# Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving no Significant Hazards Considerations

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URL: https://www.frixlaw.com/law-library/documents/fr%3A96-11295

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** May 8, 1996
- **Citation:** 61 FR 20842

## Text

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving no Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 13, 1996, through April 26, 1996. The
last biweekly notice was published on April 24, 1996 (61 FR 18162).

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 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By June 7, 1996, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714

[[Page 20843]]

which is available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: 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 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

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 28, 1996.
Description of amendments request: Pursuant to 10 CFR 50.90, the
Baltimore Gas and Electric Company (BGE) hereby requests an amendment
to Operating License Nos. DPR-53 and DPR-69 to reduce the moderator
temperature coefficient (MTC) limit shown on Technical Specification
Figure 3.1.1-1. This proposed change is necessary to support changes in
the safety analyses made to accommodate a larger number of plugged
steam generator (SG) tubes for future operating cycles. The proposed
limit will be more restrictive than the existing limit to match the
analytical assumptions. In addition, the licensee provided information
to clarify the relationship of the MTC to an Anticipated Transient
Without Scram event in its licensing basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The safety analyses for the current fuel cycles assume 500 tubes
per steam generator (SG) are plugged and the maximum beginning-of-
cycle moderator temperature coefficient (MTC) is assumed to follow
the curve in Technical Specification Figure 3.1.1.-1. For the fuel
cycle to be installed in Unit 1 in spring 1996, Baltimore Gas and
Electric Company (BGE) assumes in the analyses that more SG tubes
are plugged than the current limit, and it is necessary to credit a
more restrictive (less positive) limit on the maximum positive MTC
to mitigate the

[[Page 20844]]

Reactor Coolant System pressure and temperature increase analyzed
for these events. Therefore, we are proposing a change to the
allowable positive MTC limits shown on Technical Specification
Figure 3.1.1-1. The proposed limit will be more restrictive than the
existing limit to match the analytical assumptions. Since the safety
analyses supporting an increase in the number of plugged SG tubes
are applicable to both Units 1 and 2, BGE is requesting this change
for both Units.
The proposed change makes the limit on the maximum positive MTC
more restrictive. From an operational standpoint, a more restrictive
limit on MTC will help mitigate the effect of plant transients on
control of plant parameters (e.g., reactor power, pressurizer
pressure, pressurizer level, etc.) Therefore, the probability of a
previously analyzed accident will not be significantly increased.
The reason for the proposed change is to mitigate the effect
(increased reactor coolant temperatures) of increased SG U-tube
plugging on the results of the affected safety analyses. Using the
more restrictive limit on the maximum positive MTC, the Loss of
Load, Loss of Feedwater Flow, Feed Line Break, and Control Element
Assembly Withdrawal events were reanalyzed using previously accepted
methodologies. The results of these analyses are within the
acceptance limits for these events. Therefore, the consequences of a
previously analyzed accident will not be significantly increased.
The proposed change is similar to the examples of amendments
that are considered not likely to involve significant hazards
considerations given in the Statements of Consideration for 10 CFR
50.92 (51 FR 7744). The example of interest is, ``A change that
constitutes an additional limitation, restriction, or control not
presently included in the technical specifications, e.g., a more
stringent surveillance requirement.'' The proposed change provides a
more restrictive limit on the positive MTC given in Technical
Specification Figure 3.1.1-1. Based on the above arguments and the
similarity to an example in the Federal Register, BGE has determined
that the proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed change makes the limit on the maximum positive MTC
more restrictive. The proposed change does not involve installation
of new or different equipment, modify the interfaces with existing
equipment, change the equipment's function, or change the method of
operating the equipment. The proposed change does not affect normal
plant operations or configurations. The more restrictive MTC limit
will help mitigate the effect of plant transients on control of
plant parameters.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The proposed change provides for a more restrictive limit for
the allowable positive MTC. The more restrictive limit on the
maximum positive MTC was evaluated using previously approved
methodologies and compared to the existing acceptance criteria. The
analyses show that the proposed change preserves the margin of
safety by ensuring that the results of the safety analyses for the
Loss of Load, Loss of Feedwater Flow, Feed Line Break, and Control
Element Assembly Withdrawal events meet established NRC acceptance
limits for these events.
In addition, this proposed change is similar to the example of
amendments that are considered not likely to involve significant
hazards considerations given in the Statements of Consideration for
10 CFR 50.92 (51 FR 7744). The example of interest is, ``A change
that constitutes an additional limitation, restriction, or control
not presently included in the technical specifications, e.g., a more
stringent surveillance requirement.'' The proposed change provides a
more restrictive limit on the positive MTC given in Technical
Specification Figure 3.1.1-1. Based on the above arguments and the
similarity to an example in the Federal Register, BGE has determined
that the proposed change does not involve a significant reduction in
the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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: Susan F. Shankman, Acting.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina

Date of amendment request: February 8, 1996.
Description of amendment request: The proposed amendment would
remove Technical Specifications (TS) 3.3.4, Turbine Overspeed
Protection; TS 3.7.12, Area Temperature Monitoring; and TS 3.11.2.6,
Gas Storage Tanks; and their associated bases; and relocate them to
licensee-controlled documents, such as the Final Safety Analysis
Report. The licensee revised the original amendment request dated
October 24, 1994, to provide supplemental information to TS 6.8.4 for
administrative control program related to TS 3.11.2.6, by letters dated
August 31, 1995 and February 8, 1996.
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 was previously presented in the Federal Register
(59 FR 60397). The staff reviewed and determined that the proposed
license amendment's revisions do not alter the original conclusion that
no significant hazards considerations exist pursuant to 10 CFR 50.92.
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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Eugene V. Imbro.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: December 21, 1995.
Description of amendment request: The proposed amendments would
delete the requirement to place the reactor mode switch in the Shutdown
position if a stuck open safety/relief valve cannot be closed within
two minutes. The operator would still be required to scram the reactor
if suppression pool average water temperature reaches 110 degrees
Fahrenheit or greater. The licensee also proposed changes to the TS
index pages to reflect Bases page changes that were accepted by the NRC
staff in a letter dated May 23, 1995. Because the changes to the index
pages require a license amendment, they have been included as part of
this submittal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The proposed change does not involve a significant increase in
the probability or

[[Page 20845]]

consequences of an accident previously evaluated in the UFSAR. A
stuck open SRV event is a mild transient which neither affects fuel
limits nor radiological consequences. The two minute requirement to
manually scram after a SRV becomes stuck open is not assumed or used
in any transient or accident analysis in the FSAR. Removing the two
minute requirement to manually scram after a SRV becomes stuck open
does not change the probability of any accident evaluated in the
FSAR. Removing the two minute requirement to manually scram after a
SRV becomes stuck open also does not change the capability of the
suppression pool during this event in case of any accident involving
reactor blowdown, because the suppression pool average water
temperature limit in Technical Specification 3.6.2.1 is still valid
and enforced. The suppression pool average water temperature limit
is the only requirement during operational conditions 1 and 2 that
assures sufficient heat sink capacity in case of a LOCA in the
containment. Therefore, removing the two minute requirement to
manually scram after a SRV becomes stuck open would not increase the
probability or consequences of any postulated accident analyzed in
the FSAR.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated
in the UFSAR. This change does not effect any hardware. This is a
procedural change to assure that the reactor will not be
unnecessarily scrammed by the operator after a SRV is stuck open for
two minutes. The reactor will still be scrammed if suppression pool
average water temperature increases above 110 degrees F. Since the
design basis of the suppression pool is protected by this average
water temperature limit, this procedural change of removing the two
minute requirement to manually scram after a SRV becomes stuck open
introduces no new accident or malfunction.
(3) Involve a significant reduction in the margin of safety
because:
The proposed change does not reduce the margin as defined in the
bases for any Technical Specification. On the contrary, if the two
minute requirement to manually scram after a SRV becomes stuck open
is not removed, the operator has to scram the reactor thus
challenging the RPS, the rector vessel, and other associated
components, and reducing the related margin to safety. This scram
would be unnecessary if the suppression pool average water
temperature is below the 110 degree F limit allowed by the design
basis of the suppression pool. Reactor safety or suppression pool
design basis is not compromised because the suppression pool average
water temperature limit alone guarantees that there would not be any
reduction in margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: March 4, 1996.
Description of amendment request: The proposed amendments would
change the McGuire Units 1 and 2 Updated Final Safety Analysis Report
to delete the seismic qualification requirement for the Containment
Atmosphere Particulate Radiation Monitors.
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:

This proposed change has been evaluated against the standards in
10 CFR 50.92 and has been determined to involve no significant
hazards considerations, in that operation of the facility in
accordance with the proposed amendment would not:
1. [I]nvolve a significant increase in the probability or
consequences of an accident previously evaluated; or
EMF38(L) is not used directly for any phase of power generation
or conversion or transmission, normal decay heat removal, fuel
handling, or the processing of radioactive fluids. As such, it is
not an ``accident initiator''. No ``accident initiator'' is affected
by the change. Thus, the probability of accidents evaluated in the
FSAR is not affected by the change. It is determined that sufficient
ability to determine conditions inside containment remain available
for any earthquake up to and including the SSE [safe-shutdown
earthquake]. Furthermore, should either EMF38(L) or EMF39(L) be
found to not be functional following any earthquake, including those
smaller than the OBE [Operating Basis Earthquake], the appropriate
steps will be taken; i.e., declare the monitor(s) inoperable and
apply the action statement for TS [technical specification] 3.4.6.1
which may require that the associated unit(s) be taken to Cold
Shutdown (Mode 5) if the minimum required Reactor Coolant Leakage
Detection Systems are not operable. Cold Shutdown is a mode for
which neither the Emergency Core Cooling System nor the containment
safeguards are required. Finally, no equipment provided to mitigate
any accident is adversely affected by the change. For these reasons,
the proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated in
the SAR [safety analysis report].
2. [C]reate the possibility of a new or different type of
accident from any accident previously evaluated; or
As stated above, no equipment used in direct support of power
generation or conversion or transmission, normal decay heat removal,
fuel handling, or processing of radioactive fluids is affected with
the update. No new failure modes are identified with the change. The
upper bound to an undetected leak in the Reactor Coolant System is a
Loss of Coolant Accident [LOCA]. As noted above, no equipment
provided to mitigate a LOCA is affected by the change. For these
reasons, the change will not create a new or different type of
accident from any accident previously evaluated.
3. [I]nvolve a significant reduction in a margin of safety.
It has been determined that sufficient means remain at the
disposal to the operators to assess conditions within the
containment following any earthquake up to and including the SSE. In
particular, the ability to determine leakage with the sensitivity
comparable to that of EMF38(L) can be established. This meets the
intent of the Regulatory Position of RG [Regulatory Guide] 1.45. In
addition, should it be determined that either EMF38(L) or EMF39(L)
is not functional following any earthquake, the appropriate steps
will be taken; i.e, declare the monitor(s) inoperable and apply the
action statement for TS 3.4.6.1 which may require that the
associated unit(s) be taken to Cold Shutdown (Mode 5) if the minimum
required Reactor Coolant Leakage Detection Systems are not operable.
This brings the unit(s) to a mode in which TS 3.4.6.1 does not
apply. It ensures that at least the minimum required Reactor Coolant
System leakage detection systems will be functional before power
operations are continued following a postulated earthquake smaller
than the OBE. It ensures protection of the reactor coolant pressure
boundary, one of the fission product barriers. No other fission
product barrier is affected by the change. Therefore, the margin of
safety is not reduced.
Therefore, based on the information contained in this submittal,
it is determined that no significant hazard is associated with the
proposed change.

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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South

[[Page 20846]]

Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas

Date of amendment request: April 11, 1996.
Description of amendment request: The proposed technical
specification amendment modifies the reactor building leak testing
requirements per Option B to 10 CFR 50, Appendix J. Option B permits
performance based determination of the reactor building leak testing
frequency.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does Not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
The proposed changes to the Technical Specifications implement
Option B of 10 CFR 50 Appendix J at ANO. The proposed changes will
result in increased intervals between containment leakage tests
determined through a performance based approach. The intervals
between such tests are not related to conditions which cause
accidents. The proposed changes do not involve a change to the plant
design or operation. Therefore, this change does not involve a
significant increase in the probability of any accident previously
evaluated.
NUREG-1493, ``Performance-Based Containment Leak-Test Program,''
contributed to the technical bases for Option B of 10 CFR 50
Appendix J. NUREG-1493 contains a detailed evaluation of the
expected leakage from containment and the associated consequences.
The increased risk due to lengthening of the intervals between
containment leakage tests was also evaluated and found acceptable.
Using a statistical approach, NUREG-1493 determined the increase in
the expected dose to the public from extending the testing frequency
is extremely small. It also concluded that a small increase is
justifiable due to the benefits which accrue from the interval
extension. The primary benefit is in the reduction in occupational
exposure. The reduction in the occupational exposure is a real
reduction, while the small increase to the public is statistically
derived using conservative assumptions. Therefore, this change does
not involve a significant increase in the consequences of any
accident previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Does Not Create the Possibility of a New or Different Kind of
Accident from any Previously Evaluated.
The proposed change to the Technical Specifications incorporates
the performance based approach authorized by Option B of 10 CFR 50
Appendix J. The interval extensions allowed by this change do not
involve a change to the plant design or operation. No safety related
equipment or safety functions are altered as a result of this
change. The reduced testing frequency does not affect the testing
methodology. As a result, the proposed change does not affect any of
the parameters or conditions that could contribute to initiation of
any accidents. No new accident modes are created by extending the
test intervals. Therefore, this change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does Not Involve a Significant Reduction in the Margin of
Safety.
The proposed change does not change the performance methodology
of the containment leakage rate testing program. However, the
proposed change does affect the frequency of containment leakage
rate testing. With an increased frequency between tests, the
proposed change does increase the probability that a increase in
leakage could go undetected for a longer period of time. Operational
experience has demonstrated the leak tightness of the containment
buildings has been significantly below the allowable leakage limit.
The margin to safety that has the potential of being impacted by
the proposed change involves the offsite dose consequences of
postulated accidents which are directly related to containment
leakage rates. The limitation on containment leakage rate is
designed to ensure the BWN total leakage volume will not exceed the
value assumed in our accident analysis. The margin to [sic] safety
for the offsite dose consequences of postulated accidents directly
related to containment leakage is maintained by meeting the 1.0 L.
acceptance criteria. The proposed change maintains the 1.0 L.
acceptance criteria.
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, DC 20005-3502.
NRC Project Director: William D. Beckner.

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

Date of amendment request: April 11, 1996.
Description of amendment request: The proposed technical
specification (TS) amendment adds low-temperature overpressure
protection (LTOP) requirements to the TSs to resolve Generic Issue 94
in accordance with Generic Letter 90-06.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does Not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
This proposed change provides additional controls in the ANO-2
Technical Specification [(TS)] for ensuring that LTOP [low-
temperature overpressure protection] protection is available when
required. The limiting condition involving the simultaneous
injection of two HPSI [high pressure safety injection] and three
charging pumps to an RCS [reactor coolant system] water solid
condition, was used in the calculation of the ANO-2 proposed LTOP
setpoints. The methodology utilized in the LTOP setpoint analysis is
based on ASME [American Society of Mechanical Engineers] Code Case
N-514. The code case establishes a factor of 110 percent of the
operating pressure temperature curves instead of 100 percent. The
safety factor utilized by the code case provides a more reasonable
vessel overpressure allowance for conditions expected under pressure
loading from low temperature transients. The SITs [safety injection
tanks] are required to be isolated, if not depressurized, prior to
entering the LTOP enable temperature and are periodically verified
to be isolated when LTOP conditions exist. The LTOP setpoint of the
relief valves proposed by this technical specification [TS] change
is not considered to be an initiator of any transients, but is used
to mitigate an overpressure condition if such a transient were to
occur.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does Not Create the Possibility of a New or Different Kind of
Accident from any Previously Evaluated.
The design basis event for establishing LTOP limits is the
simultaneous injection of two HPSI and three charging pumps to an
RCS water solid condition. The LTOP vent size of 6.38 square inches
and the valve pressure setpoint of less than or equal to 430 psig
are currently used for mitigation of low temperature overpressure
conditions. The change in the enable setpoint was analyzed by the
application of Code Case N-514 and determined to adequately ensure
that this temperature [sic] setpoint will mitigate a

[[Page 20847]]

LTOP transient. The operator action to enable the LTOP relief valves
at 220 degrees ensures that the RCS including the reactor vessel
will not undergo system pressures at low temperature conditions
beyond their design limits. Therefore, there will not be any impact
to systems, structures or components beyond their design
requirements.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Does Not Involve a Significant Reduction in the Margin of
Safety.
The addition of a new specification to the ANO-2 Technical
Specification [TS] will not significantly reduce the margin of
safety. The LTOP safety factors are based on reanalyzed conditions
for 21 effective full power years (EFPY) of operation utilizing
methodology contained in ASME Code Case N-514. The LTOP evaluation
under Code Case N-514 for low temperature transients is considered
more appropriate than the ASME Section XI. The code case establishes
a factor of 110 percent of the operating pressure temperature curves
instead of 100 percent. The safety factor utilized by the code case
provides a more reasonable vessel overpressure allowance for
conditions expected under pressure loading from low temperature
transients. Although the proposed setpoint may involve a slight
reduction in a margin of safety, the enable temperature setpoint
will provide an equivalent level of safety to the reactor vessel
during LTOP transients and will satisfy the purpose of 10 CFR 50.60
for fracture toughness. Therefore, based on the refined methodology
used to calculate ANO-2 LTOP setpoints for 21 EFPY the margin of
safety will not be significantly reduced.

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, DC 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: April 18, 1996.
Description of amendment request: The licensee has proposed to
delete a restriction on the 24-hour emergency diesel generator
operation test in Surveillance Requirement 3.8.1.14 (Page 3.8-12) of
the Technical Specifications (TSs) for the Grand Gulf Nuclear Station,
Unit 1. The deletion would allow the test to also be conducted during
power operation (i.e., during Modes 1 and 2), instead of the current
requirement to only conduct the test when the plant is shut down.
The frequency of conducting this test, the conditions of the test,
and the criteria to pass the test are not being changed by this
amendment request.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration for the amendment request, which is presented below:

Entergy Operations, Inc. [(EOI)] propose[d] to change the
current Grand Gulf Nuclear Station [GGNS] Technical Specifications
[(TSs)]. The specific change is to modify note 2 to Surveillance
3.8.1.14. Presently, this note prohibits the performance of the 24
hour diesel maintenance run while the unit is in either Mode 1 or 2.
The proposed change would remove this restriction thus allowing the
24 hour run to be performed during any mode of operation (i.e.,
modes 1, 2, 3, 4 or 5).
The Commission has provided standards for determining whether a
no significant hazards considerations exists as stated in 10 CFR
50.92 (c). A proposed amendment to an operating license involves no
significant hazards consideration if operation of the facility in
accordance with the proposed amendment would not: (1) involve a
significant increase in the probability or consequences of an
accident previously evaluated; (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.
Entergy Operations, Inc. [EOI] has evaluated the no significant
hazards consideration in its request for this license amendment and
determined that no significant hazards considerations results from
this change. In accordance with 10 CFR 50.91(a), Entergy Operations,
Inc. [EOI] is providing the analysis of the proposed amendment
against the three standards in 10 CFR 50.92(c). A description of the
no significant hazards consideration determination follows:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The GGNS UFSAR [Updated Final Safety Analysis Report] assumes
that the AC electrical power sources are designed to provide
sufficient capacity, capability, redundancy and reliability to
ensure that the fuel, reactor coolant system and containment design
limits are not exceeded during an assumed design basis event.
Specifically, the UFSAR assumes that the onsite EDG's [emergency
diesel generator's] provide emergency power in the event offsite
power is lost to either one or all three ESF [engineered safety
feature] buses. In the event of a loss of preferred power, the ESF
electrical loads are automatically connected to the EDG's in
sufficient time to provide for safe reactor shutdown and to mitigate
the consequences of a design basis accident such as a LOCA.
The proposed change to permit the 24 hour testing of the EDG's
during power operation does not increase the chances or consequences
of any previously evaluated accident. The capability of the EDG's to
supply power in a timely manner will not be compromised by
permitting performance of EDG testing during periods of power
operation. Design features of the EDG's and electrical systems
ensures that if a LOCA [loss of coolant accident] or LOP [loss of
offsite power] signal, either individually or concurrently, should
occur during testing that the EDG would be returned to its ready-to-
load operation (i.e., EDG running at rated speed and voltage
separated from the offsite sources) or separately connected to the
ESF bus providing ESF loads. As such, an EDG being tested is
considered to be Operable and fully capable of meeting its intended
design function. Additionally, the testing of an EDG is not a
precursor to any previously evaluated accidents.
Therefore, the proposed change allowing testing of EDG's during
power operation will not significantly increase the probability or
consequences of an accident previously evaluated.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
As previously discussed [above], the proposed change to permit
the performance of EDG testing during power operation will not
affect the operation of any system or alter any system's response to
previously evaluated design basis events. The EDG's will
automatically transfer from the test configuration to the ready-to-
load configuration following receipt of a valid signal (i.e., LOCA
or LOP). In the ready-to-load configuration, the EDG will be running
at rated speed and voltage separated from the offsite source capable
of automatically supplying power to the ESF buses in the event that
preferred power is actually loss.
Surveillance Requirement 3.8.1.17 demonstrates that the EDG will
automatically override the test mode following generation of a LOCA
signal. In addition the ability of the EDG's to survive a full load
reject is verified by the performance of surveillance requirement
3.8.1.10. These existing surveillance requirements along with system
design features ensures that the performance of EDG testing during
power operation will not create the possibility of a new or
different kind of accident from any previously evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The AC electrical power sources are designed to provide
sufficient capacity, capability, redundancy, and reliability to
ensure the availability of necessary power to ESF systems so that
the fuel, reactor coolant system and containment design limits are
not exceeded. Specifically, the EDG's must be capable of
automatically providing power to ESF loads in sufficient time to
provide for safe reactor shutdown and to mitigate the consequences
of a design basis accident in the event of a loss of preferred
power.

[[Page 20848]]

Testing of EDG's during power operation will not affect the
availability or operation of any offsite source of power. In
addition, the EDG being tested remains capable of meeting its
intended design functions. Therefore the proposed change to the
Technical Specification Surveillance Requirement 3.8.1.14 will not
result in a reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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, MS 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.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: April 15, 1996 (TSCR No. 244).
Description of amendment request: The proposed amendment would
revise Specification 5.3.1.B of the Oyster Creek Technical
Specifications. The current specification prohibits handling a load
greater in weight than one fuel assembly over irradiated fuel in the
spent fuel storage facility. The proposed change will facilitate the
off load of spent fuel to the Oyster Creek Independent Spent Fuel
Storage Installation (ISFSI). Specifically, the shield plug for the dry
shield canister (DSC) and the associated lifting hardware will be moved
over irradiated fuel which is contained in the DSC within the transfer
cask located in the Cask Drop Protection System (CDPS).
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. State the basis for the determination that the proposed
activity will or will not increase the probability of occurrence or
consequences of an accident.
The design features and capacity of the reactor building crane
provide a significant safety factor. In addition, personnel training
and other administrative controls further reduce risk. Thus, the
dropping of the DSC shield plug onto a loaded DSC and causing damage
to the spent fuel assemblies is not a credible event. Therefore, it
does not increase the probability of or consequences of an accident.
2. State the basis for the determination that the activity does
or does not create the possibility of an accident or malfunction of
a different type than any previously identified in the SAR [safety
analysis report].
This activity will not create the possibility of a new or
different type of accident than previously evaluated in the SAR
because the proposed heavy load handling exception does not create a
new credible accident scenario. Dropping the shield plug on a loaded
DSC and damaging spent fuel assemblies is not considered a credible
event.
3. State the basis for the determination that the margin of
safety is not reduced.
This activity will not involve a significant reduction in the
margin of safety because the proposed heavy load handling evolution
does not create a credible accident scenario.

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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine

Date of amendment request: April 19, 1996.
Description of amendment request: The proposed amendment would
revise Technical Specification 5.14 to add the appropriate references
identifying the detailed methodology and conditions for analyzing the
Small Break Loss-of-Coolant Accident (SBLOCA) to the list of the
approved Core Operating Limits Report methods.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the Proposed Amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
These Proposed Changes are administrative in nature and are
consistent with the guidance set forth in the NRC Generic Letter 88-
16 identifying the requirements for the inclusion of analytical
methodology references in Technical Specifications as used in
determining compliance with the regulatory limits.
The references, as proposed to be included in section 5.14 of
the Technical Specifications, have previously been reviewed and
approved by the NRC for generic applicability to PWRs [Pressurized
Water Reactors]. The reports identified in the Proposed Change have
been accepted by the NRC for referencing in plant licensing
applications.
Since the references listed in the Proposed Change have
previously been found to meet the conditions of 10 CFR 50.46 and 10
CFR Appendix K, and that the plant specific safety analysis
acceptance limits have not changed or been modified, the use of
these references in the analysis of SBLOCA accident for the Maine
Yankee plant is consistent with prior plant specific and industry
requirements and practices.
Therefore, we have concluded that the Proposed Change will not
result in a significant increase in the probability or consequences
of an accident previously evaluated.
2. Does the Proposed Amendment create the possibility for a new
or different kind of accident?
The Proposed Changes introduce no new mode of plant operation;
do not involve the physical modification of any structure, system,
or component; do not affect the function, operation or surveillance
for any equipment necessary for safe operation or shutdown of the
plant; and, do not involve any changes to setpoints or limits or
operating parameters. The Proposed Changes are administrative in
nature only.
Therefore, we have concluded that the Proposed Change cannot
result in the possibility of a new or different kind of accident
from that previously evaluated.
3. Does the Proposed Amendment involve a significant reduction
in a margin of safety?
The Proposed Changes are administrative in nature, consistent
with the guidance of Generic Letter 88-12, and have been reviewed
previously by the NRC and found acceptable with regard to the
requirements of 10 CFR 50.46 and 10 CFR Appendix K. Additionally,
the plant specific safety analysis acceptance criteria has not
changed from that used in the latest core reload analysis.
Therefore, we have concluded that the Proposed Change does not
involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011.

[[Page 20849]]

NRC Deputy Director: John A. Zwolinski.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: February 7, 1996.
Description of amendment request: The amendment would change the
operating license, the Technical Specifications, and associated Bases
to permit the use of 10 CFR Part 50, Appendix J, Option B, Performance-
Based Containment Leakage Rate Testing in accordance with the
implementation guidance in NRC's Regulatory Guide 1.163 dated September
1995. The change to the operating license would delete, in paragraph
2.D.ii, reference to certain exemptions to Appendix J previously
granted by the NRC, which would no longer be applicable once Option B
is implemented.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
NMP2 [Nine Mile Point, Unit 2] is currently implementing Option
A of Appendix J of 10 CFR 50 for Type A, B and C testing. The
proposed change to the Operating License, the Technical
Specifications and the Bases would implement Option B to Appendix J
of 10 CFR 50 at NMP2 for Type A, B and C testing. Option B would
allow increased testing intervals after satisfying certain
performance based criteria. The proposed change also corrects an
inconsistency between the restoration statements and the
applicability requirements of LCO [Limiting Condition of Operation]
3.6.1.2. In addition, the proposed change affects the testing
intervals for the verification of the interlocks on the primary
containment air lock and for the measuring of the Hydrogen
Recombiner System leakage rate.
Appendix J describes the requirements for leakage testing of the
primary containment and its components penetrating the primary
containment. The leakage testing interval of the primary containment
and its components is not a precursor or initiator to an accident.
The primary containment and its penetrations minimizes the leakage
of radioactivity into the environment during an accident which
pressurizes the primary containment.
The testing intervals of the air lock interlocks and of the
Hydrogen Recombiner System leakage rate are also not precursors or
initiators to an accident. The interlocks function to provide
assurance that at least one air lock door will be closed and thereby
perform its accident mitigating function of minimizing the leakage
of radioactivity into the environment during accident conditions.
The Hydrogen Recombiner System is manually initiated following a
loss-of-coolant accident (LOCA) to maintain the hydrogen
concentration within the primary containment below its flammable
limit during post-LOCA conditions.
An inconsistency exists between the applicability statement of
LCO 3.6.1.2 and the requirement of the restoration statements to
restore prior to increasing reactor coolant system temperature over
200 deg.F. Eliminating this inconsistency does not diminish the
requirements contained in the Technical Specifications.
Therefore, the proposed change does not involve a significant
increase in the probability of an accident previously evaluated.
The proposed change to the Operating License, the Technical
Specifications and the Bases would replace the detailed and
prescriptive technical requirements contained in Option A of
Appendix J with performance based requirements and supporting
regulatory/industry documents contained in Option B of Appendix J.
This proposed change includes a description of the 10 CFR 50
Appendix J Testing Program Plan in Section 6.8.4.f of the Technical
Specifications.
This program plan, with one exception, is consistent with RG
[Regulatory Guide] 1.163. This exception to the RG is acceptable as
it is technically equivalent to and replaces an exemption that was
applicable to Option A of Appendix J. Therefore, this program plan
establishes leakage-rate test methods, procedures, acceptance
criteria and analyses which comply with Option B of Appendix J to 10
CFR 50.
The implementation of this program continues to provide adequate
assurance that during a DBA [Design Basis Accident]-LOCA the primary
containment and its components will continue to limit leakage rates
to less than the allowable leakage rates described in the Technical
Specifications and thereby limit leakage consistent with the
assumptions of the accident analyses. Therefore, the increased test
intervals permitted by Option B for the primary containment and its
penetrations will continue to implement the safety objectives
underlying the requirements of Appendix J.
As discussed under the margin of safety, the impact of the
proposed change on the consequences of a release is negligible. The
slight increase in the risk to the population is compensated by the
corresponding risk reduction benefits associated with the reduction
in component cycling, stress, and wear associated with increased
test intervals.
At least one air lock door in each air lock will continue to be
closed during the onset of an accident that would release
radioactivity into primary containment. Therefore, the air lock
interlocks continue to provide assurance that at least one leak
tested barrier will limit leakage during accident conditions.
The Hydrogen Recombiner System will continue to operate to
maintain the hydrogen concentration within the primary containment
below its flammable limit during post-LOCA conditions. This provides
assurance that primary containment integrity will not be challenged
by hydrogen burns.
Eliminating the inconsistency between the restoration statements
and the applicability requirements of LCO 3.6.1.2 does not diminish
the requirements contained in the Technical Specifications. The
Technical Specifications continue to require that the leakage limits
of LCO 3.6.1.2 be met prior to entering OPERATIONAL CONDITIONS 1, 2,
or 3 (i.e., temperature greater than 200 deg.F).
Accordingly, operation with the proposed change to the Operating
License, the Technical Specifications and the Bases will not
significantly increase the consequences of an accident previously
evaluated.
2. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change would implement Option B of Appendix J of 10
CFR 50 for Type A, B and C testing. Option B would allow increased
testing intervals after satisfying certain performance based
criteria. The proposed change also corrects any inconsistency
between the restoration statements and the applicability
requirements of LCO 3.6.1.2. In addition, the proposed change
affects the testing intervals for the interlocks on the primary
containment air lock and for the measuring of the Hydrogen
Recombiner System leakage rate.
No new plant operating modes, system operating configurations
nor failure modes are introduced by the proposed change. The primary
containment and its penetrations will continue to perform their
accident mitigating function. The Hydrogen Recombiner System will
continue to function to prevent hydrogen burns within primary
containment during post-LOCA conditions.
Accordingly, operation with the proposed change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant reduction in
a margin of safety.
A regulatory impact analysis of implementing performance-based
requirements indicates that relaxing the frequency of Type A, B and
C testing leads to an increase in overall reactor risk of
approximately two percent. As indicated in the Staff's Regulatory
Impact Analysis, this increase is considered to be marginal to
safety.
As indicated above, increasing test intervals can slightly
increase the risk to the population associated with the consequences
of a release; however, this is compensated by the corresponding risk
reduction benefits associated with the reduction in component
cycling, stress, and wear associated with increased test intervals.
Therefore, when considering the total integrated risk, the risk
associated with increased test intervals is negligible.

[[Page 20850]]

The proposed change is consistent with current plant safety
analyses. In addition, the proposed change does not require
revisions to the design of NMP2. As such, the proposed individual
changes will maintain the same level of reliability of the equipment
associated with containment integrity, assumed to operate in the
plant safety analysis, or provide continued assurance that specified
parameters affecting leak rate integrity, will remain within their
acceptance limits.
The as-left leakage after performing a required leakage test
continues to be less than 0.60 La for combined Type B and C leakage
and less than or equal to 0.75 La for Type A leakage. These as-left
acceptance criteria and the testing frequency as established by the
10 CFR 50 Appendix J Testing Program Plan provide assurance that the
measured leakage rate will not exceed the maximum allowable leakage
of La during plant operation.
Visual examination of accessible interior and exterior surfaces
of the primary containment continues to be performed prior to
initiating a Type A test. The total number of visual examinations
performed will continue to be three times during a 10-year period.
Therefore, visual examinations of the primary containment will
continue to allow for the timely uncovering of evidence of
structural deterioration and satisfy the requirements of RG 1.163.
The primary containment air lock interlocks will be tested prior
to conducting an air lock seal leakage test. This testing
requirement continues to provide adequate assurance that at least
one leak tested air lock door in each air lock will be closed during
accident conditions.
The measuring of the Hydrogen Recombiner System Leakage rate
will continue to be included as part of the overall integrated
leakage rate test. The test schedule for measuring system leakage
will also continue to coincide with the schedule for performing a
Type A test.
The leakage limits of LCO 3.6.1.2 will continue to be met prior
to entering into OPERATIONAL CONDITIONS 1, 2, or 3 (i.e.,
temperature greater than 200 deg.F). Satisfying these leakage
limits provides assurance that the measured leakage rate will not
exceed the maximum allowable leakage rate of La during plant
operation. Therefore, operation with the proposed change will not
involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Susan Frant Shankman, Acting

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: January 17, 1996.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) including revisions to
Specifications 3/4.3.1, ``Reactor Protection System Instrumentation,''
3/4.3.2, ``Isolation Actuation Instrumentation,'' 3/4.3.3, ``Emergency
Core Cooling System Actuation Instrumentation,'' 3/4.3.4.2, ``End-of-
Cycle Recirculation Pump Trip System Instrumentation,'' and the
associated Bases to relocate response time limit tables from the TSs to
the Updated Safety Analysis Report (USAR). The proposed revisions to
the TSs also include several administrative changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment relocates Tables 3.3.1-2, ``Reactor
Protection System Response Times,'' 3.3.2-3, ``Isolation System
Instrumentation Response Times'' 3.3.3-3, ``Emergency Core Cooling
System Response Times'' and 3.3.4.2-3 ``End-of-Cycle Recirculation
Pump Trip System Response Time'' from the Technical Specifications
to the USAR. The Technical Specification Surveillance Requirements
and associated actions are not affected and remain in the Technical
Specifications. This change to the reactor protection system
instrumentation, isolation actuation instrumentation, and emergency
core cooling system instrumentation is being done in accordance with
the guidance provided in Generic Letter 93-08, ``Relocation of
Technical Specification Tables of Instrument Response Time Limits,''
and the change to the end-of-cycle recirculation pump trip system
instrumentation is consistent with NUREG 1433, ``Standard Technical
Specifications, BWR/4.'' This change allows NMP2 [Nine Mile Point
Unit 2] to administratively control subsequent changes to the
response time limits in accordance with 10CFR50.59. Additionally,
procedures which contain the various response time limits are also
subject to the change control provisions of 10 CFR 50.59. Relocating
this information does not affect the initial conditions of a design
basis accident or transient analysis. The proposed Technical
Specification changes do not affect the capability of the associated
systems to perform their intended functions within their required
response times. Since any subsequent changes to the USAR or
procedures which contain the response time limits are evaluated in
accordance with 10CFR50.59, the proposed amendment does not involve
an increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed change would relocate the response time limit
tables from the Technical Specifications to the USAR. Subsequent
changes to the USAR, or in procedures which contain the various
response time limits, would be evaluated in accordance with the
requirements of 10CFR50.59, which would evaluate the possibility of
the creation of a new or different kind of accident. The proposed
change does not involve any physical alteration of the plant, change
in a Limiting Condition for Operation or change in Surveillance
Requirements. No new failure modes are introduced. Therefore, this
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant reduction in
a margin of safety.
The proposed change would relocate the response time limit
tables from the Technical Specifications to the USAR. Future changes
to the response time limits in the USAR, or in procedures which
contain the various response time limits, would be in accordance
with 10CFR50.59, which would evaluate the proposed change to
determine whether it involved any reduction in the margin of safety.
The response time limits to be transposed from the Technical
Specifications to the USAR are the same as the existing Technical
Specifications. 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Susan Frant Shankman, Acting.

[[Page 20851]]

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: January 25, 1996.
Description of amendment request: The proposed amendment would
change a footnote in Table 3.3.3-1 and the corresponding footnote in
surveillance Table 4.3.3.1-1 (both referenced by Technical
Specification 3/4.3.3 ``Emergency Core Cooling System Actuation
Instrumentation'') to more clearly define when, during cold shutdown
and refueling (i.e., Operational Conditions 4 and 5), the Loss of
Voltage and Degraded Voltage relays associated with the 4.16 kV
Emergency Bus Undervoltage are required to be operable. The footnotes
currently state: ``Required when ESF [Engineered Safety Features]
equipment is required to be OPERABLE.'' The proposed amendment would
change the footnotes to state: ``Required when the associated diesel
generator is required to be OPERABLE.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequence of an accident previously evaluated.
The proposed change would require the Loss of Power instruments
to be OPERABLE in Operational Conditions 4 and 5 only when the
associated diesel generator is required to be OPERABLE. The Loss of
Power relays provide a support function to initiate the associated
diesel generator start and bus unloading sequences. If that diesel
generator is not in service, the loss of power relays perform no
safety function. Therefore, relating diesel generator OPERABILITY
and Loss of Power instrument OPERABILITY will not involve an
increase in the probability of an accident previously evaluated.
The proposed change does not affect the requirements of ESF
OPERABILITY. The change does not affect diesel generator response to
a loss of voltage or degraded voltage on the Divisional 4.16 kV
electrical busses when the diesel generator is required to be
OPERABLE. Automatic response of the ESF functions is unaffected by
removing the Loss of Power relays from service under these
conditions, therefore, the proposed change will not involve a
significant increase in the consequences of an accident previously
evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does not involve a modification of plant
equipment nor does it change the way the equipment will be
maintained or operated. The revision to Technical Specifications
will continue to require the Loss of Power instrumentation to be
OPERABLE when the associated diesel generator is required to be
OPERABLE. The Loss of Power instruments will continue to perform
their safety function of initiating the diesel generator start and
bus unloading sequences.
Therefore, this proposed change will not create the possibility
of a new or different kind of accident from any previously
evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed change will not affect the OPERABILITY, operation
or reliability of any ESF function including the diesel generators.
All ESF functions will remain available during postulated accidents
with a loss of offsite electrical power. The change simply clarifies
when the Loss of Power instruments are required to be OPERABLE
during Operational Conditions 4 and 5. Therefore, the proposed
change will not involve a significant reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Susan Frant Shankman, Acting.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: March 15, 1996.
Description of amendment request: The proposed amendment would
revise the surveillance requirements of Technical Specification (TS)
4.6.2.1 ``Containment Systems--Depressurization Systems--Suppression
Pool'' to extend the time interval for performing the containment
drywell-to-suppression chamber bypass leakage test from 18 months to an
interval corresponding to that required for the Containment Integrated
Leak Rate Test. The provisions of TS 4.0.2 (which would provide an
extension of up to 25% of the specified surveillance interval) will not
apply. Specifically, existing TS 4.6.2.1.d would become subparagraphs d
and e to require that the suppression pool be demonstrated operable:

d. At least once per 18 months by conducting a visual inspection
of the exposed accessible interior and exterior surfaces of the
suppression chamber.*
e. At least every outage by requiring the performance of a
Containment Integrated Leak Rate Test, as scheduled in conformance
with the criteria specified in the 10 CFR 50 Appendix J Testing
Program Plan described in Section 6.8.4.f, by conducting a drywell-
to-suppression chamber bypass leak test at an initial differential
pressure of 3 psi and verifying that the [drywell-to-suppression
chamber bypass flow area] A/the square root of K calculated from the
measured leakage is within the specified limit of 0.0054 square
feet.
1. If any drywell-to-suppression chamber bypass leak test fails
to meet the specified limit, the test schedule for subsequent tests
shall be reviewed and approved by the Commission.
2. If two consecutive tests fail to meet the specified limit, a
test shall be performed at least each refueling outage until two
consecutive tests meet the specified limit, at which time the
original test schedule may be resumed.
3. The provisions of Specification 4.0.2 do not apply.

*Includes each vacuum relief valve and associated piping.
The proposed changes would also add a new surveillance requirement
for the testing of the bypass leakage path containing the suppression
chamber vacuum breakers, with associated acceptance criteria, which
would be performed each refueling outage that the bypass leak test is
not performed. Specifically, a new TS 4.6.2.1f would require that the
suppression pool be demonstrated operable:

f. During each refueling outage for which the drywell-to-
suppression chamber bypass leak test in Specification 4.6.2.1.e is
not conducted, by conducting a test of the four drywell-to-
suppression chamber bypass leak paths containing the suppression
chamber vacuum breakers at a differential pressure of at least 3 psi
and
1. Verifying that the total leakage area A/the square root of K
contributed by all four bypass leak paths is less than or equal to
24% of the specified limit, and
2. The leakage area for any one of the four bypass leak paths is
less than or equal to 12% of the specified limit.

By separate action, the NRC has provided notice of a proposed
amendment to change the frequency of Containment Integrated Leak Rate
Tests in accordance with Option B of 10 CFR Part 50 Appendix J . The
proposed changes described herein are intended

[[Page 20852]]

to be consistent with the changes proposed under Option B.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The operation on Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed TS changes involve the drywell-to-suppression
chamber bypass leak test frequency. There are no physical or
operational changes to the plant as a result of these proposed TS
revisions. Furthermore, the primary containment acts as an accident
mitigator and not as an accident initiator. Therefore, the proposed
TS changes do not affect the probability of any previously evaluated
accident.
The continued testing of bypass leakage pathways containing the
suppression chamber vacuum breakers on a refueling frequency, and
the continued requirement for visual inspection of containment
structural features assures that the bypass leakage path will not
degrade beyond the TS allowable limit during the interval between
performance of the bypass leakage test. Therefore, radioactivity
release following an accident will not be increased since the
pressure suppression capability of the containment is not reduced
from the existing design, and there will be no significant increase
in the consequences of any accident previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed TS changes involve the drywell to suppression
chamber bypass leak test frequency. There are no physical or
operational changes as a result of these proposed TS changes. These
proposed TS changes also include a requirement to continue
performing a surveillance test on the bypass leakage pathways
containing the vacuum breaker assemblies each refueling outage for
which the drywell-to-suppression chamber test is not conducted. This
test, along with the visual inspection required every refueling
cycle, will ensure that acceptable bypass leakage is maintained
during those intervals when the bypass leak test is not required.
Accordingly, the possibility of a new or different type of accident
is not introduced. Therefore, the proposed TS changes do not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The drywell-to-suppression chamber bypass leak test data
obtained during previous testing at NMP2 [Nine Mile Point Unit 2]
demonstrates conformance, by a large margin, to the TS and design
leakage requirements. The test data and engineering evaluations
indicate that there is negligible risk that the bypass leakage will
change adversely in future years. Furthermore, the proposed test
frequency is judged to be acceptable based on the small risk of
bypass leakage through paths other than those containing the
suppression chamber vacuum breakers.
A test of the bypass leak pathways containing the vacuum
breakers will be used to verify acceptable bypass leakage during
those outages when the bypass leak test is not performed. The
proposed test of the bypass leak pathways containing the vacuum
breakers, with stringent acceptance criteria, combined with the
other negligible potential leakage areas provide an acceptable level
of assurance that the bypass leakage can be measured. This
capability ensures that an adverse condition can be detected and
corrected such that the existing level of confidence that the
primary containment will function as required during a LOCA [loss-
of-coolant accident] is maintained. Therefore, the proposed TS
changes do not involve a significant reduction in the margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Susan Frant Shankman, Acting.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: March 20, 1996.
Description of amendment request: The proposed amendment would
revise Tables 3.3.1-1 and 4.3.1-1 of Technical Specification 3/4.3.1
``Reactor Protection System Instrumentation'' to delete the operability
requirement for the Average Power Range Monitor (APRM) Neutron Flux-
Upscale, Setdown and Inoperative functions in Operational Conditions
(OCs) 3 (Hot Shutdown) and 4 (Cold Shutdown). These same functions
would also be revised for OC 5 (Refueling) to indicate that operability
will only be required during shutdown margin demonstrations performed
per TS 3.10.3.
Basis for proposed no significant hazards consideration
determination: The revisions to the APRM functions are proposed to
support licensee's plans to replace Local Power Range Monitors during
the next refueling outage. The revisions also provide for the eventual
replacement of the existing APRM System with the Nuclear Measurement
Analysis and Control Power Range Neutron Monitoring System, and the
eventual installation of the Oscillation Power Range Monitor system for
the detection of reactor instability conditions. These modifications
are based upon Report NEDO-31960, ``BWR Owners' Group Long-Term
Solutions Licensing Methodology, approved by the Commission July 12,
1993; the licensee's response of November 8, 1994, selecting Option III
in NEDO-31960 for Nine Mile Point, Unit 2; NRC Generic Letter 94-02,
``Long-Term Solutions and Upgrade of Interim Operating Recommendations
for Thermal-Hydraulic Instabilities in Boiling Water Reactors'' dated
July 11, 1994; and General Electric Licensing Topical Report, NEDC-
32410P-A, ``Nuclear Measurement Analysis and Control Power Range
Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip
Function,'' which was approved by the Commission September 5, 1995.
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 operation of Nine Mile Point Unit 2 in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The Reactor Protection System (RPS) initiates a reactor scram
when one or more monitored parameters exceed their specified limits
to preserve the integrity of the fuel cladding and the Reactor
Coolant System and to minimize the energy that must be absorbed
following a loss-of-coolant accident. The proposed changes will
revise the OCs in which the APRM Neutron Flux-Upscale, Setdown and
Inoperative RPS Instrumentation is required. These changes do not
affect the probability of precursors of any accidents previously
evaluated, and therefore, do not increase their probability.
During normal operation in OCs 3 and 4, all control rods are
fully inserted and the reactor mode switch position control rod
withdrawal blocks do not allow control rods to be withdrawn.
Therefore, the RPS APRM functions are not required. Specification
3.9.10 does allow one control rod to be removed from the core in OC
4 by placing the mode switch in the refuel position. However, with
the reactor mode switch in the refuel position, refueling interlocks
are in place (i.e., one-rod out, etc.), which together with

[[Page 20853]]

adequate shutdown margin will preclude unacceptable reactivity
excursions. The APRM Neutron Flux-Upscale, Setdown function is not
required during OC 5 except during shutdown margin demonstrations.
The SRMs [source range monitors], IRMs [intermediate range
monitors], and refueling interlocks provide adequate protection from
reactivity excursions during OC 5. The exception is during the
shutdown margin demonstration when more than one control rod will be
withdrawn and the APRMs will continue to be required to be operable
as a backup to the IRMs. Testing of the RPS APRM functions will
continue to be performed in those OCs for which operability is
required. Consequently, the reliability and performance of the RPS
APRM functions in these OCs will not be adversely affected.
Therefore, the proposed change will not result in a significant
increase in the consequences of any accidents previously evaluated.
The operation of Nine Mile Point Unit 2 in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes will revise the applicable OCs in which the
APRM neutron Flux-Upscale, Setdown and Inoperative RPS
instrumentation is required. Changes to OC requirements will not
introduce any new accident precursors and will not involve any
physical alternations to plant configurations which could initiate a
new or different kind of accident. NMP2 is analyzed for a single
control rod withdrawal error during refueling. Since the core is
designed to meet shutdown requirements with the highest worth rod
withdrawn, the core remains subcritical even with one rod withdrawn.
The one-rod-out interlock which allows only one control rod to be
withdrawn in OC 5 is not affected by the proposed changes.
Consequently, the proposed changes do not create an accident
different than the previously analyzed single control rod withdrawal
error event. Surveillance testing will continue to be performed to
assure reliability and maintain current performance levels.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any previously evaluated.
The operation of Nine Mile Point Unit 2 in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
The proposed changes to the RPS APRM function instrumentation
Technical Specification requirements will not adversely affect the
design or the performance characteristics of the RPS instrumentation
nor will it affect the ability of the RPS APRM instrumentation to
perform its intended function. As discussed above, the subject RPS
instrumentation is not required in OC 3, 4, and 5 except for
shutdown margin demonstrations. Accordingly, deletion of the
requirement to have these functions operable in these OCs will not
significantly reduce a margin of safety. Surveillance testing will
continue to be performed for those OCs in which the instrumentation
is required to assure reliability. Therefore, the proposed changes
do not involve a significant reduction in a margin of safety

.The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Susan Frant Shankman, Acting.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London, Connecticut

Date of amendment request: March 28, 1996.
Description of amendment request: The proposed amendment would
change Technical Specification Section 3.7.7, ``Sealed Source
Contamination,'' by making the criteria for testing sealed sources for
contamination and leakage at Millstone Unit No. 2 the same as those at
Millstone Unit No. 3, the Haddam Neck Plant, and Seabrook Station.
Specifically, the sealed sources that are required to be free of
greater than or equal to 0.005 microcuries of removable contamination
would be those that would exceed ``100 microcuries of beta and/or gamma
emitting material or 5 microcuries of alpha emitting material.'' The
Bases Section 3/4.7.7, ``Sealed Source Contamination,'' would also be
changed to reference the appropriate section of 10 CFR 70.39.
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 (SHC), which is presented below:

Pursuant to 10 CFR 50.92, NNECO [Northeast Nuclear Energy
Company] has reviewed the proposed changes and concludes that the
changes do not involve a significant hazards consideration (SHC)
since the proposed changes satisfy the criteria in 10 CFR 50.92(c).
That is, the proposed changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The changes make the criteria for testing sealed sources for
contamination and leakage at Millstone Unit No. 2 the same as those
at Millstone Unit No. 3, the Haddam Neck Plant and Seabrook Station.
Although the leakage criteria for sealed sources that are to be
tested is being changed, the allowable leakage remains small. Any
leakage that is identified would not cause a significant radiation
exposure. The source storage area is routinely surveyed by Health
Physics in accordance with Health Physics Department procedures and
any significant leakage would be detected. Therefore, this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change in the criteria for testing sealed sources
for contamination and leakage will not change the way the sources
are used. Therefore, this change will not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The possible radiation exposure to both the workers and the
public from this change is very small. All protective systems which
would detect any release of material from the site remain in place
so there is no reduction in safety for the public. Likewise, all
protective systems for the workers remain in place. Workers using
the sources routinely pass through the whole body contamination
monitors. In addition, the source storage areas are surveyed
routinely by Health Physics in accordance with Health Physics
Department procedures, and any significant leakage would be
detected. The bases section is being revised to reference the
appropriate section of 10 CFR 70.39. Therefore, there is no
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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and Waterford Library, Attn: Vince Juliano, 49 Rope
Ferry Road, Waterford, CT 06385.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee.

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

Date of amendment request: March 12, 1996.

[[Page 20854]]

Description of amendment request: The proposed changes would remove
a requirement to interconnect two or more accumulators for the purpose
of cross checking instrumentation in the event that one of the two
pressure or level instrument channels on an accumulator is declared
inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response: The design basis accident for which the accumulators
were designed is the double ended guillotine break of a cold leg.
Interconnecting or not interconnecting accumulators does not have
any effect on the probability of occurrence of this event. By
eliminating the requirement to interconnect accumulators, the
proposed amendment assures that a minimum of three accumulators are
available, as assumed in the safety analyses, to mitigate the
consequences of a large-break loss-of-coolant [LBLOCA] accident.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: The proposed amendment does not involve any physical
changes to plant equipment or setpoints and does not create the
possibility of a new or different kind of accident. Eliminating the
requirement to interconnect accumulators ensures that the plant
configuration is maintained consistent with that assumed in the
safety analysis and no new failure modes are created.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?

Response: There is no margin of safety specified in the
Technical Specifications for these instrument channels. There are no
setpoints or allowable values associated with these instrument
channels which affect Safety Limits or Limiting Safety System
Settings. The proposed amendment ensures that the safety analysis
assumption regarding the accumulators remains valid and the
resulting peak fuel clad temperature meets specified acceptance
criteria. The proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Susan Frant Shankman, Acting.

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

Date of amendment request: March 14, 1996.
Description of amendment request: The proposed changes would allow
a one-time extension of the inspection interval for the steam generator
tubes that is due in July 1996.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: The proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated. As stated in the Basis of the IP3
[Indian Point Unit 3] Technical Specifications, the program for
inservice inspection of steam generator tubes regarding equipment,
procedures, and sample selection is based upon the guidance and
recommendations in Regulatory Guide 1.83 and NRC Generic Letter 85-
02. The addition of the footnote to extend the surveillance due date
will not increase the deviation from the guidance and recommendation
stated above, and, therefore will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: The proposed license 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
addition of any new or different type of equipment, nor does it
involve the operation of equipment required for safe operation of
the facility in a manner different from those addressed in the Final
Safety Analysis Report. Therefore, the proposed change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: The proposed license amendment does not involve a
significant reduction in a margin of safety. The proposed change
does not adversely affect any safety related system or component
operation or operability, instrument operation, or safety system
setpoints and does not result in increased severity of any of the
accidents considered in the safety analysis. This change has no
adverse effect on any margin of safety and, therefore, does not
create a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Susan Frant Shankman, Acting

Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: March 22, 1996.
Description of amendment request: The amendment proposes changes to
the Technical Specifications (TS) to establish operability requirements
for avoidance and protection from thermal hydraulic instabilities to be
consistent with Boiling Water Reactor Owners Group long-term solution
Option I-D. Editorial changes are also made to support the revised
specifications, improve readability of Bases sections, and enhance the
presentation of requirements for single loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Operation of the FitzPatrick 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 because:
The implementation of BWR Owners' Group long-term stability
solution Option I-D at FitzPatrick does not modify the assumptions
contained in the existing accident analysis. The use of an exclusion
region and the operator actions required to avoid and minimize
operation inside the region do not increase the possibility of an
accident. Conditions of operation outside of the exclusion region
are within the analytical envelope of the existing safety analysis.
The operator action requirement to exit the

[[Page 20855]]

exclusion region upon entry minimizes the possibility of an
oscillation occurring. The actions to drive control rods and/or to
increase recirculation flow to exit the region are maneuvers within
the envelope of normal plant evolutions. The flow referenced scram
has been analyzed and will provide automatic fuel protection in the
event of an instability. Thus, each proposed operating requirement
provides defense in depth for protection from an instability event
while maintaining the existing assumptions of the accident analysis.
2. Create the possibility of a new or different kind of accident
from those previously evaluated because:
The proposed operating requirements either mandate operation
within the envelope of existing plant operating conditions or force
specific operating maneuvers within those carried out in normal
operation. Since operation of the plant with all of the proposed
requirements are within the existing operating basis, an unanalyzed
accident will not be created through implementation of the proposed
change.
3. Involve a significant reduction in the margin of safety
because:
Each of the proposed requirements for plant thermal hydraulic
stability provides a means for fuel protection. The combination of
avoiding possible unstable conditions and the automatic flow
referenced reactor scram provides an in depth means for fuel
protection. Therefore, the individual or combination of means to
avoid and suppress an instability supplements the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Susan Frank Shankman, Acting.

Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: March 22, 1996.
Description of amendment request: The amendment proposes to revise
Technical Specification (TS) Table 3.2-2, ``Core and Containment
Cooling System Initiation and Control Instrumentation Operability
Requirements.'' The proposed changes will revise allowed outage times
(AOTs) for 4kV Emergency Bus Undervoltage Trip Functions. The AOTs for
these trip functions were extended by Amendment 227; however, the AOT
extensions for these trip functions were not consistent with the
requirements of Standard Technical Specifications (STS), NUREG-1433,
and differed from the recommendations in the associated Licensing
Topical Report. Additional changes are proposed to TS Table 3.2-2 and
to TS Table 4.2-2, ``Core and Containment Cooling System
Instrumentation Test and Calibration Requirements.'' These changes
will: (1) replace the generic actions for inoperable instrument
channels with function-specific actions, (2) replace the generic test
AOT with function-specific test AOTs, and (3) relocate selected trip
functions from the TS to an Authority controlled document.
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:

Operation of the FitzPatrick 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 limited to replacement of the generic
actions and test AOT with function-specific actions and test AOTs,
and relocation of selected trip functions from the TS to an
Authority controlled document. The changes do not introduce any new
modes of plant operation, make any physical changes, or alter any
operational setpoints. Therefore, the changes do not degrade the
performance of any safety system assumed to function in the accident
analysis. Consequently, there is no effect on the probability or
consequences of an accident.
2. Create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed changes do not introduce any new accident
initiators or failure mechanisms since the changes do not introduce
any new modes of plant operation, make any physical changes, or
alter any operational setpoints. Therefore the changes do not create
the possibility of a new or different kind of accident.
3. Involve a significant reduction in the margin of safety.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The relocated requirements do not satisfy
the 10 CFR 50.36 criteria for inclusion in the Technical
Specifications. Therefore, the changes do not involve a significant
reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Susan Frant Shankman, Acting.

Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: March 27, 1996.
Description of amendment request: The amendment proposes to revise
the Technical Specifications to support adoption of the primary
containment leakage rate testing requirements of Option B to 10 CFR 50,
Appendix J at the FitzPatrick plant, and clarify the numerical value of
the allowable containment leakage rate (La) as 1.5 percent per
day.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The Authority has evaluated the proposed TS Amendment and
determined that it does not represent a significant hazards
consideration. Based on the criteria for defining a significant
hazards consideration established in 10 CFR 50.92, operation of the
James A. FitzPatrick Nuclear Power Plant in accordance with the
proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The proposed changes do not involve a change to the design or
operation of the plant. The systems affected by this proposed TS
change are not assumed in any safety analyses to initiate any
accident sequence. Therefore, the probability of any accident
previously evaluated is not increased by this proposed TS change.
The clarification of the allowable containment leakage rate
(La) is consistent with the accident analyses. There is no
change to the consequences of an accident previously evaluated
because maintaining leakage within limits assumed in the accident
analyses ensures that the dose consequences resulting from an
accident are not increased. The proposed TS changes maintain an
equivalent level of reliability

[[Page 20856]]

and availability for all affected systems. The ability of the
affected systems associated with maintaining leak rate integrity to
perform their intended function is unaffected by the proposed TS
changes. Implementation of these changes will provide continued
assurance that specified parameters associated with containment
integrity will remain within acceptance limits, and as such, will
not significantly increase the consequences of a previously
evaluated accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The proposed changes allow adoption of those requirements
specified in Option B to 10 CFR 50, Appendix J, and do not involve a
change to the plant design and operation. As a result, the proposed
changes do not affect the parameters or conditions that could
contribute to the initiation of any accidents. The methods of
performing primary containment leakage rate testing are not changed.
No new accident modes are created by allowing extended intervals for
Type A, B and C testing, or by clarifying the numerical value of the
allowable containment leakage rate (La). No safety-related
equipment or safety functions are altered, or adversely affected, as
a result of these changes. The proposed changes will not introduce
failure mechanisms beyond those already considered in the current
plant safety analyses. Extension of the test intervals, and
clarification of the allowable leakage rate, does not contribute to
the possibility of a new or different kind of accident or
malfunction from those previously analyzed.
3. Involve a significant reduction in the margin of safety
because: The proposed changes affect the frequency of primary
containment leakage rate testing, and the numerical definition of
the allowable containment leakage rate (La). The design of the
FitzPatrick plant is not changed. The methodology for test
performance is unchanged and Type A, B and C tests will continue to
be performed at Pa. The proposed changes provide
sufficient controls to ensure that proper maintenance and repairs
are performed on the primary containment, and systems and components
penetrating the primary containment. The reliability of containment
systems assumed to operate in the plant safety analyses is not
reduced. The numerical value of La specified in Specification
6.20 is consistent with the accident analyses, therefore, the dose
consequences of any analyzed accidents are not increased. Therefore,
the proposed changes provide continued assurance of the leak
tightness of the containment without adversely affecting the public
health and safety and, as such, will not involve a significant
reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Susan Frant Shankman, Acting.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey

Date of amendment request: April 22, 1996.
Description of amendment request: The amendments would change the
Technical Specifications to implement 10 CFR Part 50, Appendix J,
Option B, for the Type A test by referring to Regulatory Guide 1.163,
``Performance-Based Containment Leakage-Test Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Containment leak rate testing is not an initiator of any
accident. The proposed changes do not make any physical changes to
the containment. The proposed changes do not affect performance of
the containment, reactor operations or accident analysis. Therefore,
the proposed changes will not involve an increase in the probability
of any previously evaluated accident.
Since the allowable leakage rate is not being changed and since
the analysis documented in NUREG-1493, ``Performance-Based
Containment Leak-Test Program'' concludes that the impact on public
health and safety due to extended intervals is negligible, the
proposed changes will not involve an increase in the consequences of
any previously evaluated accident. Therefore, adoption of a
performance-based verification of leakage rates for the overall
containment boundary will provide an equivalent level of safety and
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change makes no physical changes to the plant.
Since no physical changes are involved and since the analysis
documented in NUREG-1493 confirms that the performance based
schedule continues to maintain a minimal impact on public risk, it
can be concluded that the effect of the containment on any accident
will not change. The proposed change does not affect normal plant
operations or configuration, nor does it affect leak rate test
pressure.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes are based on NRC-accepted provisions, and
maintain necessary levels of reliability of containment integrity.
The performance-based approach to leakage rate testing recognizes
that historically good results of containment testing provide
appropriate assurance of future containment integrity. This supports
the conclusion that the impact on the health and safety of the
public as a result of extended test intervals is negligible. Since
the analysis documented in NUREG-1493 confirms that the performance
based schedule continues to maintain a minimal impact on public
risk, it can be concluded that the margin of safety is not
significantly affected by the proposed changes.
The test history at Salem Units 1 and 2 (no ILRT failures)
provides continued assurance of the leak tightness of the
containment structure.
Therefore, the proposed amendment will not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: April 4, 1996 (TS 96-01).
Description of amendment request: The proposed change would revise
the appropriate technical specifications, surveillances, and bases as
needed for the conversion from Westinghouse nuclear fuel to Framatome
Cogema Mark-BW17 nuclear fuel.
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:

[[Page 20857]]

TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The analyses provided in Topical Report BAW-10220P show that the
changes do not significantly change the results of previously
evaluated events. These analyses provide the template for accident
analyses assumptions that must be met by the cycle-specific reload
analysis.
The SQN Units 1 and 2 Cycle 9 reload cores with Mark-BW fuel
will be designed to operate within the approved limits for accident
analysis. The limits provided in the TS and described in the Updated
Final Safety Analysis Report (UFSAR) provide the framework for
accident analyses. By maintaining these limits, the probability or
consequences of accidents related to the core changes do not
significantly change. Thus, it is concluded that there is no
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The change to Mark-BW fuel cores and mixed (transition) cores
has been evaluated in the Topical Report BAW-10220P. It was
concluded that the change did not create new or different kinds of
accidents. The change in fuel suppliers has been evaluated for
consideration of the effects of power distribution and peaking
factors such that there are no restrictions on the use of Mark-BW
fuel assemblies beyond those already established in the UFSAR and
TS. Adherence to the safety analysis limits restricts the
possibility of new or different accidents. Historically, new
accidents have not been associated with changes in fuel suppliers as
long as safety analysis limits continue to be met. It is concluded
that transition to Mark-BW fuel does not create the possibility of a
new or different kind of accident from any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The margin of safety is established by the acceptance criteria
used by NRC. Meeting the acceptance criteria assures that the
consequences of accidents are within known and acceptable limits.
The loss-of-coolant accident (LOCA) acceptance criteria are
unchanged: peak cladding temperature of 2200 degrees
Fahrenheit, peak cladding oxidation of 17 percent,
average clad oxidation of 1 percent, and long-term
coolability. These requirements continue to be met. The methods used
to demonstrate conformance with these limits have changed, and were
reviewed to assure that the methods, as well as the results, are
acceptable. The acceptance criteria for Departure from Nucleate
Boiling (DNB) events has not changed and is still the 95 percent
probability and 95 percent confidence interval that DNB is not
occurring during the transient. The DNB correlation, and methods
used to demonstra

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A96-11295. Public record. Not legal advice.