limitation of 40 years, as contained in 10 CFR
50.51 is not a limitation resulting from concerns over plant aging
effects. ``In fact, the limit was a compromise between the efforts
of the Justice Department and electric cooperatives, who championed
a 20-year limit on the basis of antitrust concerns, and the view of
the utility industries that a longer period was necessary to ensure
full amortization of a nuclear power plant.'' (56 FR 64961, December
13, 1991)). Therefore, the probability or consequences of previously
analyzed accidents are not significantly increased.
b. The change would not create the possibility of a new or
different kind of accident from any previously analyzed.
The proposed change will not add any plant equipment or
introduce any new modes of plant operation. The change will only
amend the operating license to allow 40 years of full power
operations. The proposed change does not affect the current
maintenance or surveillance practices, which are designed to
maintain and monitor the current service life of plant structures,
systems and components in accordance with regulatory requirements.
Therefore, the proposed change does not create the possibility of
new equipment failure modes or a new or different kind of accident
from any accident previously evaluated.
c. The change would not involve a significant reduction in a
margin of safety.
The proposed change does not involve a significant reduction in
a margin of safety since it only provides for 40 years of full power
operations for which the plant is designed. Current Technical
Specification surveillance requirements (e.g. associated with 10 CFR
50 Appendix H) and other regulatory requirements remain in place and
will ensure continued compliance with applicable safety margins.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania

Date of amendment request: June 20, 1995
Description of amendment request: The proposed Technical
Specifications (TS) changes would remove the surveillance interval text
for the 10 CFR Part 50, Appendix J, Type A test (Integrated Leak Rate
Test or ILRT), and Drywell-to-Suppression Chamber (bypass) leakage test
specified in TS Surveillance Requirements (SR) 4.6.1.2.a, 4.6.1.2.b,
and 4.6.2.1.e.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The primary containment and the suppression chamber are not
considered to be accident initiators, they are accident mitigators.
There are no physical or operational changes to the containment or
suppression structure, system or components being made as a result
of the proposed changes. These changes will not impose different
requirements and adequate control of information will be maintained.
These TS changes will not alter assumptions made in the safety
analysis and licensing basis. Therefore, the proposed TS changes to
eliminate the details of the test intervals will not increase the
probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes remove the specific surveillance test
interval text from TS and address the interval by direct reference
to the applicable regulation. The proposed TS changes do not make
any physical or operational changes to existing plant systems or
components. Furthermore, the primary containment and suppression
chamber act as

[[Page 42606]]
accident mitigators not initiators. Therefore, the possibility of a new
or different kind of accident than from any accident previously
evaluated is not introduced.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
LGS [Limerick Generating Station] TS Bases 3/4 6.1.2 state that
surveillance testing is consistent with 10 CFR 50, Appendix J and
does not specify a SR test interval. TS Bases 3/4 6.2, describing
the bypass test does not specify a SR test interval. However, the
NRC Safety Evaluation related to amendment Nos. 68 (Unit 1) and 31
(Unit 2) concluded that it is acceptable for the drywell-to-
suppression chamber test frequency to coincide with the 10 CFR 50,
Appendix J, Type A test, since individual vacuum breaker leakage
tests are an acceptable alternative to an integrated suppression
pool bypass test during outages for which a Type A containment
integrated leak rate test is not conducted. The alternative bypass
test requirement, TS SR 4.6.2.1.f, is not affected by these changes.
The Type A test, and bypass SR test intervals are adequately
presented in the test implementing procedures, and TS will directly
reference 10 CFR 50, Appendix J, for the appropriate test interval.
Therefore, the proposed TS changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz

Power Authority of The State of New York, Docket No. 50-286, Indian
PointNuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: July 21, 1995
Description of amendment request: The proposed amendment would
change Technical Specifications Section 6.0 (Administrative Controls)
to replace the title-specific list of members on the Plant Operating
Review Committee (PORC) with a more general statement of membership
requirements. The scope of disciplines represented on the PORC would
also be expanded to include nuclear licensing and quality assurance.
The proposed amendment would also change the title ``Resident Manager''
to ``Site Executive Officer.'' This title change would not affect the
reporting relationship, authority, or responsibility of the position.
Basis for proposed no significant hazards
considerationdetermination: 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:
Operation of the Indian Point 3 Nuclear Power Plant in
accordance with the proposed amendment would not involve a
significant hazards consideration as defined in 10 CFR 50.92, since
it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes are administrative in nature and do not
involve plant equipment or operating parameters. There is no change
to any accident analysis assumptions or other conditions which could
affect previously evaluated accidents. The proposed changes will not
decrease the organization's ability to respond to a design basis
accident.
2. Create the possibility of a new or different kind of accident
from those previously evaluated.
Since the proposed changes are administrative in nature and do
not involve hardware design, modifications or operation, the
possibility of new or different accidents is not created.
3. Involve a significant reduction in the margin of safety.
The proposed title change for the Resident Manager is an
administrative change and does not affect the responsibilities,
authority, or reporting relationships for this management position.
Replacing the title specific list of PORC members with a statement
of membership requirements for the committee does not reduce the
effectiveness of the committee to advise the Resident Manager (Site
Executive Officer) on matters regarding nuclear safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Ledyard B. Marsh

Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 30, 1995
Description of amendment request: The proposed change to the
Technical Specifications (TS) would change TS Table 3.3.1-2, ``Reactor
Protection System Response Times'', TS Table 3.3.2-3, ``Isolation
System Instrumentation Response Time'', TS Table 3.3.3-3, ``Emergency
Core Cooling System Response Times'', and associated Bases. The
proposed changes to the above-referenced TS Tables would eliminate the
requirement to perform response time testing for certain classes of
equipment.
Basis for proposed no significant hazards
considerationdetermination: 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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The purpose of the proposed Technical Specification change is to
eliminate response time testing requirements for selected
instrumentation in the Reactor Protection System, Isolation System,
and Emergency Core Cooling System. However, because of the continued
application of other existing Technical Specification requirements
such as channel calibrations, channel checks, channel functional
tests, and logic system functional tests, the response time of these
systems will be maintained within the acceptance limits assumed in
plant safety analyses and required for successful mitigation of an
initiating event. The proposed Technical Specification changes do
not affect the capability of the associated systems to perform their
intended function within their required response time.
The BWR Owners' Group has completed an evaluation (NEDO-32291,
``System Analyses for the Elimination of Selected Response Time
Testing Requirements'') which demonstrates that response time
testing is redundant to the other Technical Specification
requirements listed in the preceding paragraph. These other tests
are sufficient to identify failure modes or degradation in
instruments response time and ensure operation of the associated
systems within acceptance limits. There are no known failure modes
that can be detected by response time testing that cannot be
detected by the other Technical Specification tests. Hope Creek
Generating Station is specifically bounded by the assumptions and
justifications in General Electric Company Licensing Topical Report,
NEDO-32291, ``System Analyses for Elimination of Selected Response
Time Testing Requirements.''
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
As discussed above, the proposed Technical Specification changes
do not affect the capability of the associated systems to perform
their intended function within the acceptance limits assumed in
plant safety analyses and required for successful mitigation of an
initiating event. The proposed elimination of response time testing
would not result in any new

[[Page 42607]]
equipment, operating modes, or plant configurations.
3. Will not involve a significant reduction in a margin of
safety.
The current Technical Specification response times are based on
the maximum allowable values assumed in the plant safety analyses.
These analyses conservatively establish the margin of safety. As
described above, the proposed Technical Specification changes do not
affect the capability of the associated systems to perform their
intended functions within the allowed response time used as the
basis for the plant safety analyses. Plant and system response to an
initiating event will remain in compliance within the assumptions of
the safety analyses, and therefore the margin of safety is not
affected.
Although not explicitly evaluated, the proposed Technical
Specification changes will provide an improvement to plant safety
and operation by:
a) Reducing the time safety systems are unavailable
b) Reducing safety system actuations
c) Reducing shutdown risk
d) Limiting radiation exposure to plant personnel
e) Eliminating the diversion of key personnel to conduct
unnecessary testing.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey

Date of amendment request: April 18, 1995
Description of amendment request: The proposed changes to the
Technical Specifications (TS) would change TS Table 4.3.7.1-1
``Radiation Monitoring Instrumentation Surveillance Requirements.''
This change would increase the channel functional test interval from
monthly to quarterly for each instrument.
Basis for proposed no significant hazards
considerationdetermination: 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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change involves no hardware changes, no changes to
the operation of any systems or components, and no changes to
existing structures. Increasing the interval between channel
functional tests for the radiation monitoring instrumentation
represent changes that do not affect plant safety and do not alter
existing accident analyses.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed change is procedural in nature concerning the
channel functional test frequency for the radiation monitoring
instrumentation not already on a quarterly surveillance. The channel
functional test methodology for these instruments remains unchanged.
The proposed changes, while slightly increasing the possibility of
an undetected instrument error, will not create a new or unevaluated
accident or operating condition.
3. Will not involve a significant reduction in a margin of
safety.
The proposed change is in accordance with recommendations
provided by the NRC regarding the improvement of Technical
Specifications. These changes will result in perpetuation of current
safety margins while reducing regulatory burden and decreasing
equipment degradation.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey

Date of amendment request: May 4, 1995
Description of amendment request: The proposed change to the
Technical Specifications (TS) would change TS 3/4.6.1.8, ``Drywell and
Suppression Chamber Purge System'', to increase the annual operational
limit for the drywell and suppression chamber purge system from 120 to
500 hours.
Basis for proposed no significant hazards
considerationdetermination: 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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change involves no hardware changes and no changes
to existing structures. Increasing the annual operational limit of
the drywell and suppression chamber purge system will not increase
the probability of a loss-of-coolant accident. While increased usage
of the purge system will result in a slight increase in the
possibility that these valves will be open during a LOCA, it will
not alter or impact previous LOCA analyses.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed change will not result in an unanalyzed condition.
While the increase in purge system operation will slightly increase
the possibility of the containment vent and purge valves being open
at the onset of a LOCA event, the valves have been established as
capable of isolating the containment within five seconds. This is
well within the bounds of existing LOCA analyses which assume an
open duration of 175 seconds. Therefore, this change will not
require a new or different accident analysis.
3. Will not involve a significant reduction in a margin of
safety.
The proposed change will not alter existing systems, equipment,
components, or structures. The method of operating the drywell and
suppression chamber purge system will not be altered by the
increased annual usage. While there is a slight increase in the
possibility of purge operations at the onset of a LOCA, any
resulting release would be insignificant and bounded by existing
LOCA analyses. Operation of the drywell and suppression chamber
purge system based on these proposed changes will remain within the
guidance provided in the NRC's Branch Technical Position CSB 6-4.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Pennsville Public Library,
190 S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz

Saxton Nuclear Experimental Corporation (SNEC), Docket No. 50-146,
Saxton Nuclear Experimental Facility (SNEF), Bedford County,
Pennsylvania

Date of amendment request: June 2, 1995, as supplemented on June
23, 1995.
Description of amendment request: The proposed changes to the
technical specifications are administrative in

[[Page 42608]]
nature. The proposed amendment would revise the organization structure
associated with the SNEF to allow General Public Utilities Nuclear
Corporation resources to be applied to SNEC activities within their
normal organizational structure; eliminating the need to identify and
compartmentalize a portion of the organization as specific to SNEC. The
proposed amendment would also revise the description and drawing of the
SNEF site to reflect multiple gates in the SNEF fence.
Basis for proposed no significant hazards
considerationDetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below: The proposed changes
do not involve a significant hazards considerations because the changes
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The administrative changes will not impact the physical
condition of the containment vessel as it relates to the risk of
fire, flood or radiological hazard.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
In its present condition, the only accidents applicable to the
site are those addressed above.
3. Involve a significant reduction in a margin of safety.
The proposed administrative changes would have no effect on any
margins of safety for any evaluated accidents.
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: Saxton Community Library, 911
Church Street, Saxton, Pennsylvania 16678Attorney for the Licensee:
Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge,
2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: Seymour H. Weiss

South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: June 30, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) for the pressurizer power
operated relief valves (PORVs) to follow the guidance of Generic Letter
(GL) 90-06, Generic Issue 70, and the improved Westinghouse
Standardized Technical Specifications (NUREG-1431, Rev. 1).
Basis for proposed no significant hazards
considerationdetermination: 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 probability or consequences of an accident previously
evaluated is not significantly increased.
There is no increase in the probability of an accident because
the physical characteristics of the PORVs and their block valves
remain unchanged. No changes to any hardware or software that
affects these components is planned.
The PORVs are pressure relieving devices and only two failure
modes need to be considered. The first is that one or more PORVs or
block valves fail to open when required. This is not
a significant concern and is not a credible cause of any
accident. The second mode is failing to close which includes
depressurization of the RCS [reactor coolant system] and a reactor
trip on low pressurizer pressure or overtemperature [delta]T. The
consequences for the more limiting Pressurizer Safety Valve
Accidental Depressurization event has been analyzed with acceptable
results.
There is no increase in the consequences of an accident as a
result of this change, because only one PORV is required to mitigate
the consequences of a design basis Steam Generator Tube Rupture.
There is sufficient redundancy to ensure one PORV is available to
perform this function even if one PORV is inoperable or incapable of
being manually cycled. The validation of the Emergency Operating
Procedures on the VCSNS [Virgil C. Summer Nuclear Station] simulator
demonstrated that one pressurizer PORV has sufficient capacity to
depressurize the RCS in a time frame which will not cause the
offsite doses presented in the FSAR [Final Safety Analysis Report]
to be exceeded.
The PORVs are utilized to depressurize the RCS and equalize the
pressure between the primary and secondary systems. This stops the
intrusion of RCS water into the secondary which can be released into
the atmosphere. By the time the PORVs are called upon, the affected
steam generator (SG) has been identified and steps have been taken
to isolate the faulted SG. This acts to minimize the radiological
impact on the health and safety of the public. In all cases, the
dose results are within 10 CFR 100 limits.
2. The possibility of an accident or a malfunction of a
different type than any previously evaluated is not created.
The proposed TSCR [TS Change Request] does not involve any
physical changes to the plant or decrease the number of PORVs and
block valves that must be capable of performing their intended
function. These components are used to mitigate the effects of
postulated events and their failure has already been considered. The
worst case failure, either not opening or not closing, has been
evaluated and is bounded by other more limiting accidents.
3. The margin of safety has not been significantly reduced.
The currently approved TS permits all three PORVs and/or their
block valves to be inoperable as long as precautions are taken to
assure that RCS would not leak-by, assuming single failures and
spurious operation. The proposed TSCR would require a minimum of two
PORVs and block valves to be operable, or at least capable of being
manually cycled, in Modes 1, 2, and 3. This is in fact an increase
in margin and provides for greater reliability with the added
benefit that the probability of challenges to the pressurizer code
safety valves will be lessened.
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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: Frederick J. Hebdon

South Carolina Electric & Gas Company (SCE&G), South Carolina
Public Service Authority, Docket No. 50-395, Virgil C. Summer
Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: July 28, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to exclude the requirement to
perform the slave relay test of the 36-inch containment purge supply
and exhaust valves on a quarterly basis while the plant is in Modes 1,
2, 3, or 4.
Basis for proposed no significant hazards
considerationdetermination: 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?
No, the probability or consequences of an accident previously
evaluated would not be increased since no credit is taken for the
valves in FSAR [Final Safety Analysis Report] Chapter 15.
The only credible accident discussed in FSAR Chapter 15 that
applies to these valves is a fuel handling accident inside

[[Page 42609]]
containment (15.4.5.1). The analysis assumes the escaped gases are
released instantaneously to the environment via the Reactor
Building purge system. The analysis does not take credit for
these valves nor for filtration or holdup time during release. The
result of the analysis is acceptable and offsite doses are within
the limits of 10 CFR 100.
TS 3.6.1.7 requires that these valves be sealed shut during
Modes 1, 2, 3, and 4. When sealed shut, these valves will not open
via any signal.
With these valves already in a shut position, neither the
probability nor the consequences of an accident are increased.
2. Does the change create the possibility of a new or different
kind of accident from any previously evaluated?
No, the 36'' [inch] containment purge exhaust and supply valves
will not be placed in a condition different from that evaluated
previously.
The only credible accident discussed in FSAR Chapter 15 that
applies to these valves is a fuel handling accident inside
containment (15.4.5.1). The analysis assumes the escaped gases are
released instantaneously to the environment via the Reactor Building
purge system. The analysis does not take credit for these valves nor
for filtration or holdup time during release. The result of the
analysis is acceptable and offsite doses are within the limits of 10
CFR 100.
Additionally, TS 3.6.1.7. requires that these valves be sealed
shut during Modes 1, 2, 3, and 4. When sealed shut, these valves
will not open via any signal.
3. Does the change involve a significant reduction in the margin
of safety?
TS 4.3.2.1. requires that this slave relay test be performed
quarterly. This surveillance is accomplished for the 36'' [inch]
containment purge exhaust and supply valves by cycling the
respective K615 relay. This will not provide assurance that the
valve will perform its safety function since the valve is sealed
closed. The proposed change will exclude the requirement to perform
the K615 relay test (auto actuation logic and actuation relays -
slave relay test) on a quarterly basis while the plant is in Modes
1, 2, 3,or 4.
TS 3.6.1.7. requires that these valves be sealed shut during
Modes 1, 2, 3, and 4. When sealed shut, these valves will not open
via any signal. Since this relay would not be needed to supply a
signal to place these valves in the closed position, the margin of
safety is not affected.
Based on the preceding analysis, SCE&G has determined that this
change does no involve a significant hazards consideration.
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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama

Date of amendment request: June 2, 1995 (TS 353)
Description of amendment request: The proposed amendment supports
replacement of the existing power range neutron monitoring equipment
and implements ARTS/MELLL [average power range monitor and rod block
monitor technical specifications/maximum extended load line limit]
analysis improvements.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Group A Changes: This proposed TS change is associated with the
NUMAC PRNM [nuclear measurement analysis and control power range
neutron monitor] retrofit design. The proposed TS change involves
modification of the LCOs [limiting condition for operations] and SRs
[surveillance requirements] for equipment designed to mitigate
events which result in power increase transients. For the APRM
[average power range monitor] system mitigative action is to block
control rod withdrawal or initiate a reactor scram which terminates
the power increase when setpoints are exceeded. For the RBM [rod-
block monitor] system mitigative action is to block continuous
control rod withdrawal prior to exceeding the MCPR [minimum critical
power ratio] safety limit during a postulated Rod Withdrawal Error
[RWE]. The worst case failure of either the APRM or the RBM systems
is failure to initiate mitigative action (failure to scram or block
rod withdrawal). Failure to initiate mitigative action will not
increase the probability of an accident. Thus, the proposed change
does not increase the probability of an accident previously
evaluated.
For the APRM and the RBM systems, the NUMAC PRNM design,
together with revised operability requirements (LCOs) and revised
testing requirements (SRs), results in equipment which continues to
perform the same mitigation functions under identical conditions
with reliability equal to or greater than the equipment which it
replaces. Because there is no change in mitigation functions and
because reliability of the functions is maintained, the proposed
change does not involve an increase in the consequences of an
accident previously evaluated.
Group B Changes: This proposed change is associated with
implementation of the ARTS/MELLL analysis. The proposed change will
permit expansion of the current allowable power/flow operating
region and will apply a new methodology for assuring that fuel
thermal and mechanical design limits are satisfied. Reference 3
evaluates operation in the MELLL region with assumed implementation
of the ARTS changes. The conclusion of reference 3 is that for all
events and parameters considered there is adequate design margin for
operation in the MELLL region. Because operation in the MELLL region
maintains adequate design margin, the proposed change does not
significantly increase the probability of an accident previously
evaluated.
In support of operation in the MELLL region, the proposed change
modifies flow-biased APRM scram and rod block setpoints and
implements new RBM power-biased setpoints. This potentially changes
the way in which the APRM and RBM systems perform their mitigation
functions. However, no credit for the flow-biased APRM scram or rod
block is taken in mitigation of any design basis event; thus,
changing the APRM setpoints does not impact the consequences of any
accident previously evaluated. The proposed changes to the RBM
system potentially impact mitigation of the RWE. However, per
discussion in reference 3, the proposed RBM changes will assure that
the RWE is not a limiting event; thus, the consequences of the RWE
are not increased. The proposed change does not increase the
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 changes (Group A and Group B) involve modification
and replacement of the existing power range neutron monitoring
equipment, modification of the setpoints and operational
requirements for the APRM and RBM systems, implementation of a new
methodology for administering compliance with fuel thermal limits,
and operation in an extended power/flow domain. These proposed
changes do not modify the basic functional requirements of the
affected equipment, create any new system interfaces or
interactions, nor create any new system failure modes or sequence of
events that could lead to an accident. The worst case failure of the
affected equipment is failure to perform a mitigation action, and
failure of this mitigative equipment does not create the possibility
of a new or different kind of accident. The proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
Group A Changes: This proposed TS change is associated with the
NUMAC PRNM retrofit design. The NUMAC PRNM change does not impact
reactor operating parameters nor the functional requirements of the
power

[[Page 42610]]
range neutron monitoring system. The replacement equipment continues to
provide information, enforce control rod blocks and initiate reactor
scrams under appropriate specified conditions. The proposed change
does not revise any safety margin requirements. The replacement
APRM/RBM equipment has improved channel trip accuracy compared to
the current system and meets or exceeds system requirements
previously assumed in setpoint analysis. Thus, the ability of the
new equipment to enforce compliance with margins of safety equals or
exceeds the ability of the equipment which it replaces. The proposed
change does not involve a reduction in a margin of safety.
Group B Changes: This proposed change is associated with
implementation of recommendations presented in the ARTS/MELLL
analysis. Operation in the MELLL region does not affect the ability
of the plant safety-related trips or equipment to perform their
functions, nor does it cause any significant increase in offsite
radiation doses resulting from any analyzed event. Analyses
documented in reference 3 demonstrate that for operation in the
MELLL region adequate margin to design limits is maintained.
Implementation of the ARTS improvements provides flow- and power-
dependent thermal limits which maintain existing margins of safety
in normal operation, anticipated operational occurrences and
accident events. Implementation of power-biased RBM setpoints
improves the margin of safety in a postulated RWE by assuring that
the RWE is not a limiting event. The proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama

Date of amendment request: June 8, 1995 (TS 361)
Description of amendment request: The proposed amendment clarifies
the definition of operability for the RHRSW system standby coolant
supply capability and revises the instrument numbers for several
instruments that have been upgraded.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change to TS 3.5.C.3 clarifies the operability
requirements of the standby coolant supply capability. It does not
change or degrade the nuclear safety characteristics of the RHRSW
and RHR systems and will not affect the intent of the TS. The
operation of the standby coolant supply capability is not a
precursor to any design basis accident or transient analyzed in the
BFN FSAR. The proposed changes to instrument numbers are
administrative changes for the upgraded drywell temperature and
pressure instrumentation. The proposed changes do not affect the
design basis or the safety functions of the Primary Containment
system, since the function and instrumentation range is not changed.
Therefore, the probability of occurrence or the consequences of an
accident or malfunction of equipment important to safety previously
evaluated in the safety analysis report has not been increased.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The possibility for an accident or malfunction of a different
type than any evaluated previously in the safety analysis report is
not created by this change. The change to TS 3.5.C.3 adds the
indication of associated valves of the function involved and a
clarification of operability for the standby coolant supply
connection to be commensurate with the RHR cross-connect capability.
The proposed changes to instrument numbers are administrative
changes effected by the upgrade of instrumentation. There are no
automatic actions affected or compromised by these changes.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change to TS 3.5.C.3 does not affect any acceptable
limit of operation or analysis assumption in the TS or Bases. The
changes affect neither setpoints, calibration intervals, nor
functional test intervals. The change does not affect any acceptable
limit of operation or analysis assumption found in the TS or their
bases. The proposed administrative changes to the instrument numbers
do not affect the setpoint, calibration interval or function of the
instrumentation. These changes do not affect any limiting conditions
of operation or analysis assumption in the TSs or their bases.
Therefore, the change does not reduce the margin of safety as
defined in the basis for any TS.
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 llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama

Date of amendment request: June 16, 1995 (TS 360)
Description of amendment request: The proposed change will revise
the BFN Units 1, 2, and 3 Technical Specifications (TS) to permit the
Traversing In-Core Probe (TIP) system to be considered operable with
less than five TIP machines operable. The proposed amendment will allow
the utilization of substitute data in lieu of data from inaccessible
TIP measurement locations. The substitute data will be derived from
either symmetric TIP measurement locations (under certain core
conditions) or from normalized TIP data as calculated by the on-line
core monitoring system.
Basis for proposed no significant hazards
considerationdetermination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The TIP system is not used to prevent, or mitigate the
consequences of any previously analyzed accident or transient; nor
are any assumptions made in any accident analysis relative to the
operation of the TIP system. The primary containment isolation
function (TIP withdrawal) is not affected. The
proposed TS change does not alter the fundamental process
involved in calibrating neutron instrumentation (LPRMs) [local power
range monitors], but requires that only the equipment associated
with the TIP channels necessary for recalibrating LPRMs and for core
monitoring functions be operable. Collection and storage of TIP data
without using all TIP channels is acceptable because TIP machine
normalization factors are ultimately derived from the most recent
full core TIP set, which intercalibrates the TIP machines in a
common core location.
Additionally, the use of symmetric detectors and analytical
values as substitute data for inaccessible TIP channels does not
compromise the ability of the process computer to accurately
represent the spatial neutron flux distribution of the reactor core.

[[Page 42611]]
The core monitoring methodology is presently based on symmetry of rod
patterns and fuel loading. This is not changed but extended to use a
higher order of symmetry (octant symmetry) which exists with ``type
A'' sequence rod patterns. Therefore, this change does not increase
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 involve the installation of any new
equipment, or the modification of any equipment designed to prevent
or mitigate the consequences of accidents or transients. Therefore,
the proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The total core TIP reading uncertainties will remain within the
assumptions of the licensing basis. Therefore, the margin of safety
to the MCPR [minimum critical power ratio] safety limits is not
reduced. The ability of the process computer to accurately represent
the spatial neutron flux distribution for the reactor core is not
compromised. Additionally, the computer's ability to accurately
predict the LHGR [linear heat generation rate], APLHGR [average
planar linear heat generation rate], MCPR and its ability to provide
for LPRM calibration is not compromised. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: October 21, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.6.1.2, ``Primary Containment
Leakage.'' The changes would clarify that the main steam line isolation
valves leakage is accounted for separately from the integrated primary
containment leak rate or combined local leak rate results. Also, two
references would be deleted, the test duration for use of Bechtel
Corporation Topical Report BN-TOP-1 would be clarified, and the
requirement to perform the third integrated leak rate in each 10-year
service period in conjunction with the 10-year plant inservice
inspection would be deleted. Exemptions to 10 CFR Part 50 Appendix J,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' are also being requested in conjunction with the proposed
TS changes.
Basis for proposed no significant hazards
considerationdetermination: 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:
Part A - Formalize the Approval for Excluding the Main Steam
Line Isolation Valve Leakages from Inclusion in i) the Overall
Integrated Primary Containment Leak Rate and ii) the Combined Local
Leak Rate, and Clarify that the Main Steam Lines are Not Required to
be Vented and Drained for Type A Testing
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Since Appendix J was originally envisioned, alternative means of
meeting the intent of these requirements have been developed which
provide an equivalent level of protection of the public health and
safety. However, since some of these alternatives deviate from the
specific wording of Appendix J, exemptions are appropriate for these
alternatives. Implicit in the FSAR treatment of the main steam line
leakage, as well as the TS requirements for main steam line leakage,
are several deviations from the specific requirements of Appendix J.
Although PNPP's methods and practices for Appendix J testing have
been previously described in correspondence to the NRC, a formal
exemption was not recognized to be needed at that time in that the
NRC's approval was perceived to be received by the issuance of the
PNPP TS. Exemption to four separate paragraphs of 10 CFR 50 Appendix
J will document the approvals previously received and incorporated
into the TS for main steam line isolation valve testing during the
initial licensing of the PNPP. This TS change adds references to
footnotes within the TS LCO 3.6.3.1 to clarify which conditions
represent exemptions to Appendix J. These exemptions are described
in the Bases.
PNPP utilized the criteria described in the Standard Review Plan
(SRP), Section 15.6.5, Appendix D, ``Radiological Consequences of a
Design Basis Loss-of-Coolant Accident: Leakage from Main Steam
Isolation Valve Leakage Control System (Rev. 1 - July 1981).'' This
is an alternative, NRC approved method for assessing the MSIV
leakage contribution and determining the radiological consequences.
In accordance with the SRP, the safety analysis for a design
basis LOCA includes the maximum main steam line leak rate separately
from the maximum containment leak rate. Within Appendix J it is
implied that Type A tests are intended to measure the primary
containment overall integrated leak rate, but this vas before the
SRP Section was developed which allows the MSIV contribution to be
accounted for separately in the safety analysis. Therefore, the MSIV
leak rate should not be included in the measurement of the ILRT.
Including the MSIV leakage in the combined local leak rate limit is
also not necessary since a specific Type C MSIV leak rate has been
specified in TS 3.6.1.2.
In summary, there is no change in the probability or
consequences of any accident since the addition of the references
and footnotes to clarify the TS LCO and Actions do not change the
design of the plant, nor the operational characteristics of any
plant system, nor the procedures by which the Operators run the
plant. These changes only cite formal Appendix J exemptions which
are requested to document the approval previously received. A formal
request for exemption to the applicable paragraphs of 10 CFR 50
Appendix J is also being submitted in a separate letter in
conjunction with this proposed TS change.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
There are no design changes being made that would create a new type
of accident or malfunction, and the method and manner of plant
operation remains unchanged. The only change being made is an
exemption to 10 CFR 50 Appendix J which will be cited in the TS to
document the implicit and explicit approvals of the PNPP design and
testing methods for main steam line isolation valves. The
requirements and bases for which the formal exemption is sought are
currently presented and implemented in the licensing basis and the
TS for PNPP. The objective of the regulation is being met and will
continue to be met. The exemption to 10 CFR 50 Appendix J is being
submitted in a separate letter in conjunction with this proposed TS
change.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
These changes do not involve a significant reduction in the
margin of safety because they are administrative in nature. The
proposed change will only cite the NRC exemption that grants the
deviation from Appendix J. The proposed changes do not affect any
USAR design bases or accident assumptions. Therefore, the proposed
changes do not reduce the margin of safety as defined in the bases
for any Technical Specification.
Part B - Revise Surveillance Requirement 4.6.1.2 to Eliminate
Unnecessary References and ClarifY the Use of BN-TOP-1

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3AX95-20816. Public record. Not legal advice.