95-10127. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

[Federal Register Volume 60, Number 80 (Wednesday, April 26, 1995)]
[Notices]
[Pages 20513-20539]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-10127]

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NUCLEAR REGULATORY COMMISSION

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

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 31, 1995, through April 14, 1995. The
last biweekly notice was published on Wednesday, April 12, 1995 (60 FR
18621).

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 [[Page 20514]] margin of
safety. The basis for this proposed determination for each amendment
request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By May 26, 1995, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee. [[Page 20515]]
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.

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: March 20, 1995.
Description of amendment request: The licensee proposes a revision
to Technical Specification (TS) 2.2.1, Reactor Trip System
Instrumentation Setpoints, and to relocate cycle specific Overpower and
Overtemperature Delta T trip setpoint parameters to the Core Operating
Limits Report (COLR).
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 change does not involve a significant hazards consideration
for the following reasons:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change of relocating Overtemperature Delta T * * *
and the Overpower Delta T * * * trip setpoint parameters to the COLR
has no influence or impact to the probability or consequences of an
accident. The revised TS will continue to implement the Reactor Trip
System Instrumentation [Overtemperature Delta T] and [Overpower
Delta T] setpoint limits through reference to the parameters in the
COLR. In addition, the COLR is subject to the existing controls of
TS 6.9.1.6, including the establishment of the parameter values
using an NRC approved methodology. Given that this change
administratively relocates the selected trip setpoint parameter
values to another TS-controlled document, there would be no increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
No safety-related equipment, safety function, or plant operation
will be altered as a result of this proposed change. The limits are
simply being relocated to another TS-controlled document. The TS
will continue to require operation within the required limits as
established per NRC approved methodologies. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
Relocation of the Reactor Trip System Instrumentation
[Overtemperature Delta T] and [Overpower Delta T] setpoint limits to
the TS-controlled COLR has no effect on the trip system setpoints
currently in force in TS 2.2.1. Future revisions to the trip
setpoint parameters are governed by TS 6.9.1.6. TS 6.9.1.6 lists
each TS that references values in the COLR and the NRC approved
methodologies utilized in developing those values. Since this change
is only an administrative relocation of the selected trip setpoint
parameter values to another TS controlled document, the proposed
changes do not involve a significant reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
NRC Project Director: David B. Matthews.

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: March 30, 1995.
Description of amendment request: The licensee proposes to revise
the Emergency Diesel Generator (EDG) surveillance requirements
contained in Technical Specification (TS) 4.8.1.1.2 to be consistent
with NUREG-1431, Standard Technical Specifications for Westinghouse
Plants, and to eliminate the need for duplicate EDG testing that has
already been implemented to satisfy the requirements of the Station
Blackout Rule and the Maintenance Rule.
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 change does not involve a significant hazards consideration
for the following reasons:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
A failure of the Emergency Diesel Generator (EDG) is not an
initiator for any previously evaluated FSAR Chapter 15 accident
scenario. By committing to and implementing an EDG reliability
program that satisfies the requirements of the Station Blackout Rule
and the Maintenance Rule, the Shearon Harris Nuclear Power Plant
(SHNPP) will continue to ensure that target EDG reliability and
availability is being achieved by conducting appropriate monitoring,
testing, and maintenance activities. This program will be developed
and controlled as a Plant Operating Manual procedure and will
incorporate industry, vendor, and TDI Owners Group recommendations.
Therefore, with commensurate levels of testing and inspection in
place to provide assurance that the EDGs will perform their intended
safety function in the event of an accident, the proposed changes
will have no effect on the probability or consequences of such an
accident.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
A failure of the EDG is not an initiator for any previously
evaluated FSAR Chapter 15 accident scenario nor would the proposed
changes to the EDG surveillance requirements result in the
possibility of a new or different kind of accident from any accident
previously evaluated. By committing to and implementing an EDG
reliability program that satisfies the requirements of the Station
Blackout Rule and the Maintenance Rule, SHNPP will continue to
ensure that target EDG reliability and availability is being
achieved by conducting appropriate monitoring, testing, and
maintenance activities. This program will be developed and
controlled as a Plant Operating Manual procedure and will
incorporate industry, vendor, and TransAmerica Delaval Inc. Owners
Group recommendations. Therefore, with commensurate levels of
testing and inspection in place to provide assurance that the EDGs
will perform their intended safety function in the event of an
accident, the proposed changes would not increase the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes will not affect any parameters which relate
to the margin of safety as defined in the Technical Specifications
or the FSAR. Testing, inspection and maintenance necessary to verify
the EDGs' ability to perform their intended safety function will
continue to be [[Page 20516]] performed. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of amendment request: March 24, 1995.
Description of amendment request: The proposed amendments would
acknowledge the acceptability of performing containment leakage rate
testing in accordance with 10 CFR Part 50, Appendix J, and all approved
exemptions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed amendment does not involve a significant
increase in the probability of occurrence or consequences of any
accident previously evaluated.
The proposed changes to Technical Specifications to add an
allowance to test in accordance with approved exemptions to 10 CFR
50 Appendix J are administrative in nature and will not affect any
accident initiators or precursors. 10 CFR 50 Appendix J provides the
requirements to periodically test the primary reactor containment.
The objective of these requirements is to ensure that leakage from
the primary reactor containment structure and systems and components
that penetrate the containment is maintained below the limits
established for containment leakage. The performance of periodic
integrated leakage rate testing (Type A) and local penetration
testing (Type B and C) during containment life provides a current
assessment of potential leakage from containment during accident
conditions.
10 CFR 50.12 allows the Commission to grant specific exemptions
to the requirements of 10 CFR 50 Appendix J when those exemptions
are authorized by law, will not present undue risk to the public,
and are consistent with the common defense and security. In
addition, special circumstances must exist as described in Section
50.12. Since all exemptions to 10 CFR 50 Appendix J receive NRC
review and approval prior to being implemented, all containment
leakage rate testing will continue to be performed in accordance
with NRC approved methodologies when relying upon the allowance that
is added to the Technical Specifications by the proposed amendment.
The proposed changes are consistent with the requirements provided
in NUREG-1431, ``Standardized Technical Specifications, Westinghouse
Plants'' which has been approved by the NRC.
The proposed changes will not affect any accident initiators or
precursors and will not change or alter the design assumptions for
the systems used to mitigate the consequences of an accident. The
proposed changes do not involve the addition of any new or different
type of equipment, nor do they involve the operation of equipment
required for safe operation of the facility in a manner different
from those addressed in the UFSAR. There are no changes to
parameters governing plant operation as a result of the proposed
changes. The results and conclusions in the Zion Updated Final
Safety Analysis Report (UFSAR) are unaffected by this proposed
License Amendment.
Based on the previous discussion, the proposed changes do not
involve a significant increase in the probability of occurrence or
consequences of any accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any previously analyzed.
The proposed changes to Technical Specifications to add an
allowance to perform containment leakage rate testing in accordance
with approved exemptions to 10 CFR 50 Appendix J are administrative
in nature and do not involve the addition of any new or different
types of safety related 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 safety analyses. The proposed
changes may only affect the methods used to perform containment
leakage rate testing while in a shutdown condition. No safety
related equipment or function will be altered as a result of the
proposed changes. Also, the procedures governing normal plant
operation and recovery from an accident are not changed by the
proposed Technical Specification changes. Since no new failure modes
or mechanisms are added by the proposed changes, the possibility of
a new or different kind of accident is not created.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Given the proposed changes to Technical Specifications, testing
would be allowed in accordance with approved exemptions to Appendix
J. Exemptions are allowed by the Commission in accordance with 10
CFR 50.12 when it is shown that the exemption is authorized by law,
will not present an undue risk to the public health and safety, and
is consistent with the common defense and security. In addition,
special circumstances must exist.
The proposed changes will not impact any margin of safety and
testing in accordance with approved exemptions will not involve a
significant reduction in a level of safety since containment leakage
testing is performed while in a shutdown condition. In addition, it
is likely that any test methodology that significantly reduces a
margin of safety would not be approved by the NRC.
The ability to safely shut down the operating unit and mitigate
the consequences of all accidents previously evaluated will be
maintained. Therefore, the margin of safety is not significantly
affected.

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: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
NRC Project Director: Robert A. Capra.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan

Date of amendment request: December 15, 1994.
Description of amendment request: The proposed amendment would
revise Technical Specification 11.3.1.5 ACTION a. to eliminate the need
to demonstrate that the actuation circuitry of the unaffected reactor
depressurization system channels is operable. In addition, an editorial
change correcting a typographical error is also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change will eliminate the probability of a
subsystem failure caused by additional testing (which unnecessarily
introduces the potential for human and equipment problems),
therefore eliminating the probability that the facility would have
to be challenged and brought to the SHUTDOWN condition within 12
hours and to the COLD SHUTDOWN condition within the following 24
hours.
[[Page 20517]]

2. Will the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not alter the plant configuration,
systems, components, or operation; and does not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed change is expected to eliminate
unnecessary challenges to a safety system that has already been
determined to be operable by routine surveillance testing; therefore
contributing to the overall safe operation of the facility.
3. Will the proposed change involve a significant reduction in
the margin of safety?
The RDS [Reactor Depressurization System] provides for both
manual and automatic depressurization of the primary system to allow
injection of the core spray following a small-to-intermediate size
break in the primary system. This will allow core cooling with the
objective of preventing excessive fuel clad temperatures. The design
of the system is based on the specified initiation set points
described in the Technical Specifications. Transient analysis
demonstrated that these conditions result in adequate safety margins
for both the fuel and the system pressure. The proposed change does
not affect these setpoints, therefore the margin of safety is not
changed.

In addition, the proposed editorial change to correct a
typographical error is administrative in nature and, therefore, would
have no effect on the three standards of 10 CFR 50.92 discussed above.
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: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: Cynthia A. Carpenter, Acting.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: January 26, 1995, as supplemented March
9, 1995.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to increase the allowable
nominal fuel enrichment from 4.2 to 5.0 weight percent for reload fuel
assemblies. TS impose a limit on fuel enrichment of stored fuel
assemblies to prevent inadvertent criticality. Presently, the Crystal
River Unit 3 (CR3) TS specify a maximum enrichment of 4.5 weight
percent for storage pool A and dry fuel (new fuel) storage racks, and
4.2 weight percent for fuel pool B. The licensee proposed to revise TS
3.7.15, 4.2, and 4.3, and associated TS bases to allow increasing the
enrichment limits from 4.2 to 5.0 weight percent for the dry fuel
storage racks and for A and B fuel pools. Additionally, a typographical
error in TS 4.3.1.2.b will also be corrected.
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. This amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
An increase in fuel enrichment will not by itself affect the
mixture of fission product nuclides. A change in fuel cycle design
which makes use of an increased enrichment may result in fuel burnup
consisting of a somewhat different mixture of nuclides. The effect
in this instance is insignificant because:
a. The isotopic mixture of the irradiated assembly is relatively
insensitive to the assembly's initial enrichment.
b. Most accident doses are such a small fraction of 10 CFR 100
limits, a large margin exists before any change becomes significant.
c. The change in Pu content which would result from an increase
in burnup would produce more of some fission product nuclides and
less of other nuclides. Small increases in some doses are offset by
reductions in other doses. The radiological consequences of
accidents are not significantly changed.
2. This amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
An unplanned criticality event will not occur as Keff
[effective neutron multiplication factor] will not exceed 0.95 with
the maximum allowable enriched fuel in Pool A and Pool B, when
flooded with unborated water, and Keff will not exceed 0.98 in
the new fuel storage racks assuming conditions of optimum
hypothetical low density moderation. The new fuel storage racks have
two rows of storage cells physically blocked to ensure reactivity
limits are not exceeded. Administrative controls assure fuel is
stored in configurations which meet the requirements of the safety
analysis.
3. This amendment will not involve a significant reduction in a
margin of safety.
While the increased enrichment in Pool A, Pool B, and the dry
storage racks may lessen the margin to criticality, this reduction
is not significant because the overall safety margin is within NRC
criteria of Keff [less than or equal to] 0.95 (NRC Standard
Review Plan, Section 9.1.2.)
Therefore, this amendment request satisfies the criteria
specified in 10 CFR 50.92 for amendments which do not involve a
significant hazards consideration.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629.
Attorney for licensee: A. H. Stephens, General Counsel, Florida
Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida
33733.
NRC Project Director: David B. Matthews.

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

Date of amendment request: March 16, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specification 4.6.1.2, regarding the overall
integrated containment leakage rate tests, so that it would reference
10 CFR Part 50, Appendix J directly, rather than paraphrase the
regulation, and allow approved exemptions to the test frequency
requirements. In addition, there is an associated proposed exemption,
from the requirements of 10 CFR Part 50, Appendix J, to provide a one-
time interval extension for the Unit 2 Type A test (containment
integrated leak rate test) from the current scheduled 48 months to
approximately 66 months.
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:

Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated

The proposed change specific to Unit 2 will provide a onetime
exemption from the 10 CFR 50, Appendix J Section III.D.I.(a) leak
rate test schedule requirement. This change will allow for a one-
time test interval for Type A Integrated Leak Rate Tests of
approximately 66 months.
Leak rate testing is not an initiating event in any accident;
therefore, this proposed [[Page 20518]] change does not involve a
significant increase in the probability of a previously evaluated
accident.
Type A tests are capable of detecting both local leak paths and
gross containment failure paths. Experience at South Texas Project
Unit 2 demonstrates that excessive containment leakage paths are
local leakage detected by Type B and C Local Leak Rate Tests.
Administrative controls govern maintenance and testing of
containment penetrations such that the probability of excessive
penetration leakage due to improper maintenance or valve
misalignment is very low. Following maintenance on any containment
penetration, a Local Leak Rate Test is performed to ensure
acceptable leakage levels. Following a Local Leak Rate Test on a
containment isolation valve, an independent valve alignment check is
performed. Therefore, Type A testing is not necessary to ensure
acceptable leakage rates through containment penetrations.
While Type A testing is not necessary to ensure acceptable
leakage rates through containment penetrations, Type A testing is
necessary to demonstrate that there are no gross containment
failures. Structural failure of the containment is considered to be
a very unlikely event, and in fact, since South Texas Project Unit 2
has been in operation, it has successfully passed each Type A
Integrated Leak Rate Test. Therefore, a one-time exemption
increasing the interval for performing an Integrated Leak Rate Test
results [sic] in a significant decrease in the confidence in the
leak tightness of the containment structure. Therefore, this change
does not involve a significant increase in the consequences of an
accident previously evaluated.
The proposed amendment revised Technical Specification 4.6.1.2
to reference the testing frequency requirements of 10 CFR 50,
Appendix J, and to state that Nuclear Regulatory Commission approved
exemptions to the applicable regulatory requirements are permitted.
This portion of the proposed change is applicable to Unit 1 and Unit
2. The current language of Technical Specification 4.6.1.2
paraphrases the requirements of Section III,D.I.(a) [sic] of
Appendix J. The proposed administrative revision simply deletes the
paraphrased language and directly references Appendix J. No new
requirements are added, nor are any existing requirements deleted.
Any specific changes to the requirements of Section III.D.I.(a) will
require a submittal from Houston Lighting & Power under 10 CFR 50.12
and subsequent review and approval by the Nuclear Regulatory
Commission prior to implementation.
The proposed amendment, in itself, does not affect reactor
operations or accident analysis and has no radiological
consequences. The change provides clarification so that future
Technical Specification changes will not be necessary to correspond
to applicable Nuclear Regulatory Commission-approved exemptions from
the requirements of Appendix J.
Therefore, this proposed amendment does not involve a
significant increase in the probability or consequences of any
accident previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident From Any Previously Evaluated

The proposed Unit 2 exemption request does not affect normal
plant operations or configuration, nor does it affect leak rate test
methods. The proposed change allows a one-time test interval of
approximately 66 months for the Integrated Leak Rate Test. Because
the test history of South Texas Project Unit 2 demonstrates no Type
A test failures during plant lifetime, the relaxation in schedule
should not significantly decrease the confidence in the leak
tightness of the containment.
The proposed Technical Specification amendment for Units 1 and 2
provides clarification to a specification that paraphrases a
codified requirement.
Since the proposed change and amendment would not change the
design, configuration or method of operation of the plant, they
would not create the possibility of a new or different kind of
accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety

The purpose of the existing schedule for Integrated Leak Rate
Tests is to ensure that release of radioactive materials will be
restricted to those leak paths and leak rates assumed in accident
analyses. The relaxed schedule for Integrated Leak Rate Tests does
not allow for relaxation of Type B and C Local Leak Rate Tests.
Therefore, methods for detecting local containment leak paths and
leak rates are unaffected by this proposed change. A one-time
increase of the South Texas Project Unit 2 test interval does not
leak to a significant probability of creating a new leakage path or
increased leakage rates because the test history for Integrated Leak
Rate Tests shows no failure during plant life. The margin of safety
inherent in existing accident analyses is maintained.
The proposed Technical Specification amendment for Units 1 and 2
is administrative and clarifies the relationship between the
requirements of Technical Specification 4.6.1.2, Appendix J, and any
approved exemptions to Appendix J. It does not, in itself, change a
safety limit, a Limiting Condition of Operation, or a surveillance
requirement on equipment required to operate the plant. Nuclear
Regulatory Commission approval of any proposed change or exemption
to III.D.1.(a) of Appendix J will be required prior to
implementation.
Therefore, this change and amendment 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: William D. Beckner.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa

Date of amendment request: March 10, 1995.
Description of amendment request: The proposed amendment would
remove redundant Limiting Conditions of Operation and Surveillance
Requirements for the containment hydrogen and oxygen monitors in the
Technical Specifications (TS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. No physical changes will result from this
amendment. This change deletes requirements that are redundant and
unduly restrictive. The annual surveillance deleted by this
amendment is redundant to the semi-annual surveillance required in
Table 4.2-H. The Limiting Conditions for Operation are not changed
by the proposed amendment.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No physical changes will result from this amendment.
Functional tests are performed on the hydrogen and oxygen analyzers
semiannually as required in TS Table 4.2-H. Deleting the annual
requirement for a functional test of the same equipment will not
reduce the amount of testing performed or increase the possibility
of degraded equipment being undetected.
3. The proposed amendment does not involve a significant
reduction in a margin of safety. No physical changes will result
from this amendment. The existing requirement for a semi-annual test
of the hydrogen and oxygen analyzer in Table 4.2-H exceeds the
requirements to be deleted in Section 3.7/4.7-H. The frequency of
testing of the hydrogen and oxygen analyzers will not be reduced as
a result of this amendment.

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.
[[Page 20519]]
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis
& Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: Gail H. Marcus.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa

Date of amendment request: March 28, 1995.
Description of amendment request: The proposed amendment would
revise and clarify Technical Specification Table 3.2-A that lists
allowable out-of-service times and surveillance test intervals for
instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed changes to TS Table 3.2-A will not significantly
increase the probability or consequences of an accident previously
evaluated. The changes do not alter the physical design or operation
of the plant and serve to describe more accurately and clearly the
actual logic configurations. The existing logic designs are in
conformance with the Architect/Engineer's design documentation since
plant startup. These changes will assure that the information in the
tables is clearer and more consistent with the column headings of
the table. The proposed changes do not affect assumptions contained
in the plant safety analysis.
The Bases changes provide additional information about the logic
arrangements as appropriate to identify unique or different logic
configurations. Changes to the Allowed Outage Time (AOT)
descriptions for the MSL Flow--High and MSL Tunnel Temperature--High
provide clarification regarding application of the AOT to these
logic arrangements, since multiple instrument channels provide input
into multiple logic channels. This application conforms to the
single failure criterion of the design basis (NEDO-10139, Compliance
of Protection Systems to Industry Criteria: General Electric BWR
Nuclear Steam Supply System, dated June 1970) and to the analytical
basis for the TS (NEDC-31677P-A, Technical Specification Improvement
Analysis for BWR Isolation Actuation Instrumentation, dated July
1990).
2. The proposed changes to Table 3.2-A will not introduce a new
or different kind of accident from any accident previously
evaluated. The changes do not alter the physical design of the plant
or affect any modes of operation. The proposed changes serve to
clarify the existing information to better assure that the trip
instrumentation will be maintained as assumed in the accident
analyses contained in the Updated Final Safety Analysis Report.
3. The proposed changes do not involve a significant reduction
in a margin of safety. Clarification of the logic arrangements in
both Table 3.2-A and the TS Bases and how the AOT is applied does
not affect the ability of the isolation logic to perform its
intended function. No physical changes to the plant are being made
as part of this amendment.

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: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis
& Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: Gail H. Marcus.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit No. 1, Berrien County, Michigan

Date of amendment request: March 17, 1995.
Description of amendment request: The proposed amendment would
defer performance of the Type A containment integrated leakage rate
test until the next refueling outage.
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:

Per 10 CFR 50.92, a proposed change does not involve a
significant hazards consideration if the change does 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.

Criterion 1

The Cook Nuclear Plant Type A test history provides substantial
justification for the proposed test schedule. Three Type A tests
were performed over a seven year period with successful results. The
tests indicate that the Cook Nuclear Plant has a low leakage
containment. In addition, there are no adverse trends in the results
from the previous Types A, B, and C tests or visual inspections that
indicate a gradual degradation of the containment boundary. Further,
there are no structural modifications planned which would adversely
affect the structural capability of the containment and that would
be a factor in deferring the Type A test one refueling outage.
Containment leak rate testing is not an initiator of any accident.
The proposed interval extension does not affect reactor operations
or the accident analysis and has no radiological consequences,
except for the dose savings associated with not performing the test.
There will be no changes to 10 CFR 100 dose limits or the control
room dose limits. Extending the test interval will not increase the
probability of a malfunction of equipment important to safety. Based
on these considerations, it is concluded that the change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.

Criterion 2

The proposed change does not involve physical changes to the
plant or changes in plant operating configuration. The proposed
change only relaxes the schedular requirements for conducting one
Type A test from the T/Ss and defers performance of the test one
cycle. The purpose of the test is to provide periodic verification
of the leak-tight integrity of the primary reactor containment, and
systems and components which penetrate containment. The tests assure
that leakage through containment and systems and components
penetrating containment will not exceed the allowable leak rate
values established in 10 CFR 50, Appendix J. Thus, it is concluded
that the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.

Criterion 3

The proposed change to the schedule for performing the Type A
test does not reduce the margin of safety assumed in the accident
analysis for any release of radioactive materials or reduce any
margin of safety preserved by the technical specifications. The
methodology, acceptance criteria, and the technical specification
leak rate limits for the performance of the Type A test will not
change. Type A tests will continue to be performed in accordance
with 10 CFR 50, Appendix J and the applicable Cook Nuclear Plant
Technical Specifications beginning in 1997. In addition, there are
no adverse trends in the results from the previous Type A, B, and C
tests or visual inspections that indicate a gradual degradation of
the containment boundary. Based on these considerations, it is
concluded that the change does not involve a significant reduction
in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and [[Page 20520]] Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

Date of amendment requests: March 31, 1995.
Description of amendment requests: The proposed amendments would
modify the Containment Ventilation System Technical Specifications (and
associated Bases) to allow limited containment purge operation in Modes
1, 2, 3, and 4 for pressure control, ALARA [as low as is reasonably
achievable], and respirable air quality considerations.
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:

Criterion 1

The purpose of this amendment is to allow flexibility in the use
of the containment purge system during MODES 1, 2, 3, and 4. The use
of this system during these modes of operation has previously been
approved (Amendment No. 66). Therefore, this amendment request does
not involve a significant increase in the probability or
consequences of an accident previously evaluated because the
proposed change to the T/Ss does not affect the assumptions,
parameters, or results of any UFSAR [Updated Final Safety Analysis
Report] accident analysis. Based on the existing system design and
demonstrated closure capability it is concluded that the proposed
changes do not modify the response of the containment during a
design basis accident. The proposed amendment does not add or modify
any existing equipment. Based on these considerations, it is
concluded that the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.

Criterion 2

The proposed change does not involve physical changes to the
plant or changes in the plant operating configuration. Thus, it is
concluded that the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.

Criterion 3

The margin for safety presently provided is not reduced by the
proposed change. As discussed previously, the containment purge
valves have been designed and demonstrated capable of closure
against the dynamic forces resulting from a loss of coolant
accident. The proposed amendment does not impact the ability of the
purge valves to perform their intended function (i.e. achieve
closure) in the event of an accident. Based on these considerations,
it is concluded that the changes do not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut

Date of amendment request: March 31, 1995.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to increase the as-found
setpoint tolerance of the safety/relief valves (SRVs) from plus or
minus 1% to plus or minus 3%. In addition, the proposed amendment (1)
would allow the as-found condition of one SRV to be inoperable, (2)
clarifies the 1325 psig safety limit wording, (3) increases the number
of SRVs to be tested during each refueling outage, (4) makes editorial
changes to reflect the TS changes, and (5) revises the bases for the
applicable sections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

NNECO has reviewed the proposed changes in accordance with 10
CFR 50.92 and concluded that the changes do not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10 CFR 50.92(c) are not
compromised. The proposed changes do not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The safety function of the SRVs is to mitigate the effects of a
RPV [reactor pressure vessel] overpressurization, therefore a
failure to open until the upper setpoint limit (+3%) is reached
cannot affect the probability of an accident. The lowest allowable
limit (-3%) is still above normal operating pressure and therefore
does not significantly increase the probability of an inadvertent
opening.
Should the SRVs open in response to an RCS [reactor coolant
system] overpressure event, opening of the SRVs below the nominal
setpoints does not adversely affect the consequences of an accident.
The fuel reload analysis demonstrates that actuation of five valves
at or below 103% of nominal provides sufficient pressure reduction
to maintain peak RCS pressure below the safety limit of 1375 psig
and to maintain vessel steam space pressure below 1325 psig. The
hydrodynamic loads on the SRV discharge pipe (i.e., tail pipe) and
the torus remain within the design limits.
The performance of the high pressure systems; FWCI [feedwater
coolant injection], SLC [standby liquid control] and IC [isolation
condenser] remain acceptable. There is also no adverse impact on the
operability of the APR [automatic pressure relief] system.
The SRV setpoints will continue to be required to be within
[plus or minus] 1% prior to plant startup from a refueling outage.
This ensures that the SRVs are restored to the optimal conditions at
the start of each fuel cycle.
Therefore, increasing the ``as-found'' tolerance from [plus or
minus] 1% to [plus or minus] 3% does not result in a significant
increase in the probability or consequences of a previously analyzed
accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Revising the acceptable as-found setpoint tolerance from [plus
or minus] 1% to [plus or minus] 3% does not change the type of
action that these valves are expected to perform, nor does it change
the initial ``as-left'' requirements for the valves. Plant operating
parameters have not changed. Therefore, this change cannot create
the possibility of a new or different kind of accident.
3. Involve a significant reduction in the margin of safety.
The margin of safety established and stated in the Millstone
Unit No. 1 Technical Specifications, is a peak RCS pressure of 1375
psig and a peak vessel steam space pressure of 1325 psig. While
allowing the SRV setpoint tolerance to increase to [plus or minus]
3% would allow peak pressures from an MSIV [main steam isolation
valve] closure event to approach that safety limit, the safety limit
will not be exceeded. 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: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360. [[Page 20521]]
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania

Date of application for amendments: February 10, 1995.
Description of amendment request: The proposed changes provide for
the correction of administrative errors made in the past during the
processing of technical specification changes related to control room
ventilation filter surveillance testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the changes are purely administrative and do not involve any
physical changes to plant SSC [systems, structures, or components].
Therefore, these changes will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
because the changes will not alter the plant or the manner in which
the plant is operated. The changes do not allow plant operation in
any mode that is not already evaluated in the safety analysis. The
changes will not alter assumptions made in the safety analysis and
licensing bases. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety because they are purely administrative and
have no impact on any safety analysis assumptions.

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: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101.
NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania

Date of amendment request: March 15, 1994.
Description of amendment request: This amendment would reflect an
exemption from 10 CFR Part 50, Appendix J, Section II.H.4, concerning
the scope of Type `C' testing on specified emergency core cooling
system and reactor core isolation cooling containment isolation valves
by revising Technical Specification Table 3.6.3-1, Primary Containment
Isolation Valves. The subject valves on systems which terminate below
the minimum water level of the suppression pool and are associated with
closed systems would be tested using requirements of the American
Society of Mechanical Engineers' Section XI Code.
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:

I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to the scope of Type `C' testing for the
subject valves does not affect the probability of the design basis
accidents. The valves will continue to be maintained in an operable
state, and in their current design configuration. There is no
correlation between the scope of the Type `C' testing and accident
probability.
PP&L reviewed the postulated consequences of design basis events
on primary containment isolation under the proposed change. GDC 50
design conformance states that the primary containment structure,
including access openings, penetrations and the containment heat
removal system, is designed so that the containment structure and
its internal compartments can withstand, without exceeding the
design leakage rate (1.0% per day), the peak accident pressure and
temperature that could occur during any postulated LOCA.
For the purposes of considering the consequences of LOCAs under
the proposed change, a single active failure of a CIV or a passive
failure of the closed system were reviewed, within the limits of the
existing licensing basis. Under the existing licensing basis, a pipe
rupture of seismically qualified ECCS piping does not have to be
assumed concurrent with the LOCA, except if it is a consequence of
the LOCA. Consequential failures can be eliminated, since a LOCA
inside containment is separated from the ECCS piping by the
containment structure. Consequential failures of the ECCS piping
from LOCA's outside containment are outside the Appendix J design
considerations, although they are adequately addressed through the
redundancy and separation of the ECCS design. A single active
failure of the CIV, under the LOCA condition, can be accommodated
since the closed and filled system piping remains as the leakage
barrier. The ECCS passive failure criterion does require
consideration of system leaks, but not pipe breaks, beyond the
initiating LOCA. Pipe leakage, equivalent to the leakage from a
valve or pump seal failure, should be considered at 24 hours or
greater post-LOCA. The capability to make-up inventory to the
suppression pool is adequate to ensure that postulated seat leakage
and pipe leakage does not result in a condition that jeopardizes
pool level. Make-up capability exists to the suppression pool via
the Condensate Storage Tank and Spray Pond. Actions to make-up to
the suppression pool are delineated in Emergency Operating
Procedures.
Therefore, the proposal to eliminate the subject Type `C' tests
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The acceptability of the proposed change to the scope of Type
`C' testing for the subject valves is based on maintaining the
existing barriers to primary containment leakage, and ensuring that
the suppression pool level is assured for 30 days during all design
basis, post-accident modes of operation. By meeting these dual
objectives, the plant response to the design basis events will be
unchanged, and no new accident scenarios will be encountered. These
two objectives are related, in that, the suppression pool inventory
creates a passive barrier to primary containment atmospheric leakage
for penetrations which are located below the minimum water level of
the pool. The subject valve lines terminate below the minimum
suppression pool water level.
The subject valves are all single isolation valves associated
with lines that penetrate the primary containment, but are not
connected directly to the primary containment atmosphere or the
reactor coolant pressure boundary. The redundant isolation boundary
for each of the affected valves is the closed system associated with
the valve. This configuration is described in General Design
Criteria (GDC) 57. The proposed exemption, and Technical
Specification change, does not alter the configuration of these
systems. The valves will continue to be tested and maintained to
ensure their operability. The closed system [[Page 20522]] piping
meets PP&L's design conformance to GDC 56 and is verified via a
10CFR50 Appendix J Type `A' test. The integrity of the closed
systems is also monitored and controlled via Technical Specification
6.8.4.a.
The subject valves may be open, or change state, postaccident to
support the design function of their associated ECCS systems (HPCI,
Core Spray, RHR) or RCIC. The subject valves function as system
valves during the periods when they are open or in an intermediate
state, not as containment isolation valves. Reliance is placed on
the suppression pool seal and the closed system piping to maintain
the barrier between primary and secondary containment atmospheres.
Therefore, with the valve and closed system configuration
unaffected by the proposed change, the existing barriers to primary
containment atmospheric leakage are maintained, so long as the
suppression pool level is ensured.
The suppression pool is designed and operated so that it is
filled with water in accordance with Technical Specifications 3/
4.5.3, ``Suppression Chamber,'' 3/4.6.2, ``Depressurization
Systems--Suppression Chamber,'' and the associated Bases. The supply
of water in the suppression pool is assured for 30 days during all
design basis, post-accident modes of operation. Type `C' leak rate
testing has historically been performed on valves associated with
lines that connect to the suppression pool. The acceptance criteria
for combined leakage from these penetrations is 3.3 gpm. This
leakage rate is at a level which ensures the 30 day post-accident
suppression pool level. However, for the valves discussed in this
change, seat leakage past the CIV is into a closed and filled
system. Thus ``leakage'' from the suppression pool, past the CIV, is
a function of closed system leakage.
As mentioned above, the integrity of the closed system piping is
verified via a 10CFR50 Appendix J Type `A' test and is monitored and
controlled via Technical Specification 6.8.4.a. TS 6.8.4.a
establishes a program to monitor and control leakage from systems
located outside containment that could contain highly radioactive
fluids during a serious transient or accident. This program applies
to the ECCS systems and RCIC affected by the proposed change and
ensures that leakage into secondary containment via packing,
flanges, seals, etc., is controlled. Leakage from these systems,
plus the Scram Discharge Volume, Reactor Water Clean-up, and PASS,
has been found to be very low, and well below the 5 gpm limit
established for these systems. Current leakage for Unit 1 is 0.14
gpm and for Unit 2, 0.043 gpm. The proposed change is not expected
to contribute to higher levels of system leakage. Any leakage from
these systems is processed via Standby Gas Treatment and the
radwaste system to maintain ALARA and comply with regulatory
guidance. The closed systems are maintained filled, so that a supply
of water exists on both sides of the isolation valves.
While suppression pool leakage is a function of closed system
leakage for the subject penetrations, a review of Type `C' test data
for the subject CIVs showed that the valves have had low leakage
rates during previous tests. This leakage is on the order of 0.6
gpm, per unit. Proposed testing of the valves under Section XI and
the current requirements of the Generic Letter 8910 program will
ensure valve operability.
Therefore, leakage past the CIV and out of the closed system is
expected to be low and in keeping with the design basis for the
suppression pool. However, the capability does exist to make-up
water to the suppression pool from the Condensate Storage Tank or
Spray Pond if necessary. Existing Emergency Operating Procedures
require actions if suppression pool level is less than 22 feet or
greater than 24 feet. Thus, the level of the suppression pool is
ensured, independent of the current CIV Type `C' testing
requirement.
The proposed change to the scope of Type `C' testing for the
subject valves maintains the existing barriers to primary
containment leakage, and ensures that the suppression pool level is
assured for 30 days during all design basis, post-accident modes of
operation. Therefore, the plant response to the design basis events
is unchanged, and the proposal does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
As discussed in questions I and II, the proposed change does not
alter the plant response to existing accident scenarios, and does
not introduce new or different scenarios. So the margin of safety
from a design basis accident standpoint is maintained.
Historically, the leakage rate through the subject valves has
been determined under the Type `C' testing program. This leakage
rate has been found to be very low, and is currently on the order of
0.6 gpm. Quantifying leakage past the CIVs has been used to ensure
that the suppression pool level is assured for 30 days post-
accident. Under the proposed change, this leakage rate will not be
quantified. This is acceptable since leakage from the suppression
pool is in reality a function of closed system leakage, not solely
CIV leakage. Closed system leakage is monitored and controlled by an
existing Technical Specification program. Closed system leakage has
been found to be very low on both units, and is currently a small
fraction of a gallon per minute compared with a 5 gpm allowable.
Therefore, leakage past the CIV and out of the closed system is
expected to be low and in keeping with the design basis for the
suppression pool. However, the capability does exist, and is
proceduralized, to make-up water to the suppression pool from the
Condensate Storage Tank or Spray Pond if necessary. Thus the current
capability to maintain adequate suppression pool level for 30 days
postaccident is assured under the proposed change.
Therefore the proposed change to the scope of Type `C' testing
for the subject valves 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania

Date of amendment request: March 31, 1995.
Description of amendment request: These amendments would modify the
surveillance requirement for reactor coolant system pH analysis in
section 4.4.4 of the Technical Specifications (TS) for each unit. Also,
they would clarify in the TS that the pH analysis would be taken at
least every 72 hours whenever reactor coolant conductivity exceeds 1.0
mho/cm.
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:

I. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The pH limits on reactor coolant are not affected by this
change. The pH will be measured whenever it is theoretically
possible for it to be outside the Tech Spec [Technical
Specifications] limits of <5.6 or="">8.6 (i.e., whenever the
conductivity is greater than 1.0 mho/cm). Because of the
theoretical relationship between pH and conductivity as shown in
Attachment A [see application dated March 31, 1995, for this
reference], it is possible to establish pH limits on the reactor
coolant by limiting the conductivity. As shown in this figure, the
pH must be >5.6 and <8.6 if="" the="" conductivity="" is="" less="" than="" or="" equal="" to="" 1.0="">mho/cm. Attachment A was taken from Regulatory Guide
1.56 Revision 1, July 1978 ``Maintenance of Water Purity in Boiling
Water Reactors''. As noted in both FSAR final safety analysis report
and Technical Specification Bases, the pH and conductivity limits
for OPERATIONAL CONDITION 1 are consistent with this theoretical
relationship. The Bases for Section 3/4.4.4 of the Tech Specs
[Technical Specifications] contains [contain] the following
statement: ``When the conductivity is within limits, the pH,
[[Page 20523]] chlorides and other impurities affecting conductivity
must also be within their acceptable limits[''].
Since the conductivity is measured by grab sampling at least
every 72 hours to verify that it is within limits, this will also
verify that pH is within limits every 72 hours. If the conductivity
should exceed 1.0 mho/cm, pH measurements will be made to
determine if the Tech Spec [Technical Specifications] pH limits have
been exceeded. It should also be noted that inline conductivity
instrumentation is very stable and reliable and is used to
continuously monitor the reactor coolant per Tech Spec [Technical
Specifications] requirements, with instrumentation connected to
redundant sources (reactor water cleanup influent and reactor
recirculation loop). Therefore, the proposed change will not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
II. Create the possibility of a new or different kind of
accident from any accident previously evaluated.
As stated above, the pH limits on reactor coolant are not
affected by this change. Since the conductivity is monitored
continuously, to verify that it is within limits, this will also
verify that pH is within limits. If the conductivity should exceed
1.0 mho/cm, pH measurements will be made to determine if
the Tech Spec [Technical Specifications] pH limits have been
exceeded. Therefore, the incorporation of this change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
III. Involve a significant reduction in a margin of safety.
The in-line conductivity instrumentation has been determined to
be very stable and reliable in its use to continuously monitor the
reactor coolant per Tech Spec [Technical Specifications]
requirements. To maintain this reliability, this instrumentation is
connected to redundant sources (reactor water cleanup influent and
reactor recirculation loop). Based on this continuous monitoring of
reactor coolant conductivity, as provided by this instrumentation,
the incorporation of this change will have no impact on current
safety margins, nor will it 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania

Date of amendment request: March 31, 1995.
Description of amendment request: These amendments would delete
from the Technical Specifications of each unit, the operational
condition restriction in Surveillance Requirement 4.8.1.1.2.d.7 which
requires that 24-hour emergency diesel generator testing be performed
with at least one unit in operational condition 4 or 5 (cold shutdown
or refueling).
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 proposed changes do not:

I. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to permit the 24 hour testing of the
emergency diesel generators during power operation does not increase
the chances for a previously analyzed accident to occur. The
function of the EDGs [emergency diesel generators] is to supply
emergency power in the event of a loss of offsite power. As stated
above [,] the diesel generator being tested has been determined to
remain operable and available to supply the emergency loads within
the required times. In addition, the three remaining EDGs will be
operable during this test. Operations [Operation] of an EDGs [EDG]
is not a precursor to any accident. If, however, an offsite
disturbance were to occur that affected the operability of the DG
[emergency diesel generator] being tested, the remaining EDGs are
capable of feeding the loads necessary for safe shutdown of the
plant. In summary, the proposed change does not adversely affect the
performance or the ability of the diesel generators to perform their
intended safety function. Therefore, the proposed change will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
II. Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change to the 24 hour surveillance requirement will
not affect the operation of any safety system or alter its response
to any previously evaluated accident. The diesel generator will
automatically transfer from test mode if necessary to supply
emergency loads in the required time. The test mode is used for the
monthly surveillances of these diesel generators, resulting in no
new plant operating modes being introduced. In the event the EDG
fails the functional test[,] it will be declared inoperable and the
actions required for an inoperable diesel will be performed. The
remaining three EDGs will be operable and are capable of feeding the
loads necessary for safe shutdown of the plant. Therefore, the
incorporation of this change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
III. Involve a significant reduction in a margin of safety.
Changing the EDG test timing results is no reduction in the
safety margin as defined in the design basis. Because loss of an EDG
is not expected as a result of LOOP [loss of offsite power] or LOCA/
LOOP [loss-of-coolant accident with a loss of offsite power] during
the 24 hour test, SSES [Susquehanna Steam Electric Station] remains
within its design basis. In fact, because the test EDG loads the ESS
[engineered safety system] bus 8.5 seconds earlier than the non-test
EDGs during LOCA with LOOP, plant response is actually improved.
Risk of operation during the 24 hour EDG test is certainly less than
during the current 84 hour allowed outage time (AOT) because both
the impact of the initiating events evaluated (EDG in test is not
actually failed) and the frequency of the limiting plant condition
(loss of two EDGs) are less. No increase in frequency or impact of
design basis events, and no reduction in the safety margin occurs
during the 24 hour EDG test. Therefore, the incorporation of this
change will have no impact on current safety margins, nor will it
involve a significant reduction in the margin to [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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket No. 50-388, Susquehanna
Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of amendment request: March 31, 1995.
Description of amendment request: This amendment would change
Susquehanna Unit 2 Technical Specifications (TS) by incorporating the
General Electric (GE) NRC approved methodology for GE-12 type lead use
fuel assemblies (NEDE-24011-P-A-10) into the list of references in
Section 6.9.3.2. The licensee plans to insert four of these fuel
assemblies into the Unit 2 core during the fall of 1995. The addition
of the reference to the TS would allow the use of the GE methodology to
document that all [[Page 20524]] applicable requirements of the safety
analysis are met by the assemblies.
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 proposed change does not:

I. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Incorporation of this proposed change of adding reference NEDE-
24011-P-A-10, ``General Electric Standard Application for Reactor
Fuel'' to the list of references in [the] Unit 2 Technical
Specifications will allow the use of the GE methodology to calculate
the operating limits for the four GE Lead Use Assemblies which are
of a different mechanical design than the Siemens 9X9 fuel
[currently installed in the reactor core]. This NRC approved
methodology will be referenced as the approved methodology in
showing that all applicable safety limits of the safety analysis are
met by the four GE-12 LUAs. Results of incorporating this change
will not significantly increase the probability or the consequences
of an accident previously evaluated.
II. Create the possibility of a new or different kind of
accident from any accident previously evaluated.
As stated above, the incorporation of this change will allow the
use of the GE methodology to be referenced as the approved
methodology to show that all applicable limits of the safety
analysis are met by the four GE-12 LUAs. Therefore, the
incorporation of this change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
III. Involve a significant reduction in a margin of safety.
The use of the GE methodology will not result in a change in
safety margin, but will ensure that the safety margin is maintained
with the insertion of the four GE LUAs of the GE-12 type in Unit 2
Cycle 8. Therefore, the incorporation of these changes will have no
impact on current safety margins, nor will they involve a
significant reduction in the margin to [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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

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

Date of amendment request: December 2, 1994.
Description of amendment request: The proposed change to Limerick
Generating Station, Units 1 and 2, Technical Specifications (TS)
relocates the TS Fire Protection Requirements to Licensee controlled
documents consistent with NRC Generic Letter (GL) 86-10
``Implementation of Fire Protection Requirements,'' and GL 88-12,
``Removal of Fire Protection Requirements from Technical
Specifications.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed Technical Specifications changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes are administrative in nature and are
consistent with NRC GL 86-10 and GL 88-12. Removal of Fire
Protection Program (FPP) requirements does not affect any fire
protection equipment nor plant equipment important to safety, or
involve any physical modifications to plant structures, systems or
components, and therefore is not associated with an accident
initiator or accident mitigator and can not affect the probability
of occurrence of an accident or increase the consequences of an
accident. The licensee controlled Technical Requirements Manual
(TRM) containing the relocated requirements will be maintained in
accordance with TS Section 6.0. ``Administrative Controls'' and
subject to review in accordance with 10 CFR 50.59. Since future
changes to the FPP (i.e., Updated Final Safety Analysis Report and
the TRM) will be evaluated per 10 CFR 50.59, no increase
(significant or insignificant) in the probability or consequences of
an accident previously evaluated will be allowed. Therefore, these
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The changes will not alter the plant configuration (no new or
different type of equipment will be installed) or create changes in
methods governing normal plant operation that will introduce new
failure modes. These changes will not impose different requirements
and proper control of information will be maintained. These changes
will not alter assumptions made in the safety analysis and licensing
basis. Therefore, these changes do not create the possibility of a
new or different kind of accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed changes are administrative in nature and are
consistent with NRC GL 86-10 and GL 88-12. The changes will not
reduce the margin of safety since they have no impact on any safety
analysis assumptions or sequence of events used in any accident
analysis. In addition, any future changes to the FPP will be
evaluated per the requirements of 10 CFR 50.59. Therefore, the
changes do not involve a significant reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: John F. Stolz.

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

Date of amendment request: January 27, 1995.
Description of amendment request: The proposed change to Limerick
Generating Station (LGS) Units 1 and 2 Technical Specifications (TS)
will eliminate the TS active safety function designation of eight
(i.e., four per unit) Drywell Chilled Water System (DCWS) valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed Technical Specifications changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes will eliminate the TS active safety
function designation of eight (i.e., four per unit) DCWS valves. The
DCWS motor operated valves (MOVs) are designated in TS as Primary
Containment Isolation Valves (PCIVs), during operational conditions
(OPCONS) 1, 2, and 3, which mitigate the consequences of design
basis accidents. The proposed changes will prohibit the subject DCWS
valves from opening during OPCONs 1, 2, and 3, thereby,
[[Page 20525]] eliminating the active safety function, and
maintaining a passive safety function. The postulated accidents
which require the Primary Containment to act as a barrier in order
to mitigate the release of radioactivity described in the LGS
Updated Final Safety Analysis Review [Report] (UFSAR) Section 15,
are not affected by these changes. Therefore, the previously
evaluated postulated on-site and off-site radiological effects of
these accidents will not change.
The DCWS valves will be prohibited from opening during OPCONs 1,
2, and 3 by physical changes made to the electrical control
circuitry and administrative controls. Therefore, the probability of
the valves to fail in the open position will diminish, and the
required Primary Containment isolation safety function will be
maintained.
Therefore, these proposed changes will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes remove the affected automatic isolation
relays from the DCWS MOVs' circuitry. These changes eliminate any
postulated relay failure effects on the associated control circuits
and electrical power supplies. The proposed changes do not introduce
any new accident initiators or any new valve failure modes not
previously evaluated.
Therefore, these changes will not create the possibility of a
new or different kind of accident from any accidents previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed changes will prohibit the opening of the DCWS
valves which provide backup cooling from RECW [reactor enclosure
cooling water] during OPCONs 1, 2, and 3. The RECW System is not the
normal DCWS cooling alignment, is not required as a backup safety
related drywell cooling system, and this backup alignment is not an
automatic function. The proposed changes do not affect the function
or operation of DCWS, and since the proposed changes and
administrative controls ensure the valves will remain closed during
OPCONs 1, 2, and 3, the capability for Primary Containment isolation
is not affected. Therefore, the changes 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: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: John F. Stolz.

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

Date of amendment request: February 22, 1995.
Description of amendment request: The proposed change to Limerick
Generating Station (LGS) Units 1 and 2 Technical Specifications (TS)
revises various TS Surveillance Requirements to clarify the Emergency
Diesel Generator acceptable steady state voltage range.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed Technical Specifications changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed Emergency Diesel Generator steady state voltage
range of 4280 [plus or minus] 120 volts provides voltages through
the 4160V and 480V distribution systems which are within the
operating range of the connected equipment and power system
components. Therefore, the reduced steady state voltage range will
not cause the malfunction of any equipment or affect the operation
of any equipment in a manner which would increase the probability of
occurrence of an accident previously evaluated in the [Safety
Analysis Report] SAR.
Reducing the Emergency Diesel Generator steady state voltage
range in the Technical Specifications maintains the capability of
the Emergency Diesel Generator to start and attain rated voltage and
frequency within 10 seconds and to accept the engineered safeguard
loads in the required time in order to mitigate the consequences of
an accident. The Emergency Diesel Generator automatic voltage
regulator setting is calibrated to within a range of 4266.5 volts to
4308.5 volts. A review of results from recent monthly Emergency
Diesel Generator Surveillance Tests has confirmed that the voltage
regulators currently maintain the Emergency Diesel Generator steady
state voltage within the 4280 [plus or minus] 120 volt range to be
included in the Technical Specifications. Establishing, via
Technical Specification surveillance requirements and administrative
limits within Station Surveillance Test Procedures, that the
Emergency Diesel Generator voltage regulator is maintaining the
steady state voltage within a narrower range (within the existing
range) provides increased assurance that connected equipment
required to mitigate the consequences of an accident will operate as
required.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Reducing the Emergency Diesel Generator steady state voltage
range in the Technical Specifications to a range of 4280 [plus or
minus] 120 volts does not create any new accident initiators or
affect any existing accident initiators such that a different type
of accident than previously evaluated could result. The function and
operation of the Emergency Diesel Generators and their connected
loads are not changed in a manner which would create the possibility
of an accident of a different type than any previously evaluated.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
Reducing the Emergency Diesel Generator steady state voltage
range in the Technical Specifications to a range of 4280 [plus or
minus] 120 volts does not reduce the margin of safety. The reduced
range provides increased assurance that the equipment powered by the
Emergency Diesel Generators will start and operate as designed in
order to perform their design basis functions.
Therefore, the proposed TS changes do not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: John F. Stolz.

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

Date of amendment request: March 1, 1995.
Description of amendment request: The proposed changes will clarify
the concentrations of calibration gas required to calibrate the
Hydrogen and Oxygen Analyzers, and support the requirements of Limerick
Generating Station (LGS) Transient Response Implementation Plant (TRIP)
T-102, ``Primary Containment Control Bases.''
Basis for proposed no significant hazards consideration
determination: [[Page 20526]] As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:

1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS [Technical Specification] changes remove
calibration of the H2/O2 Analyzers using zero volume percent
hydrogen (H2) and 100% bottled Nitrogen (N2). A calibration gas
containing zero volume percent H2 and 100% bottled N2 is not
required for calibration of the analyzers to the required accuracy.
Calibration of the H2/O2 Analyzers is done in accordance with the
manufacturer's instructions. The proposed TS changes also revise the
span gas concentration from 5% to 7% to support the requirements of
TRIP T-120. The H2/O2 Analyzers provide indication of the
concentrations of combustible gases in the primary containment and
provide annunciation when combustible gas concentrations reach
unacceptable levels. Failure of the analyzers is not an accident
initiator. The analyzers do not connect to the reactor coolant
pressure boundary; therefore, they do not increase the probability
of a LOCA [loss-of-coolant accident]. The proposed TS changes do not
involve any design changes to analyzers. Therefore, these TS changes
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The H2/O2 Analyzers provide indication and alarms for H2 and O2
concentrations in containment. No physical or design changes to the
analyzers are being made by these TS changes. During normal
operations, the potential for an explosive atmosphere is negligible
due to the absence of H2 sources. For Post-LOCA, conditions the
levels of H2 and O2 in containment have already been evaluated in
LGS UFSAR [Updated Final Safety Analysis Report] Section 6.2.5. No
physical or design changes which could introduce a new analyzer
failure mode are being made. The failure modes of the analyzers are
evaluated in UFSAR Table 6.2-21. Therefore, these TS changes will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
These TS changes will clarify statements in the LGS UFSAR and TS
concerning calibrated ranges of the analyzers. The change of the
span gas from 5% to 7% falls within conditions previously analyzed.
The Bases for TS 3/4.3.7.5 and 3/4.6.6 require operable H2/O2
Analyzers to ensure the analyzers will be available for monitoring,
assessing and controlling H2 and O2 in containment following a LOCA.
These TS changes do not adversely affect operability of the
analyzers or their availability for use during Post-LOCA conditions;
therefore, the margin of safety is not 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: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz.

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

Date of amendment request: December 15, 1994.
Description of amendment request: In accordance with 10CFR50.90,
PSE&G proposes to remove Technical Specification Surveillance
Requirement 4.8.1.1.2.h.1, and utilize plant- controlled programs to
govern diesel generator maintenance. To ensure procedural consistency
and reduce the impact of this change on Hope Creek procedures, the
remaining Surveillance Requirements of Technical Specification
4.8.1.1.2.h are not renumbered.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change is consistent with the improved Standard
Technical Specifications (NUREG-1433) and does not result in any
changes to the existing plant design. The Hope Creek preventative
maintenance program will utilize diesel generator performance
history, engineering analyses and manufacturer's recommendations as
appropriate for determining diesel generator inspection
requirements. Since the changes do not impact the ability of the
diesel generators and the AC electrical power sources to perform
their function, the changes do not result in a significant increase
in the consequences of any accident previously evaluated. The diesel
generators will continue to function as designed. Therefore, the
proposed change will not impact the probability of occurrence of any
accident previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
This request does not result in any change to the plant design
nor does it involve a significant change in current plant operation.
The diesel generators will be inspected utilizing diesel generator
operating history, engineering analyses and manufacturer's
recommendations as appropriate, and the remaining surveillance
requirements will not be changed. As a result, no new failure modes
will be introduced, and the proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Will not involve a significant reduction in a margin of
safety. The proposed request does not adversely impact the
reliability of the diesel generators. As stated above, the diesel
generator operating history, engineering analyses and the
manufacturer's recommendations will be utilized as appropriate to
perform the diesel generator inspections. In addition, the diesel
generators will continue to perform their design functions. This
request does not involve an adverse impact on diesel generator
operation or reliability. Since the diesel generator function is not
affected by the proposed changes, this request 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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee

Date of amendment request: April 6, 1995 (TS 95-09).
Description of amendment request: The proposed change would revise
Operating Condition 2.C.(25) to extend the ice condenser Surveillance
4.6.5.1.d to October 1, 1995, to coincide with the Unit 1 Cycle 7
refueling outage.
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 NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.

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

Date of amendment request: April 6, 1995 (TS 95-11).
Description of amendment request: The proposed change would
relocate the constant numerical value found in the overtemperature
delta temperature and overpower delta temperature equations of
Technical Specifications Table 2.2-1 and place them in the Core
Operating Limits Report (COLR). This would be accomplished by revising
notes 1 and 2 in Table 2.2-1 to state that the values are located in
the COLR. The values of the constants, however, would not be changed.
Also, the ``Overtemperature and Overpower Delta Temperature Setpoint
Parameter Values for Specification 2.2.1'' would be added to the list
of core operating limits specified in Section 6.9.1.14 that are
required to be included in the COLR. In addition, a reference to WCAP-
8745-P-A, ``Design Bases for the Thermal Overpower delta-T and Thermal
Overtemperature delta-T Trip Functions,'' would be added to Section
6.9.1.14.a.
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:

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 6, 1995 (TS 94-15).
Description of amendment request: The proposed change would modify
the Technical Specifications associated with the Post Accident Sampling
(PAS) system by deleting License Condition 2.C.(23)F for Unit 1 and
2.C.(16)g for Unit 2 that require operation of the PAS system in
accordance with referenced letters no later than startup from the
second refueling outage. The submittal also includes a revised
description of operation of the PAS system for insertion into the
Updated Final Safety [[Page 20528]] Analysis Report for staff approval.
This information supersedes the information contained in the letters
referenced in the License Conditions listed above and would be
maintained in accordance with the 10 CFR 50.59 process.
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:

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 6, 1995 (TS 94-18).
Description of amendment request: The proposed change would revise
Surveillance Requirement (SR) 4.0.5 by replacing the current Inservice
Inspection program requirements with the requirements stated in the
Standard Technical Specifications (NUREG-1431). As a result, SR 4.0.5
would more clearly specify the inservice inspection (ISI) program
requirements and the inservice testing (IST) program requirements of
the American Society of Mechanical Engineers (ASME) Code Class 1, 2,
and 3 components. The licensee also proposed deletion of Technical
Specification 3/4.4.10, ``Structural Integrity ASME Code Class 1, 2 and
3 Components,'' and its related Bases information.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Operation of the facility in accordance with the proposed
amendment would not involve any increase in the probability of
occurrence or consequences of an accident previously evaluated. The
Inservice Inspection and Testing Programs, pursuant to 10 CFR 50.55a
are described in the TSs. The proposed amendment, in accordance with
NUREG-1431 and draft NUREG-1482 permits relief from an American
Society of Mechanical Engineers (ASME) code requirement in the
interim between the time of submittal of a relief request and NRC
approval of the relief. The changes being proposed do not affect
assumptions contained in plant safety analyses or change the
physical design and/or operation of the plant, nor do they affect
TSs that preserve safety analysis assumptions. Any relief from the
approved ASME Section XI code requirements that is implemented prior
to NRC review and approval will require evaluation under the 10 CFR
50.59 process to determine that no TS changes or unreviewed safety
questions exist. This evaluation process will ensure that the impact
of any code relief is thoroughly evaluated and that the structures,
systems, and components remain in conformance with assumptions made
in the safety analysis. The proposed change to delete SQN TS 3/
4.4.10, Structural Integrity, does not affect plant safety analyses
or change the physical design or operation of the plant. The
proposed amendment relocates the structural integrity requirements
under the existing TS Surveillance Requirement (SR) 4.0.5 to allow
these requirements to be governed and controlled within the
inservice inspection (ISI) program. Therefore, operation of the
facility in accordance with the proposed amendment would not affect
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The
Inservice Inspection and Testing Programs, pursuant to 10 CFR 50.55a
are described in the TSs. The proposed amendment, in accordance with
NUREG-1433 and draft NUREG-1482, permits relief from an ASME code
requirement in the interim between the time of submittal of a relief
request and NRC approval of the relief. The changes being proposed
will not change the physical plant or the modes of operation defined
in the Facility License. The changes do not involve the addition or
modification of equipment nor do they alter the design or operation
of plant systems. Any relief from the approved ASME Section XI code
requirements that is implemented prior to NRC review and approval
will require evaluation under the 10 CFR 50.59 process to determine
that no TS changes or unreviewed safety questions exist. This
evaluation process will ensure that the impact of the code relief is
thoroughly evaluated and that the structures, systems, and
components remain in conformance with assumptions made in the safety
analysis. The proposed change to delete SQN TS 3/4.4.10 does not
affect plant safety analyses or change the physical design or
operation of the plant. The proposed amendment relocates the
structural integrity requirements under the existing TS SR 4.0.5 to
allow these requirements to be governed and controlled within the
ISI program. Therefore, operation of the facility in accordance with
the proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Operation of the facility in accordance with the proposed
amendment would not involve any reduction in a margin of safety.
[[Page 20529]] The Inservice Inspection and Testing Programs,
pursuant to 10 CFR 50.55a, are required by the SQN TSs. The proposed
amendment, in accordance with NUREG-1431 and draft NUREG-1482
permits relief from an ASME code requirement in the interim between
the time of submittal of a relief request and NRC approval of the
relief. Any relief from the ASME Section XI code is required to be
evaluated under the 10 CFR 50.59 process to determine that no TS
changes or unreviewed safety questions exist. This evaluation
process will ensure that code relief does not affect the ability of
structures, systems, or components to perform their design function,
affect compliance with any TS requirements or reduce the margin of
safety. The proposed change to delete SQN TS 3/4.4.10 does not
affect plant safety analyses or change the physical design or
operation of the plant. The proposed amendment relocates the
structural integrity requirements under the existing TS SR 4.0.5 to
allow these requirements to be governed and controlled within SQN's
ISI program. Therefore, operation of the facility in accordance with
the proposed amendment would not involve a reduction in the margin
of safety.

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 6, 1995 (TS 95-06).
Description of amendment request: The proposed change would delete
the technical specification requirement that limits and controls loads
traveling over the spent fuel pool (Specification 3.9.7), the graph
that relates the Load Carried Over the Shield to the Allowable Height
Above the Shield Surface (Figure 3.9-1), the crane interlocks and
physical stops surveillance requirements (Specifications 4.9.7.1 and
4.9.7.2), and the related Bases information. These requirements would
be relocated to administratively controlled procedures.
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:

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 6, 1995 (TS 94-19).
Description of amendment request: The proposed change would revise
the Action statement for Technical Specification 3.8.1.1 by inserting a
new Action a, relabeling and modifying existing Action a to become
Action b, adding a footnote referenced to Action b, renumbering the
subsequent action statements, and adding information to the Bases that
amplifies the action statements. The proposed new Action a would no
longer address required actions for diesel generator testing. It would
require that, should one of the AC electrical power sources listed be
inoperable, then operability of the remaining offsite AC circuit be
demonstrated by performing Surveillance Requirement 4.8.1.1.1.a within
1 hour and at least once per 8 hours thereafter. If two offsite
circuits cannot be restored within 72 hours, the unit must be placed in
hot standby within the next 6 hours and in cold shutdown within the
following 30 hours.
The proposed change to Action b would address the testing
requirements should a diesel generator become inoperable. It would
require testing of operable diesel generators if the inoperability of
the affected diesel generator has the potential to be the result of a
common cause failure. A footnote would clarify that the common cause
determination must be completed regardless of how long the diesel
generator inoperability persists or Surveillance 4.8.1.1.2.a.4 must be
completed to verify diesel generator operability. The proposed change
to the Bases would provide clear guidance on the use of common cause
failure determinations to eliminate unnecessary diesel generator
testing and would define when testing is required to verify diesel
generator operability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)
change and has determined [[Page 20530]] that it does not represent
a significant hazards consideration based on criteria established in
10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in
accordance with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revisions do not alter the plant features or
operating practices. This revision will reduce unnecessary starts of
the diesel generators (D/Gs) when a common cause failure is not
involved or for an inoperable offsite circuit. This change will not
affect the accident mitigation capabilities of the D/Gs, but should
improve the reliability by reducing the wear and tear associated
with starting the D/Gs. The D/Gs are not the source of a postulated
accident and because this change does not alter plant functions or
operating practices the probability of an accident is not increased.
The D/G's operability will continue to be verified for conditions
that indicate a potential common-cause failure to ensure accident
mitigation capabilities are not affected. Therefore, this revision
will continue to provide actions that will support alternating-
current (ac) electrical power source safety functions without
unnecessary degradation of the D/Gs and will not increase the
consequences of an accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The D/Gs are not the source of accidents and the proposed
revision will not alter plant functions or actions by more
appropriately limiting the conditions when a D/G must be verified
operable. Therefore, the possibility of a new or different accident
is not created
3. Involve a significant reduction in a margin of safety.
This revision does not alter plant functions that provide the
margin of safety. The reduction of D/G testing will only be allowed
for situations where the operable D/Gs are not affected by the
conditions resulting in the ac power source inoperability. This
reduced testing should improve D/G reliability for accident
mitigation functions and further ensure the margin of safety
provided by the D/Gs. Therefore, the margin of safety is not reduced
by the proposed revision to limit unnecessary D/G starts.

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 6, 1995 (TS 95-02).
Description of amendment request: The proposed change would add
Limiting Condition for Operation 3.0.6 to allow equipment that has been
removed from service or declared inoperable to be returned to service
under administrative control in order to perform testing required to
demonstrate operability. It would be applicable for operability testing
of the inoperable equipment or other equipment that requires the
operability feature to be in service in order to perform the test. A
related change to the Bases would provide amplifying explanation on the
use of this new provision. In addition, a proposed change to Action 18
of Table 3.3-3, ``Engineered Safety Feature Actuation System
Instrumentation,'' would clarify the time interval that an instrument
channel may be in the bypass condition. For those instruments that
reference Action 18, the change would allow the bypass for 6 hours.
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:

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 6, 1995 (TS 94-04).
Description of amendment request: The proposed amendment would
change the power range neutron flux channel calibration frequency
[[Page 20531]] surveillance requirement from monthly to every 31
effective full power days and delay the requirement to perform the
surveillance for 96 hours after reaching 15 percent power. A proposed
change to the Bases would provide amplifying information.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The likelihood that an accident will occur is neither increased
or decreased by this TS change, which only affects when the first
surveillance is performed following an outage and changes the
frequency of performance of the surveillance. Before start-up
following refueling outage, the power range high trip setpoint is
set below 85 percent power, typically 60 percent, for conservatism.
The power range low trip setpoint is set at 22 percent power, TS
requires the setpoint to be less than or equal to 25 percent power.
These settings are in addition to the conservatism built into start-
up following a refueling outage. Therefore, delaying the first
performance for 96 hours will not impact on the operation of the
plant since the setpoints are set conservatively. Also, the change
of the frequency to every 31 effective full power days (EFPD) only
delays the surveillance when the plant is operated at reduced power.
During operation at reduced power changes in the neutron flux are
also reduced. Therefore, changing the frequency from monthly to
every 31 EFPD allows slow changes in neutron flux during the fuel
cycle to be more accurately detected and evaluated.
This TS change will not impact the function or method of
operation of plant equipment. Thus, there is not a significant
increase in the probability of a previously analyzed accident due to
this change. No systems, equipment, or components are affected by
the proposed change. Thus, the consequences of a malfunction of
equipment important to safety previously evaluated in the Updated
Final Safety Analysis Report are not increased by this change.
The proposed changes provide TS improvements that ensure the
system operates within the bounds of SQN's accident analysis as
contained in the Final Safety Analysis Report (FSAR) and only
affects when a surveillance is performed. This change has no impact
on accident initiators and does not involve a physical modification
to the plant. Accordingly, the proposed changes do not involve an
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.
This revision will not change any plant equipment, system
configurations, or accident assumptions. This change will more
accurately monitor changes in the condition of the core.
Fuel burn-up is necessary to change the relationship between the
incore axial power and the excore detectors response. At reduced
levels the effectiveness of the monitoring activity is reduced.
Therefore, changing the frequency to 31 EFPD allows slow changes in
neutron flux during the fuel cycle to be more accurately detected
and evaluated. Delaying the first performance of the surveillance
requirement, until 96 hours after reaching 15 percent rated thermal
power, will allow the unit to be in a more stable condition.
Therefore, this change will not affect the safety function of any
components and will create the possibility of a new or different
kind of accident.
3. Involve a significant reduction in a margin of safety.
The proposed changes provide TS improvements for SQN's power
range monitoring system that ensure the system operates within the
bounds of SQN's accident analysis as contained in the FSAR since
only the time interval between performances of the surveillance is
being extended. This change does not involve a physical modification
to SQN's power range monitoring system. Accordingly, the margin of
safety has not been reduced.

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 6, 1995 (TS 95-08).
Description of amendment request: The proposed change would (1)
change the core alteration definition to limit the term to reactor
vessel internal activities that could have an affect on core
reactivity, (2) change the quadrant power tilt ratio definition to
eliminate the conflict in the definition of the term and its use in
Surveillance Requirement 4.2.4.2, and (3) revise the Unit 1 Operational
Modes parameters in Table 1.1 to be consistent with the description in
Table 1.1 for Unit 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

[[Page 20532]] The NRC has reviewed the licensee's analysis and,
based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.

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

Date of amendment request: March 24, 1995.
Description of amendment request: This amendment request proposes
to revise Technical Specification (TS) 1.7, ``Containment Integrity,''
TS 3/4.6.1, ``Containment Integrity,'' TS 3/4.6.3, ``Containment
Isolation Valves,'' and the associated Bases. These proposed changes
will relocate TS Table 3.6-1, ``Containment Isolation Valves,'' to Wolf
Creek Generating Station procedures. This proposed change is in
accordance with the guidance provided in Generic Letter 91-08,
``Removal of Component Lists from Technical Specifications,'' dated May
6, 1991.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes simplify the technical specifications, meet
the regulatory requirements for control of containment isolation,
and are consistent with the guidelines of GL 91-08. The procedural
details of Technical Specification Table 3.6-1 have not been
changed, but only relocated to a different controlling document. The
proposed changes are administrative in nature, should result in
improved administrative practices, and do not affect plant
operations.
The probability of occurrence of a previously evaluated accident
is not increased because this change does not introduce any new
potential accident initiating conditions. The consequences of an
accident previously evaluated is not increased because the ability
of [the] containment to restrict the release of any fission product
radioactivity to the environment will not be degraded by this
change.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes are administrative in nature, do not result
in physical alterations or changes to the operation of the plant,
and cause no change in the method by which any safety-related system
performs its function. Therefore, this proposed change 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 the margin of safety.
The administrative change to relocate Technical Specification
Table 3.6-1 to appropriate plant procedures does not alter the basic
regulatory requirements for containment isolation and will not
adversely affect containment isolation capability for credible
accident scenarios. Adequate control of the content of the table is
assured by existing plant procedures.
The proposed relocation of Technical Specification Table 3.6-1
does not alter the requirements for containment isolation valve
operability currently in the technical specifications. The LCO
[limiting condition for operation] and Surveillance Requirements
would be retained in the revised technical specifications.
Therefore, the proposed change will not affect the meaning,
application, and function of the current technical specification
requirements for the valves in Table 3.6-1.

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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.

Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.

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

Date of amendment request: April 4, 1995, as supplemented by letter
dated April 5, 1995.
Description of amendment request: The proposed amendment would
change the technical specifications on moderator temperature
coefficient. The proposed change constitutes a one time deviation not
to perform the two-thirds end-of-cycle moderator temperature
coefficient test for Cycle 7.
Date of individual notice in the Federal Register: April 11, 1995
(60 FR 18432).
Expiration date of individual notice: May 11, 1995.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Notice of Issuance of Amendments to Facility Operating Licenses

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs
Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of application for amendment: February 24, 1995.
Brief description of amendment: The amendment revises Technical
Specification Section 4.6.1.2.a, Primary Containment/Containment
Leakage, to reference 10 CFR Part 50, Appendix J, as modified by
approved exemptions.
Date of issuance: April 10, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 183.
Facility Operating License No. DPR-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 8, 1995 (60 FR
12789).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 10, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina

Date of application for amendments: September 30, 1994, as
supplemented on March 24, 1995.
Brief Description of amendments: The proposed change will revise
Technical Specification requirements to eliminate the reactor scram and
isolation functions of the Main Steam Line Radiation Monitors. The
March 24, 1995, supplement provided clarifying information only and did
not affect the NRC's determination of no significant hazards
considerations.
Date of issuance: March 31, 1995.
Effective date: March 31, 1995.
Amendment No.: 160.
Facility Operating License Nos. DPR-71 and DPR-62. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55867). The March 24, 1995, submittal provided clarifying information
only and did not affect the no significant hazards consideration as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: July 22, 1994, as supplemented
March 6, 1995.
Brief description of amendment: The amendments change the Technical
Specifications to implement a performance based assessment program,
including corresponding organizational and functional changes.
Specifically, the changes affect the independent assessment of plant
activity and the independent review function, the independent
assessment of plant activity and the Independent Safety Engineering
Group. These functions will be performed by the Nuclear Assessment
Section (NAS). The NAS's fundamental role will be to: (1) Assist plant
management in the early identification of issues that may prevent the
plant from achieving quality, and (2) ensure effective correction of
deficiencies.
Date of issuance: March 31, 1995.
Effective date: March 31, 1995.
Amendment No.: 160.
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45017). The March 6, 1995, submittal added Radiation Protection to the
list of assessments in TS 6.5.5.2 and reworded Section 6.5.4.4, but did
not change the no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: August 23, 1994, as supplemented
March 2, 1995.
Brief description of amendment: The amendment increases the trip
voltage settings of the degraded grid voltage relays which are shown in
TS Table 3.5-1, Engineering Safety Feature System Initiation Instrument
Setting Limits, Item 6b.
Date of issuance: April 14, 1995.
Effective date: April 14, 1995.
Amendment No.: 161.
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1995 (60 FR
11692). The licensee's March 2, 1995, submittal provided clarifying
information that did not affect the no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 14, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.

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

Date of application for amendments: June 9, 1994.
Brief description of amendments: The amendments implement a partial
application of the Generic Electric ARTS (Average Power Range Monitor
(APRM)/Rod Block Monitor (RBM)/Technical Specification) Improvement
Program. Four new ARTS thermal limits replace the existing flow-
referenced APRM trip setpoint setdown requirements and the Minimum
Critical Power Ratio (MCPR) Kf factor.
Date of issuance: April 13, 1995.
Effective date: Immediately to be implemented within 60 days.
Amendment Nos.: 103 and 88.
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39582). [[Page 20534]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan

Date of application for amendment: July 29, 1993, as supplemented
October 8, 1993.
Brief description of amendment: The amendment revises the Technical
Specification ACTION STATEMENTS related to operability of the control
room emergency filtration system. A portion of the amendment request
was denied. A separate Notice of Denial of Amendment has been sent to
the Federal Register for publication.
Date of issuance: March 31, 1995.
Effective date: March 31, 1995, with full implementation within 45
days.
Amendment No.: 103.
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43925). The October 8, 1993, letter provided clarifying information
within the scope of the original submittal and did not change the
staff's initial no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.

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

Date of application for amendments: January 18, 1995.
Brief description of amendments: The amendments revise TS Table
4.3-3 to allow the analog channel operational test interval for
radiation monitoring instrumentation to be increased from monthly to
quarterly, and are consistent with the guidance in Generic Letter 93-
05, ``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation.''
Date of issuance: April 3, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 154 and 136.
Facility Operating License Nos. NPF-9 and NPF-17. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11132).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

Date of application for amendment: July 22, 1993.
Brief description of amendment: The amendment revised operability
requirements for the Reactor Protection System and the Engineered
Safety Features Acturation System.
Date of issuance: April 3, 1995.
Effective date: April 3, 1995.
Amendment No.: 159.
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46229).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida

Date of application for amendments: July 19, 1994, resubmitted
October 20, 1994, and supplemented February 20, 1995.
Brief description of amendments: These amendments relate to the
maximum allowable reactor thermal power operation with inoperable main
steam safety valves.
Date of issuance: April 11, 1995.
Effective date: April 11, 1995.
Amendment Nos.: 172 and 166.
Facility Operating Licenses Nos. DPR-31 and DPR-41. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 23, 1994 (59
FR 60380).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 11, 1995. The February 20, 1995
submittal provided additional information that did not change the
staff's proposed no significant hazards consideration determination.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa

Date of application for amendment: October 28, 1994.
Brief description of amendment: The amendment revised the license
by deleting the ``Plan for the Integrated Scheduling of Plant
Modifications for the Duane Arnold Energy Center,'' as a condition of
the license.
Date of issuance: April 3, 1995.
Effective date: April 3, 1995.
Amendment No.: 208.
Facility Operating License No. DPR-49. Amendment revised the
license.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11134).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street SE., Cedar Rapids, Iowa 52401.

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

Date of application for amendment: December 6, 1994, as
supplemented on February 27, and April 4, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications to allow the use of the Combustion Engineering sleeving
process for repairing steam generator tubes. (The current requirement
specifies that degraded steam generator tubes be repaired by plugging.)
Date of issuance: April 14, 1995.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 149.
Facility Operating License No. DPR-36. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 1995 (60 FR
3673). The February 27, and April 4, 1995, [[Page 20535]] letters
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 14, 1995.
No significant hazards consideration comments received: Yes.
Comments were provided by letter dated February 17, 1995, from the
Maine State Nuclear Safety Inspector, Office of Nuclear Safety,
Division of Health Engineering, Department of Human Services. The NRC
staff responded to his comments in its letter dated March 15, 1995.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.

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

Date of application for amendment: September 30, 1994, as
supplemented February 13, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications to (1) clarify the definition of core alterations, (2)
change the verbiage in the Limiting Condition For Operation (LCO)
addressing the refueling operations, (3) make changes of surveillance
requirements involving source range instrumentation, and (4) change the
LCO regarding the Residual Heat Removal and coolant circulation water
levels to be consistent with the guidance provided in NUREG-1431,
Standard Technical Specifications for Westinghouse plants.
Date of issuance: April 12, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 107.
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55877). The February 13, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Learning Resources Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.

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

Date of application for amendment: January 26, 1993, as
supplemented August 4, 1993.
Brief description of amendment: The amendment changes the Technical
Specifications governing electrical power systems, AC and DC power
sources, and onsite power distribution for shutdown conditions (modes 5
and 6).
Date of issuance: April 12, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 108.
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1993 (58 FR
28058). The August 4, 1993, submittal provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: April 23, 1993.
Description of amendment request: The amendment revises the
Appendix A Technical Specifications to allow longer surveillance test
intervals and allowed outage times for the reactor protection system
and the engineered safety features actuation system. Also, the
amendment removes the requirement to perform the reactor trip system
analog channel operational test on a staggered basis.
Date of issuance: April 10, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 36.
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41507).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 10, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library, 47
Front Street, Exeter, NH 03833.

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

Date of application for amendments: February 16, 1994 (Reference
LAR 94-03)
Brief description of amendments: The amendments revise TS 4.6.1.2,
``Containment Integrity,'' to allow a more flexible schedule for
testing the primary containment integrated leakage rate.
Date of issuance: April 11, 1995.
Effective date: April 11, 1995, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 1--Amendment No. 99; Unit 2--Amendment No. 98
Facility Operating License Nos. DPR-80 and DPR-82. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14893).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 11, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.

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

Date of application for amendments: August 17, 1994.
Brief description of amendments: The amendments revise the combined
technical specifications (TS) to change TS 3/4.4.9.1, ``Reactor Coolant
System--Pressure/Temperature Limits,'' Figures 3.4-2, ``Reactor Coolant
System Heatup Limitations--Applicable Up to 8 EFPY,'' and 3.4-3,
``Reactor Coolant System Cooldown Limitations--Applicable Up to 8
EFPY,'' to extend the applicability up to 12 effective full-power years
(EFPYs). TS 3/4/4/9/3, ``Overpressure Protection Systems,'' is revised
to [[Page 20536]] specify a new low-temperature overprotection (LTOP)
system actuation pressure setpoint. The associated Bases were also
appropriately revised. Additionally, TS 3/4.1.2.2, ``Flow Paths--
Operating;'' TS 3/4.1.2.4, ``Charging Pumps--Operating;'' TS 3/4.4.1.3,
``Hot Shutdown;'' TS 3/4.4.1.4.1, ``Cold Shutdown--Loops Filled;'' TS
3/4.4.9.3, ``Overpressure Protection Systems;'' and TS 3/4.5.3, ``ECCS
Subsystems--Tavg Less than 350 Degrees F,'' are revised to specify
a new LTOP system enable temperature.
Date of issuance: April 13, 1995.
Effective date: April 13, 1995.
Amendment Nos.: Unit 1--Amendment No. 100; Unit 2--Amendment No.
99.
Facility Operating License Nos. DPR-80 and DPR-82. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 1994 (59 FR
51622).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 13, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.

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

Date of application for amendment: June 13, 1994.
Brief description of amendment: The amendment removes License
Condition 2.E from the Facility Operating License. License Condition
2.E incorporated the requirements of U.S. Department of Interior
publication ``Environmental Criteria for Electric Transmission
Systems''--1970, which applies to the construction cleanup,
restoration, and maintenance of transmission lines. The NRC staff has
determined that removing this condition from the Facility Operating
License has no bearing on plant safety or the health and safety of the
public, and is therefore acceptable.
Date of issuance: March 31, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 224.
Facility Operating License No. DPR-59. Amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11140).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama

Date of amendments request: June 10, 1995.
Brief Description of amendments: The amendments allow modifications
to be made for both units to relocate the lower level steam generator
water level taps during the upcoming refueling outages. These
modifications affect the Technical Specifications associated with the
reactor trip system and engineered safety feature actuation system
setpoints.
Date of issuance: April 7, 1995.
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 114 and 105.
Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: March 6, 1995 (60 FR
12253).
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated April 7, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Southern Nuclear Operating Company, Inc., Joseph M. Farley Nuclear
Plant, Unit 2, Houston County, Alabama

Date of application for amendment: December 7, 1994, as
supplemented February 14 and March 20, 1995.
Brief description of amendment: The December 7, 1994, submittal
requested a permanent change to the Technical Specifications for both
units related to steam generator tube support plate voltage-based
repair criteria in accordance with the draft Generic Letter on this
issue.
Date of issuance: April 7, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 106.
Facility Operating License No. NPF-8. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8754). The February 14 and March 20, 1995, letters provided
clarifying information that did not change the scope of the original
December 7, 1994, application and the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 7, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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

Date of application for amendments: December 16, 1994; supplemented
February 10, 1995 (TS 94-07).
Brief description of amendments: The amendments revise the
technical specifications to reduce the high reactor power level
setpoints when one or more main steam safety valves are inoperable and
incorporate related changes.
Date of issuance: April 4, 1995.
Effective date: April 4, 1995.
Amendment Nos.: 196 and 187.
Facility Operating License Nos. DPR-77 and DPR-79. Amendments
revise the technical specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11140).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 4, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: February 14, 1994 (Reference LAR 94-002,
TXX-94008), as supplemented by letters dated May 17, 1994 (Reference
TXX-94142), and April 3, 1995 (TXX-95098).
Brief description of amendments: The amendments revised Technical
Specification (TS) 3/4.2.4, ``Quadrant Power Tilt Ratio,'' by replacing
the existing TS and associated Bases concerning the quadrant power tilt
ratio with a TS consistent with the improved Standard Technical
Specifications (NUREG-1431). [[Page 20537]]
Date of issuance: April 4, 1995.
Effective date: April 4, 1995.
Amendment Nos.: Unit 1--Amendment No. 36; Unit 2--Amendment No. 22.
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37087).
The additional information contained in the supplemental letter
dated April 3, 1995, was clarifying in nature and thus, within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: February 28, 1995 (Reference LAR 95-01).
Brief description of amendments: These amendments replace
Surveillance Requirement (SR) 4.6.2.1b of Technical Specification (TS)
3/4.6.2, ``Depressurization and Cooling Systems--Containment Spray
System,'' with the corresponding SR from NUREG-1431. Bases Section 3/
4.6.2.1 ``Containment Spray System'' has also been revised to expand
the detail consistent with the corresponding Bases from NUREG-1431. The
SR, and its associated Bases, for confirming the performance of the
containment spray pumps are changed by replacing the specific pump head
and flow values with the general requirement that the pumps provide the
required head at the flow test point while the specific required values
are moved to the Comanche Peak Steam Electric Station Technical
Requirements Manual.
Date of issuance: April 6, 1995.
Effective date: April 6, 1995.
Amendment Nos.: Unit 1--Amendment No. 37; Unit 2--Amendment No. 23.
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 6, 1995 (60 FR
12255).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 6, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: April 22, 1994 (LAR 94-011, TXX-94116).
Brief description of amendments: These amendments changed
Surveillance Requirement 4.7.1.2 of Technical Specification 3/4.7.1.2
``Auxiliary Feedwater System,'' for the operational test frequency of
the motor driven and turbine driven pumps from ``at least once per 31
days on a STAGGERED TEST BASIS'' to ``at least once per 92 days on a
STAGGERED TEST BASIS.'' This change is consistent with ASME Section XI
requirements and Generic Letter 93-05, ``Line-Item Technical
Specifications Improvements to Reduce Surveillance Requirements for
Testing During Power Operations.''
Date of issuance: April 7, 1995.
Effective date: April 7, 1995.
Amendment Nos.: Unit 1--Amendment No. 38; Unit 2--Amendment No. 24.
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39598).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 7, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.

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

Date of application for amendment: December 9, 1994, as
supplemented on January 27, 1995.
Brief description of amendment: The amendment revises the Technical
Specification Surveillance Requirement 4.6.1.2.a and its associated
Bases. The change defers the requirement to perform the Type A
Containment Integrated Leak Rate Test until Refuel 8 (October 1996), in
conjunction with the exemption to 10 CFR Part 50, Appendix J.
Date of issuance: April 5, 1995.
Effective date: April 5, 1995.
Amendment No.: 98.
Facility Operating License No. NPF-30. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11141).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 5, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.

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

Date of amendment request: December 12, 1994.
Brief description of amendment: The amendment clarifies the
surveillance requirements for verifying the correct required position
for the valves in the auxiliary feedwater system.
Date of issuance: April 3, 1995.
Effective date: April 3, 1995, to be implemented within 30 days of
issuance.
Amendment No.: 85.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 1995 (60 FR
3677).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 3, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: April 11, 1994, as supplemented
on November 30, and December 22, 1994, and March 3, 1995.
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant (KNPP) Technical Specification (TS) 3.1.f,
``Minimum Conditions for Criticality,'' and its associated basis, by
specifying that the moderator temperature coefficient (MTC) shall be no
greater than 5.0 pcm/ deg.F when at or below 60% rated thermal power
and shall be zero or negative when above 60% rated thermal power.
Additionally, the MTC shall be no less [[Page 20538]] negative than -8
pcm/ deg.F for 95% of the cycle time at full power. The amendment also
incorporates required actions to be implemented, if the MTC
specification is not met.
Date of issuance: April 3, 1995.
Effective date: April 3, 1995.
Amendment No.: 117.
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49442).
The November 30, and December 22, 1994, and March 3, 1995,
submittals, provided clarifying information and expanded the basis
portion of the TS, but did not change the initial no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 3, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.

Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By May 26, 1995, the licensee
may file a request for a hearing with respect to issuance of the
amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene.
[[Page 20539]] Any person who has filed a petition for leave to
intervene or who has been admitted as a party may amend the petition
without requesting leave of the Board up to 15 days prior to the first
prehearing conference scheduled in the proceeding, but such an amended
petition must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of application for amendments: March 29, 1995.
Brief description of amendments: The amendments change the
Technical Specifications by revising the periodicity of the channel
functional test of the turbine driven auxiliary feedwater pump from
quarterly to each refueling outage.
Date of issuance: April 14, 1995.
Effective date: April 14, 1995.
Amendment Nos.: 161 and 149.
Facility Operating License Nos. DPR-39 and DPR-48. The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendments, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated April
14, 1995.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
NRC Project Director: Robert A. Capra.

Dated at Rockville, Maryland, this 19th day of April 1995.

For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 95-10127 Filed 4-25-95; 8:45 am]
BILLING CODE 7590-01-P

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Document Information

Published:: 04/26/1995
Department:: Nuclear Regulatory Commission
Entry Type:: Notice
Document Number:: 95-10127
Dates:: As of the date of issuance to be implemented within 30 days.
Pages:: 20513-20539 (27 pages)
PDF File:: 95-10127.pdf