# Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

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URL: https://www.frixlaw.com/law-library/documents/fr%3AX94-11123

## Record

- **Collection:** Federal Register
- **Document type:** Uncategorized Document
- **Published:** November 23, 1994

## Text

NUCLEAR REGULATORY COMMISSION
Biweekly Notice

Applications and Amendments to Facility Operating
LicensesInvolving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from October 31, 1994, through November 10, 1994.
The last biweekly notice was published on November 9, 1994 (59 FR
55865).

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

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By December 23, 1994, 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.
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, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: August 11, 1994
Description of amendment request: The proposed amendment deletes
the requirement to perform a five-year interval hydrostatic test on the
auxiliary coolant system critical headers from Technical Specification
Section 4.1.3, Table 4.1-3, Item 11.
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. The proposed change will delete the
requirement to perform a hydrostatic test on the component cooling
water [CCW] system at five year intervals to ensure the integrity of
the system. However, adequate testing of the system is provided as
required by the ASME Code Section XI. This testing includes a 10-
year system hydrostatic test as well as a 40-month interval system
inservice test and provides assurance of system integrity and the
ability to perform the intended function. Therefore, there is 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. The proposed change will delete the requirement to
perform a hydrostatic test on the component cooling water system at
five-year intervals to ensure the integrity of the associated system
headers. Operating characteristics of the system and its physical
configuration will remain unchanged, and the system will continue to
perform its intended function. There will be an overall decrease in
the frequency of testing the CCW system due to the elimination of
redundant testing and a decrease in operational activity associated
with testing the CCW system. Since there will be no functional or
hardware changes to the system, the proposed change will not create
the possibility of a new or different type of accident.
3. The proposed amendment does not involve a significant
reduction in the margin of safety. The proposed change will delete
the requirement to perform a hydrostatic test on the component
cooling water system at five-year intervals to ensure the integrity
of the system. However, adequate testing of the system is ensured by
the required ASME Code Section XI tests. This testing includes a 10-
year system hydrostatic test as well as a 40-month interval system
inservice test and provides assurance of system integrity and the
ability to perform the intended function. Therefore, there will be
no reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman

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: October 24, 1994
Description of amendment request: The proposed amendment would
remove Technical Specifications (TS) 3.3.4, Turbine Overspeed
Protection; TS 3.7.12, Area Temperature Monitoring; and TS 3.11.2.6,
Gas Storage Tanks; and their associated bases; and relocate them to
licensee-controlled documents, such as the Final Safety Analysis
Report.
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.
The proposed changes will simplify the TS, and implement the
recommendations of the Commission's Final Policy Statement on TS
Improvements. Since the elements of these TS are being relocated to
licensee-controlled documents any future changes would be controlled
under 10 CFR 50.59. The proposed changes are administrative in
nature and do not involve any modifications to any plant equipment
or affect plant operation. Therefore, 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.
The proposed changes are administrative in nature, do not
involve any physical alterations to plant equipment, and result in
no change in the method by which any safety-related system performs
its function. 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.
These changes do not affect any Final Safety Analysis Report
(FSAR) Chapter 15 accident analyses or have any impact on margin as
defined in the Bases to the Technical Specifications. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: William H. Bateman

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

Date of amendment request: October 27, 1994
Description of amendment request: The proposed amendments will
improve consistency throughout the Technical Specifications and their
related Bases by removing outdated material and blank pages,
incorporating minor changes in text, making editorial corrections, and
resolving other inconsistencies identified by the plant operations
staff.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if operation of the facility in accordance with the
proposed amendment would not: (1) involve a significant increase in
the probability or consequences of an accident previously evaluated;
or (2) create the possibility of a new or different kind of accident
from any accident previously evaluated; or (3) involve a significant
reduction in a margin of safety. Each standard is discussed as
follows:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments consist of administrative changes to the
Technical Specifications (TS) for St. Lucie Units 1 and 2. The
amendments will update the index and remove blank pages; implement
minor changes in text to rectify reference, nomenclature, spelling,
and/or consistency-in-format errors; and otherwise improve
consistency within the TS for each unit. The proposed amendments do
not involve changes to the configuration or method of operation of
plant equipment that is used to mitigate the consequences of an
accident, nor do the changes otherwise affect the initial conditions
or conservatisms assumed in any of the plant accident analyses.
Therefore, operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed administrative revisions will not change the
physical plant or the modes of plant operation defined in the
Facility License for each unit. The changes do not involve the
addition or modification of equipment nor do they alter the design
or operation of plant systems. Therefore, operation of the facility
in accordance with the proposed amendments would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendments are administrative in nature and do not
change the basis for any technical specification that is related to
the establishment of, or the preservation of, a nuclear safety
margin. Therefore, operation of the facility in accordance with the
proposed amendments would not involve a significant reduction in a
margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Mohan C. Thadani, Acting

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

Date of amendment request: October 20, 1994.
Description of amendment request: This supersedes the licensee's
original request dated July 19, 1994, and Noticed in the Federal
Register on August 3, 1994 (59 FR 39587). The licensee proposes to
change Turkey Point Units 3 and 4 Technical Specifications and its
associated BASES, which address the maximum allowed reactor thermal
power operation with inoperable main steam safety valves (MSSVs).
Westinghouse issued Nuclear Safety Advisory Letter 94-001 which
notified the licensee of a deficiency in the basis of the Turkey Point
Technical Specification 3/4.7.1, which allows the plant to operate at
reduced power levels with a specified number of MSSVs inoperable. This
amendment request corrects the allowable power level with inoperable
MSSVs and revises the TS to conform with the guidelines of the standard
technical specifications.
The licensee also proposed changes to TS 4.7.1.1 to indicate that
the provisions of TS 4.0.4 are not applicable for entry into mode 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed maximum allowable power level values will ensure that
the secondary side steam pressure will remain below 110 percent of
the design value following a Loss of Load/Turbine Trip event, when
one or more main steam safety valves (MSSVs) are declared
inoperable. The proposed change will not impact the classification
of the Loss of Load/Turbine Trip event as a Condition II probability
event (faults of moderate frequency) per ANSI--N18.2, 1973.
Accordingly, since the proposed maximum allowable power level will
maintain the capability of the MSSVs to perform their pressure
relief function associated with a Loss of Load/Turbine Trip event,
there will be no effect on the probability or consequences of an
accident previously evaluated.
The proposed addition of ACTION statement [a] to TS 3.7.1.1,
will not [a]ffect the probability or consequences of an accident
previously evaluated, since the proposed action is consistent with
the current Technical Specifications. Reducing the Power Range
Neutron Flux High Trip Setpoint to the maximum power level will
ensure the energy transfer to the most limiting steam generator is
not greater than the available relief capacity in that steam
generator. Entry into mode 3 does not require the availability of
the MSSV, since plant conditions (i.e., not operating at reactor
power) do not create the possibility of a secondary side
overpressurization event.
In addition, the proposed change to Surveillance Requirement
4.7.1.1, will not [a]ffect the probability or consequences of an
accident previously evaluated, since the proposed plant condition is
an analyzed shutdown condition. Entry into Mode 3 for surveillance
testing does not require the availability of the MSSV, since plant
conditions (i.e., not operating at reactor power) do not create the
possibility of a secondary side overpressurization event.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve any change to the configuration
of any plant equipment, and no new failure modes have been defined
for any plant system or component. The proposed maximum allowable
power level will maintain the capability of the MSSVs to perform
their pressure relief function to ensure the secondary side steam
design pressure is not exceeded following a Loss of Load/Turbine
Trip event. Therefore, since the function of the MSSVs is unaffected
by the proposed changes, the possibility of a new or different kind
of accident from any accident previously evaluated is not created.
The proposed addition of ACTION statement [a] to TS 3.7.1.1,
will not create the possibility of a new or different kind of
accident from any accident previously evaluated, since the proposed
action is consistent with the current Technical Specifications.
Reducing the Power Range Neutron Flux High Trip Setpoint to the
maximum power level will ensure the energy transfer to the most
limiting steam generator is not greater than the available relief
capacity in that steam generator. Entry into mode 3 does not require
the availability of the MSSV, since plant conditions (i.e., not
operating at reactor power) do not create the possibility of a
secondary side overpressurization event.
In addition, the proposed change to Surveillance Requirement
4.7.1.1, will not create the possibility of a new or different kind
of accident from any accident previously evaluated, since the
proposed plant condition is an analyzed shutdown condition. Entry
into Mode 3 for surveillance testing does not require the
availability of the MSSV, since plant conditions (i.e., not
operating at reactor power) do not create the possibility of a
secondary side overpressurization event.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes to the Technical Specifications do not
involve a significant reduction in a margin of safety. The algorithm
methodology used to calculate the maximum allowable power level is
conservative and bounding since it is based on a number of
inoperable MSSVs per loop; i.e., if only one MSSV in one loop is out
of service, the required action to reduce power to the maximum
allowable power level would be the same as if one MSSV in each loop
were out of service. Another conservatism with the algorithm
methodology is with the assumed minimum total steam flow rate
capability of the operable MSSVs. The assumption is that if one or
more MSSVs are inoperable per loop, the inoperable MSSVs are the
largest capacity MSSVs, regardless of which capacity MSSVs are
actually inoperable. Therefore, since the maximum allowable power
level calculated for the proposed changes using the algorithm
methodology are more conservative and ensure the secondary side
steam design pressure is not exceeded following a Loss of Load/
Turbine Trip event, this proposed license amendment will not involve
a significant reduction in a margin of safety.
The proposed addition of ACTION statement [a] to TS 3.7.1.1,
will not involve a significant reduction in a margin of safety,
since the proposed action is consistent with the current Technical
Specifications. Reducing the Power Range Neutron Flux High Trip
Setpoint to the maximum power level will ensure the energy transfer
to the most limiting steam generator is not greater than the
available relief capacity in that steam generator. Entry into mode 3
does not require the availability of the MSSV, since plant
conditions (i.e., not operating at reactor power) do not create the
possibility of a secondary side overpressurization event.
In addition, the proposed change to Surveillance Requirement
4.7.1.1, will not involve a significant reduction in the margin of
safety, since the proposed plant condition is an analyzed shutdown
condition. Entry into Mode 3 for surveillance testing does not
require the availability of the MSSV, since plant conditions (i.e.,
not operating at reactor power) do not create the possibility of a
secondary side overpressurization event.
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: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Mohan C. Thadani, (Acting)

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

Date of amendment request: October 20, 1994
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) by removing
the schedule for the withdrawal of reactor vessel material surveillance
specimens. The control of changes to this schedule, by way of a license
amendment to modify the TS, duplicates the requirements of Section
II.B.3 of Appendix H to Part 50 of Title 10 of the Code of Federal
Regulations (10 CFR). These proposed license amendments are consistent
with the guidance provided to licensees by NRC Generic Letter (GL) 91-
01, ``Removal of the Schedule for the Withdrawal of Reactor Vessel
Material Specimens from Technical Specifications.'' Additionally, these
amendments propose to correct typographical errors in the TS BASES and
to revise the reference in the TS BASES to the American Society for
Testing and Materials (ASTM) standard by which the fracture toughness
properties of the ferritic materials in the reactor vessels are
determined.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed license amendments do not involve a change in the
probability or consequences of accidents previously evaluated since
no physical changes to the plant, their operation, nor their
procedures are involved. The proposed changes are administrative in
nature and involve the activity of relocating, from the Turkey Point
Units 3 and 4 Technical Specifications (TS) to the Updated Final
Safety Analysis Report (UFSAR), the schedule for the withdrawal of
reactor vessel material surveillance specimens. The control of
changes to this schedule, by way of a license amendment to modify
the TS, duplicates the requirements of Section II.B.3 of Appendix H
to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR).
These proposed license amendments are consistent with the guidance
provided to licensees by NRC Generic Letter (GL) 91-01, ``Removal of
the Schedule for the Withdrawal of Reactor Vessel Material Specimens
from Technical Specifications.'' The TS BASES are also revised to
remove references to the table being removed from the TS. In
accordance with GL 91-01, FPL commits to maintain, the NRC-approved
version of the specimen withdrawal schedule in the Turkey Point
Units 3 and 4 UFSAR.
The current Turkey Point Units 3 and 4 TS BASES provide
background information on the use of the data obtained from material
specimens. This background information clearly defines the purpose
and relationship of this information to the requirements included in
the regulations and the ASME Code. Therefore, the removal of the
schedule for specimen withdrawal from the TS will not result in any
relaxation of the regulatory requirements of Appendix H to 10 CFR
Part 50 and do not involve an increase in the probability or
consequences of an accident previously evaluated.
The typographical corrections in the TS BASES and the revision
to the reference to ASTM E-185 are consistent with the guidance for
implementing administrative corrections to the TS to ensure that
references in the TS BASES are proper and correct.
In summary, operation of the facility in accordance with the
proposed amendment would not involve an increase in the probability
or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed license amendments do not create the possibility of
a new or different kind of accident from any accident previously
evaluated since no physical changes to the plant, their operation,
nor procedures are involved. The proposed changes are administrative
in nature and involve the activity of relocating, from the Turkey
Point Units 3 and 4 Technical Specifications (TS) to the UFSAR, the
schedule for the withdrawal of reactor vessel material surveillance
specimens. The control of changes to this schedule, by way of a
license amendment to modify the TS, duplicates the requirements of
Section II.B.3 of Appendix H to Part 50 of Title 10 of the Code of
Federal Regulations (10 CFR). These proposed license amendments are
consistent with the guidance provided to licensees by NRC GL 91-01,
``Removal of the Schedule for the Withdrawal of Reactor Vessel
Material Specimens from Technical Specifications.'' The TS Bases are
also revised to remove references to the table being removed from
the TS.
The removal from the TS of the schedule for the withdrawal of
reactor vessel material surveillance specimens will not result in
any loss of regulatory control because changes to this schedule are
controlled by the requirements of Appendix H to 10 CFR Part 50. In
addition, to ensure that the surveillance specimens are withdrawn at
the proper time, Surveillance Requirement 4.4.9.1.2 indicates that
the specimens shall be removed and examined to determine changes in
their material properties, as required by Appendix H. In accordance
with GL 91-01, FPL commits to maintain, the NRC-approved version of
the specimen withdrawal schedule in the Turkey Point Units 3 and 4
UFSAR.
The typographical corrections in the TS BASES and the revision
to the reference to ASTM E-185 are consistent with the guidance for
implementing administrative corrections to the TS to ensure that
references in the TS BASES are proper and correct.
The current Turkey Point Units 3 and 4 TS BASES provide
background information on the use of the data obtained from material
specimens. This background information clearly defines the purpose
and relationship of this information to the requirements included in
the regulations and the ASME Code. Therefore, the removal of the
schedule for specimen withdrawal from the TS will not result in any
relaxation of the regulatory requirements of Appendix H to 10 CFR
Part 50 and would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed license amendments do not involve physical changes
to the plant, their operation, nor their procedures. The proposed
license amendments do not create the possibility of a new or
different kind of accident from any accident previously evaluated
since no physical changes to the plant, their operation, nor their
procedures are involved. The proposed changes are administrative in
nature and involve the activity of relocating, from the Turkey Point
Units 3 and 4 Technical Specifications (TS) to the UFSAR, the
schedule for the withdrawal of reactor vessel material surveillance
specimens. The control of changes to this schedule, by way of a
license amendment to modify the TS, duplicates the requirements of
Section II.B.3 of Appendix H to Part 50 of Title 10 of the Code of
Federal Regulations (10 CFR). These proposed license amendments are
consistent with the guidance provided to licensees by NRC GL 91-01,
``Removal of the Schedule for the Withdrawal of Reactor Vessel
Material Specimens from Technical Specifications.'' The TS Bases are
also revised to remove references to the table being removed from
the TS.
The removal from the TS of the schedule for the withdrawal of
reactor vessel material surveillance specimens will not result in
any loss of regulatory control because changes to this schedule are
controlled by the requirements of Appendix H to 10 CFR Part 50. In
addition, to ensure that the surveillance specimens are withdrawn at
the proper time, Surveillance Requirement 4.4.9.1.2 indicates that
the specimens shall be removed and examined to determine changes in
their material properties, as required by Appendix H. In accordance
with GL 91-01, FPL commits to maintain the NRC-approved version of
the specimen withdrawal schedule in the Turkey Point Units 3 and 4
UFSAR.
The typographical corrections in the TS BASES and the revision
to the reference to ASTM E-185 are consistent with the guidance for
implementing administrative corrections to the TS to ensure that
references in the TS BASES are proper and correct.
The current Turkey Point Units 3 and 4 TS BASES provide
background information on the use of the data obtained from material
specimens. This background information clearly defines the purpose
and relationship of this information to the requirements included in
the regulations and the ASME Code. Therefore, the removal of the
schedule for specimen withdrawal from the TS will not result in any
relaxation of the regulatory requirements of Appendix H to 10 CFR
Part 50 and would not involve a reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Mohan C. Thadani, (Acting)

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

Date of amendment request: October 28, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 1.7, ``CORE ALTERATION,'' to
indicate that movement or replacement of incore instrumentation is not
considered to be a CORE ALTERATION provided that there are no fuel
assemblies in the associated core cell. TS 3/4.9.3, ``Control Rod
Position,'' and associated Bases would be revised to be consistent with
the proposed revision of TS 1.7 by changing the requirement to verify
that all control rods be inserted only during loading of fuel
assemblies into the core rather than during CORE ALTERATIONS. The
licensee has stated that these proposed changes are consistent with the
NRC's ``Improved Standard Technical Specifications,'' (NUREG-1434) and
those to be incorporated in Revision 1.The proposed amendment would
also revise Item 1.i.3) of TS Tables 3.3.2-1 and 4.3.2.1-1 to delete
the requirement showing that the Standby Liquid Control System (SLCS)
initiates Reactor Water Cleanup (RWCU) isolation in OPERATIONAL
CONDITION 5. License Amendment No. 48 issued on September 30, 1993,
deleted the requirement for SLCS to be OPERABLE in OPERATIONAL
CONDITION 5 but due to an oversight, failed to delete item 1.i.3) and
associated notations from TS Tables 3.3.2-1 and 4.3.2.1-1. The proposed
amendment would correct this oversight.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The purpose of the definition of CORE ALTERATION is to identify
operations which have the potential for adding reactivity to the
core while the vessel head is removed and fuel is in the vessel. The
proposed definition of CORE ALTERATION explicitly states that
movement of incore instruments and undervessel replacement is not
considered to be a CORE ALTERATION. The amount of fissile material
contained in any of these instruments is insignificant and thus
would not result in any change in reactivity of the core. Similarly,
control rod movement with no fuel assemblies in the associated core
cell has negligible impact on the reactivity of the remaining core.
Removal of a control rod by either the normal control rod drive
system or uncoupling and removing the blade from the top of the
vessel with no fuel in the associated cell is not considered a CORE
ALTERATION. It has negligible impact on the reactivity of the
remaining core and is not required to be covered by Specification 3/
4.9.3. In addition, the drop of a blade on irradiated fuel is
bounded by the fuel bundle drop.
The proposed change to Specification 3/4.9.3, ``Control Rod
Position,'' making it applicable only during loading of fuel
assemblies to reflect the remaining condition that results in the
addition of positive reactivity. Specification 3/4.9.1, ``Reactor
Mode Switch,'' requires the mode switch be locked in the refuel
position. This initiates the one-rod-out interlock which prevents
the selection of more than one control rod for movement.
Specification 3/4.1.1, ``Shutdown Margin,'' requires shutdown margin
be greater than or equal to 0.38% delta k/k analytically determined
or 0.28% delta k/k determined by test. These specifications ensure
that the reactor will not become critical when all control rods are
not inserted. Removal of the note referencing Special Test Exemption
3.10.3 is to be consistent with the revised definition.
The proposed change to eliminate RWCU isolation requirement upon
initiation of SLCS in OPERATIONAL CONDITION 5 is consistent with
Amendment 48, which eliminated the requirement for SLCS to be
OPERABLE in OPERATIONAL CONDITION 5.
Therefore, these changes will not involve a significant increase
in the probability or consequences of an accident from any
previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes to the definition of CORE ALTERATION and
Specification 3/4.9.3, ``Control Rod Position,'' and deletion of the
RWCU isolation requirement on SLCS initiation in OPERATIONAL
CONDITION 5 do not involve a physical change in any system's
configuration. Systems required to be OPERABLE for CORE ALTERATIONS
are still required to be OPERABLE, however, no new modes of
operation are introduced based on the proposed definition.
The purpose of the definition of CORE ALTERATION is to identify
operations which have the potential for adding reactivity to the
core while the vessel head is removed and fuel is in the vessel. The
proposed definition of CORE ALTERATION explicitly states that
movement of incore instruments and undervessel replacement is not
considered to be a CORE ALTERATION. The amount of fissile material
contained in any of these instruments is insignificant and thus
would not result in any change in reactivity of the core. Similarly,
control rod movement with no fuel assemblies in the associated core
cell has negligible impact on the reactivity of the remaining core.
Removal of a control rod by either the normal control rod drive
system or uncoupling and removing the blade from the top of the
vessel with no fuel in the associated cell is not considered a CORE
ALTERATION. It has negligible impact on the reactivity of the
remaining core and is not required to be covered by Specification 3/
4.9.3. In addition, the drop of a blade on irradiated fuel is
bounded by the fuel bundle drop.
The proposed change to Specification 3/4.9.3, ``Control Rod
Position,'' making it applicable only during loading of fuel
assemblies to reflect the remaining condition which results in the
addition of positive reactivity. Specification 3/4.9.1, ``Reactor
Mode Switch,'' requires the mode switch be locked in the refuel
position. This initiates the one-rod-out interlock which prevents
the selection of more than one control rod for movement.
Specification 3/4.1.1, ``Shutdown Margin,'' requires shutdown margin
be greater than or equal to 0.38% delta k/k analytically determined
or 0.28% delta k/k determined by test. These specifications ensure
that the reactor will not become critical when all control rods are
not inserted. Removal of the note referencing Special Test Exemption
3.10.3 is to be consistent with the revised definition.
The proposed change to eliminate RWCU isolation requirement upon
initiation of SLCS in OPERATIONAL CONDITION 5 is consistent with
Amendment 48, which eliminated the requirement for SLCS to be
OPERABLE in OPERATIONAL CONDITION 5.
Therefore, these changes will not create the possibility of a
new or different kind of accident from any previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant reduction in the
margin of safety.
The proposed definition of CORE ALTERATION clearly details what
constitutes a CORE ALTERATION. The definition is consistent with
NUREG-1433, ``Improved Standard Technical Specifications.'' The
definition has no impact on safety limits, setpoints, or plant
design and thus does not affect a margin of safety.
The proposed change to Specification 3/4.9.3, ``Control Rod
Position,'' making it applicable only during loading of fuel
assemblies to reflect the remaining condition that results in the
addition of positive reactivity. Specification 3/4.9.1, ``Reactor
Mode Switch,'' requires the mode switch be locked in the refuel
position. This initiates the one-rod-out interlock which prevents
the selection of more than one control rod for movement.
Specification 3/4.1.1, ``Shutdown Margin,'' requires shutdown margin
be greater than or equal to 0.38% delta k/k analytically determined
or 0.28% delta k/k determined by test. These specifications ensure
that the reactor will not become critical when all control rods are
not inserted, thus does not affect a margin of safety. The removal
of the note referencing Special Test Exemption 3.10.3 is consistent
with the revised definition.
Elimination of the requirement to initiate RWCU isolation based
upon SLCS initiation in OPERATIONAL CONDITION 5 is consistent with
deletion of the requirement to have the SLCS OPERABLE during
OPERATIONAL CONDITION 5. Therefore, there is no impact on a margin
of safety.
Therefore, based upon the above, these proposed 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Michael J. Case, Acting

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

Date of amendment request: October 4, 1994
Description of amendment request: The proposed amendment relocates
the primary containment isolation valve list from Technical
Specification (TS) Section 3.7.D to the Millstone Unit 1 technical
requirements manual (TRM). This change is in accordance with the
guidance of Generic Letter (GL) 91-08. The proposed amendment also
makes administrative and editorial changes to TS Section 3.7.D and
makes changes to the associated bases.
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 change in accordance with 10 CFR
50.92 and concluded that the change does 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 change does not involve a significant hazards consideration
because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed change will not result in any hardware or operating
changes. The proposed change is based upon Generic Letter 91-08 and
merely removes the containment isolation valve table and all
references to the table. The removal of the isolation valve table
from the technical specifications does not affect the operability
requirements of any of the listed valves. The technical
specifications will continue to require the isolation valves to be
OPERABLE. LCO's [limiting condition for operation] and surveillance
requirements for the valves will also remain in the technical
specifications. The containment isolation valve table will be
relocated to the Millstone Unit No. 1 TRM which is controlled in
accordance with 10 CFR 50.59.
This change is administrative in nature and does not involve an
increase in the probability or consequence of an accident previously
evaluated. Further, the proposed change does not alter the design,
function, or operation of the valves involved, and therefore does
not affect the probability or consequence of any previously
evaluated accident.
The clarification of Surveillance Requirement 4.7.D.2 ensures
that the flow path affected by an inoperable primary containment
isolation valve is isolated and maintained in the isolated
condition. This change ensures that probability or consequence of a
previously analyzed accident is not increased.
The nonintent changes involved with this license amendment
request are administrative in nature and will not, in and of
themselves, increase the probability or consequences of any
transient or accident previously analyzed. This does not affect or
have any potential impact upon any of the design basis types of
accidents previously analyzed. There are no failure modes affected
by the changes. As such, there are no design basis accidents
affected by the changes.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change will not impose any different operational or
surveillance requirements, nor will the change remove any such
requirements. The change proposes to relocate the containment
isolation valve list from the technical specifications to the TRM.
Adequate control of information is maintained. Further, as stated
above, the proposed change does not alter the design, function, or
operation of the valves involved, and therefore no new accident
scenarios are created.
The clarification of Surveillance Requirement 4.7.D.2 ensures
that the flow path affected by an inoperable primary containment
isolation valve is isolated and maintained in the isolated
condition. Since this change only ensures that the position of a
valve in the isolated condition is recorded, this change cannot
create a new or different kind of accident.
The nonintent changes do not, by their nature, modify plant
response during operation or during any transient or accident.
Therefore, there are no failure modes that can represent a new
unanalyzed accident.
3. Involve a significant reduction in the margin of safety.
The proposed change will not reduce the margin of safety since
it has no impact on any safety analysis assumption. The proposed
change does not decrease the scope of equipment currently required
to be operable or subject to surveillance testing, nor does the
proposed change affect any instrument setpoints or equipment safety
functions.
The relocation of the valve list is consistent with the guidance
provided in GL 91-08. The intent of the technical specification will
be met since the change will not alter function or operability
requirements for any primary containment isolation valve.
The clarification of Surveillance Requirement 4.7.D.2 ensures
that the flow path affected by an inoperable primary containment
isolation valve is isolated and maintained in the isolated
condition. Therefore, this change ensures that the margin of safety
established by the safety analyses is maintained.
The nonintent changes involved with this license amendment
request are administrative in nature and will not, in and of
themselves, reduce any margin of safety. There is no impact on the
performance of any safety system. There is no increase in the
consequences of any accident and, as such, there is no reduction in
the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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.
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

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

Date of amendment request: September 28, 1994
Description of amendment request: The proposed change would revise
the Surveillance Requirement 4.6.1.2.a of the Technical Specifications
to permit a more flexible schedule for containment leakage Type A
testing. The information in the associated Bases Section would also be
changed. In conjunction with this amendment request, the licensee has
requested a partial and schedular exemption, dated September 28, 1994,
from the requirements of Section III.D.1.(a) of Appendix J to Title 10
of the Code of Federal Regulations, Part 50.
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 basis for this conclusion is that the three criteria of
10 CFR 50.92(c) are not compromised. The proposed change does not
involve a SHC [significant hazards consideration] because the change
would not:
Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Type A tests are performed to ensure that the total leakage from
containment does not exceed the maximum allowable primary
containment leakage rate at a calculated peak containment internal
pressure permitted by the Millstone Unit No. 3 Technical
Specifications and FSAR [Final Safety Analysis Report]. This assures
compliance with the dose limits of 10CFR100.
The proposed change to Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 3 Technical Specifications will increase the
flexibility for scheduling the Type A tests. They do not modify the
maximum allowable leakage rate at the calculated peak containment
pressure, do not impact the design basis of the containment, and do
not change the post-accident containment response.
The first two Type A tests of the first 10-year service period
for Millstone Unit No. 3 have been conducted. The results of these
tests demonstrate that Millstone Unit No. 3 has maintained control
of containment integrity by maintaining margin between the
acceptance criterion and the ``As-Found'' and ``As-Left'' leakage
rates.
Historically, Type A tests have a relatively low failure rate,
where Type B and C testing (local leakage rate tests) could not
detect the leakage path. Most Type A test failures are attributed to
failures of Type B or C components (containment penetrations and
isolation valves). Type B and C components are tested per
Surveillance Requirement 4.6.1.2.d of the Millstone Unit No. 3
Technical Specifications. These tests are required to be conducted
at intervals no greater than 24 months, and the acceptance criterion
for the combined leakage rate for all penetrations and valves
subject to the Type B and C tests is 0.6 La. These local
leakage rate tests provide assurance that containment integrity is
maintained. The relatively low ``As-Left'' Type B and C total
leakage resulting from each successive outage indicates that the
leakage has been maintained within the technical specification
acceptance criterion, and demonstrates that improvements are
continually being made to the Type B and C program. The Type B and C
leakage results have decreased over the last three refueling
outages. This proposal does not request any changes to the
requirements for Type B and C testing. The Type B and C tests will
continue to be performed in accordance with the requirements of
Surveillance Requirement 4.6.1.2.d. These tests confirm that the
leak-tightness of the containment isolation valves and penetrations
has been maintained.
Based on the previous Type A, B, and C tests, the Millstone Unit
No. 3 containment's structural integrity is considered to be in
sound condition. No operations are known to have occurred which
would suggest any substantial degradation of these results.
Additionally, no structural modifications are planned for the next
refueling outage.
Based on the above, the proposed change to Surveillance
Requirement 4.6.1.2.a of the Millstone Unit No. 3 Technical
Specifications does not involve a significant increase in the
probability or consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change to Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 3 Technical Specifications will increase the
flexibility in scheduling the Type A tests. They do not make any
physical or operational changes to existing plant structures,
systems, or components. In addition, the proposed change does not
modify the acceptance criteria for the Type A tests. Maintaining the
leakage through the containment boundary to the atmosphere within a
specific value ensures that the plant complies with the requirements
of 10 CFR 100. The containment boundary serves as an accident
mitigator; it is not an accident initiator. Therefore, the proposed
change to Surveillance Requirement 4.6.1.2.a does not create the
possibility of a new or different kind of accident from any
previously analyzed.
3. Involve a significant reduction in the margin of safety.
The proposed change to Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 3 Technical Specifications will increase the
flexibility for scheduling the Type A tests. They do not modify the
maximum allowable leakage rate at the calculated peak containment
pressure, do not impact the design basis of the containment, and do
not change the post-accident containment response.
Based on the previous Type A, B, and C tests, the Millstone Unit
No. 3 containment's structural integrity is considered to be in
sound condition. No operations are known to have occurred which
would suggest any substantial degradation of these results.
Additionally, no structural modifications are planned for the next
refueling outage.
Based on the above, the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
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

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

Date of amendment request: September 29, 1994
Description of amendment request: The proposed change would remove
the sections from the Techical Specifications that are entitled
``Seismic Instrumentation'' and ``Meteorological Instrumentation'' and
relocate the information and testing requirements to the Salem Updated
Final Safety Analysis Report. The proposed change conforms with the NRC
guidance presented in the ``Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors'' published in
the Federal Register (58 FR 39132).
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 changes involve no hardware changes, no changes to
the operation of any systems or components, and no changes to
existing structures. Neither the relocation of the seismic/
meteorological specifications to the Salem UFSAR nor the elimination
of the Special Report requirements represent changes that affect
plant safety or alter existing accident analyses.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed changes are procedural in nature concerning the
operability and surveillance of instrumentation that are not safety
related and will not impact the operation of any plant safety
related component or equipment. Therefore, these changes will not
create a new or unevaluated accident or operating condition.
3. Will not involve a significant reduction in a margin of
safety.
In accordance with guidance provided by the NRC regarding the
improvement of Technical Specifications (58 FR 39132), the proposed
changes relocate the seismic and meteorological instrumentation
portion of the Technical Specification, with the exception of the
Special Report requirements, to the Salem UFSAR. These instruments
are not safety related and do not have any associated safety margins
which could be affected by this change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz

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

Date of amendment request: September 29, 1994
Description of amendment request: The proposed change to the
Technical Specifications revises the surveillance interval for
performing an air or smoke flow test through each containment spray
header from once every five years to once every ten years. The proposed
change implements a recommended line-item improvement from Generic
Letter 93-05, ``Line-Item Technical Specifications Improvements to
Reduce Surveillance Requirements for Testing During Power Operation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not affect the assumptions, design
parameters or results of UFSAR accidents analyzed. The proposed
change does not involve a hardware change, a change to the operation
of any system or component, or a change to an existing structure.
The proposed change leads to a reduction in radiation exposure to
plant personnel and the reduction of an unnecessary burden on plant
staff. The Containment Spray System header and nozzles are
fabricated from corrosion resistant stainless steel and are
maintained dry. Operating experience demonstrates that the proposed
increase in the Containment Spray surveillance test interval would
not affect operability of the system. Testing the Containment Spray
System header and nozzles at the proposed increased surveillance
interval does not increase the probability or consequences of an
accident previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed change does not modify equipment, affect the system
design basis or operability. This change does not alter parameters
utilized in the analyzed accident scenarios. The Containment Spray
System piping and nozzles are fabricated from corrosion resistant
stainless steel. The proposed change in surveillance frequency is
consistent with the guidance provided in GL 93-05. Testing the
Containment Spray System header and nozzles at the proposed
increased surveillance interval does not create the possibility of a
new of different kind of accident from those previously evaluated.
3. Does not involve a significant reduction in a margin of
safety.
The proposed change only involves a decrease in the surveillance
frequency and does not alter the performance of the surveillance
itself. System equipment and operation remains unchanged.
Operability and reliability is still maintained by periodic testing.
Testing the Containment Spray System header and nozzles at the
proposed increased surveillance interval does not involve a
significant reduction in the margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz

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

Date of amendment request: October 11, 1994

Description of amendment request: The proposed amendment would make
two Technical Specification changes concerning the pressurizer heaters.
The first change would add the phrase ``capable of being powered from
an emergency power supply'' to the Limiting Condition of Operation
(LCO) 3/4.4.4. The second change would alter the frequency of
surveillance requirement 4.4.4.2 from 92 days to every 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:

0. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The request (both proposed changes) does not change any
assumption or parameter assumed to function in any of the design/
licensing basis analysis.
The proposed change as described in section IA merely relocates
the requirement to supply emergency power to the required heater
group from the action to the LCO statement.
The change as described in section IB does not eliminate the
surveillance requirement, but extends its frequency from 92 days to
once per refueling outage in accordance with NRC recommendation. The
design of the Salem Station Pressurizer heaters is identical to that
described in the NUREG 1366 (Improvements to Technical
Specifications Surveillance Requirements, published December 1992),
and Generic Letter 93-05 (Line-Item Technical Specifications
improvements to Reduce Surveillance Requirements for Testing During
Power Operation, issued on September 27, 1993), and the extension of
the surveillance requirement is a recognized enhancement and
assurance to the continued reliability of the pressurizer heaters.
Based upon the above, PSE&G concludes that the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.2. Does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The proposed changes do not introduce any design or physical
configuration changes to the facility which could create new
accident scenarios.
3. Does not involve a significant reduction in a margin of
safety.

As stated in response to question number 1 above, the request
does not change any assumption or parameter assumed to function in
any of the design/licensing basis analysis. One change merely
relocates a requirement from one section of the LCO to another, and
the second change incorporates the recommendations and enhancements
as stated in NUREG 1366 and GL 93-05.
Consequently, PSE&G concludes that the change does not involve a
significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz

Saxton Nuclear Experimental Corporation, Docket No. 50-146, Saxton
Nuclear Facility, Bedford County, Pennsylvania

Date of amendment request: August 1, 1994. This supersedes the
request dated June 23, 1993.
Description of amendment request: The proposed amendment would
revise the technical specifications to allow characterization
activities related to the decommissioning of the Saxton Nuclear
Facility and add administrative activities associated with the
characterization activities.
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 involve a significant hazards
considerations because the changes would not:
1. Involve a significant increase in the probability or consequences of
an accident previously evaluated.
The activities associated with characterization of the facility
will have a minimum impact on the physical condition of the
containment vessel as it relates to the risk of fire and has no
effect on the risk of flooding.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
In its present condition, the only accidents applicable to the
site are fire, flood, and radiological hazard. The possibility of a
new or different type of accident than that previously evaluated in
the FSAR will not be created by the implementation of activities
permitted by the approval of this amendment request.
3. Involve a significant reduction in a margin of safety.
No margins of safety relevant to the equipment at the facility
exist. Activities involved in characterization will not involve a
reduction in a margin of safety.

The NRC staff has reviewed the analysis of the licensee and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Saxton Community Library, 911
Church Street, Saxton, Pennsylvania 16678Attorney for the Licensee:
Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge,
2300 N Street, NW, Washington, D.C. 20037
NRC Project Director: Seymour H. Weiss

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: October 20, 1994
Description of amendments request: The proposed Technical
Specification changes will delete requirements for the chlorine
detections systems from Technical Specification 3/4.3.3.6 and its
associated bases.
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. Removal of the control room chlorine detection system does
not involve a significant increase in the probability or
consequences of an accident previously evaluated because on-site
gaseous chlorine will be limited to a maximum per container
inventory of 150 pounds located greater than 100 meters from the
control room, and manual isolation of the control room is provided.
This is in compliance with Regulatory Guide 1.95. Furthermore,
offsite chlorine storage and transportation meets the requirements
of Regulatory Guides 1.78 and 1.95. Therefore, the probability of
occurrence of an accident is not affected.
There are no radiological consequences associated with chlorine
release accidents. Therefore, the consequences of an accident
previously evaluated are not increased.
2. Removal of the control room chlorine detection system does
not create the possibility of a new or different kind of accident
from any accident previously evaluated since the chlorine detectors
are utilized for detection of accidental chlorine release and are
not accident initiators. Gaseous chlorine has been removed from the
plant site, except for a permissible maximum per container inventory
of 150 pounds which will be located greater than 100 meters away
from the control room. In addition, there is a provision for the
manual isolation of the control room. Therefore, on-site chlorine
storage meets the requirements of Regulatory Guide 1.95.
Furthermore, offsite chlorine storage and transportation meet the
requirements of Regulatory Guides 1.78 and 1.95.

3. Removal of the control room chlorine detection system does
not involve a significant reduction in the margin of safety related
to the protection of control room operators from excessive levels of
chlorine since the onsite chlorine storage will be limited to a
maximum per container inventory of 150 pounds at the chlorination
house, which is located greater than 100 meters from the control
room. In addition, manual isolation of the control room is also
provided. This meets the requirements of Regulatory Guide 1.95.
Therefore, onsite and offsite chlorine storage and transportation
meets the requirements of Regulatory Guides 1.78 and 1.95.
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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: William H. Bateman

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

Date of amendment request: October 7, 1994 (TS 351)
Description of amendment request: The proposed amendment clarifies
the BFN diesel generator surveillance requirements which were thought
to be too ambiguous by both the NRC staff and TVA personnel. In
addition, the applicable Bases sections are being reviewed to provide
additional background information. TVA is revising Units 1 and 2 TS
Surveillance Requirements 4.9.B.3 and Unit 3 TS Surveillance
Requirement 4.9.B.2 to more closely reflect the requirements of
Improved Standard Technical Specifications (ISTS) for BWR/4s (NUREG-
1433), Section 3.8.1, AC Sources--Operating, Condition B for plant
operation with an inoperable diesel generator.
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.
The proposed change revises the surveillance requirements for
plant operation with an inoperable diesel generator. Diesel
generator operation is not a precursor to any design basis accident
or transient analyzed in the Browns Ferry Updated Final Safety
Analysis Report. Therefore, this change does not increase the
probability of any previously evaluated accident.
The proposed change will eliminate the requirement for
unnecessary diesel generator starts and the incumbent diesel
generator wear when a diesel generator is made inoperable for
planned maintenance and testing. Thus, the proposed change will
result in an increase in the reliability and availability of the
diesel generators. Therefore, this change does not increase the
consequences of any previously evaluated accident.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change to the surveillance requirements for plant
operation with an inoperable diesel generator does not involve a
modification to plant equipment. No new failure modes are
introduced. There is no effect on the function of any plant system
and no new system interactions are introduced by this change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will eliminate the requirement for
unnecessary diesel generator starts and the incumbent diesel
generator wear. Thus, the proposed change will result in an increase
in the reliability and availability of the diesel generators. Since
the ability of the diesel generators to perform their safety
function will not be degraded, the proposed amendment does not
involve a reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

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

Date of amendment request: November 2, 1994 (TS 94-17)
Description of amendment request: The proposed change would add
Operating License Condition 2.C.(25) to provide temporary extension of
the intervals for the surveillance tests specified in the submittal on
Unit 1 to coincide with the Cycle 7 refueling outage. The tests would
be extended to October 1, 1995, which would result in extension of the
specified 18-month, 36-month and 54-month surveillances to 29.5, 48 and
71.5 months, respectively.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change is temporary and allows a one-time extension
of specific surveillance requirements (SRs) for Cycle 7 to allow
surveillance testing to coincide with the seventh refueling outage.
The proposed surveillance interval extension will not cause a
significant reduction in system reliability nor affect the ability
of the systems to perform their design function. Current monitoring
of plant conditions and continuation of the surveillance testing
required during normal plant operation will continue to be performed
to ensure conformance with TS operability requirements. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Extending the surveillance interval for the performance of
specific testing will not create the posssibility of any new or
diffferent kind of accidents. No changes are required to any system
configurations, plant equipment, or analyses. Therefore, this change
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Surveillance interval extension will not impact any plant safety
analyses since the assumptions used will remain unchanged. The
safety limits assumed in the accident analyses and the design
function of the equipment required to mitigate the consequences of
any postulated accidents will not be changed since only the
surveillance test interval is being extended. Historical performance
generally indicates a high degree of reliability, and surveillance
testing perforned during normal plant operation will continue to be
performed to verify proper performance. Therefore, the plant will be
maintained within the analyzed limits, and the proposed extension
will not significantly reduce the margin of safety.
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
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia

Date of amendment request: October 25, 1994
Description of amendment request: The proposed change would extend
the functional surveillance frequency for the hydrogen recombiners from
once per 6 months to once per 18 months. The proposed changes would
also delete the surveillance requirement to operate the containment
purge blower. Also, minor editorial changes would be made to improve
the clarity and consistency between the NA-&2 Technical Specifications
(TS).
The NRC has completed a comprehensive examination of surveillance
requirements in the TS that require testing at power. The evaluation is
documented in NUREG-1366, ``Improvements to Technical Specification
Surveillance Requirements,'' dated December 1992. The NRC staff found,
that while the majority of testing at power is important, safety can be
improved, equipment degradation decreased, and an unnecessary burden on
personnel resources eliminated by reducing the amount of testing at
power that is required by the TS. Based on the results of the
evaluations documented in NUREG-1366, the NRC issued Generic Letter
(GL) 93-05, ``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation,'' dated
September 27, 1993.
The Hydrogen Recombiner System (HRS) removes the hydrogen gasses
that accumulate in the containment atmosphere following a design-basis
loss-of-coolant accident. Using the guidelines provided by GL 93-05,
Item 8.5 and NUREG-1366, the licensee is requesting a change to the
functional surveillance testing frequency for the hydrogen recombiners
from once per 6 months to once per 18 months. These changes in the
surveillance requirements do not affect plant or HRS operations. In
addition, several other changes are being requested for clarity and
consistency between NA-1&2 TS.
TS Surveillance Requirement 4.6.4.2.a states in part that ''...
each purge blower operates for 15 minutes.'' NA-1&2 are equipped with
two different types of ``purge blowers.'' One type of purge blowers is
an integral part of the HRS. These hydrogen recombiner purge blowers
are capable of exhausting containment gasses directly to the atmosphere
even with the recombiner incapable of removing hydrogen gas. The second
type of purge blowers is the containment purge blowers which exhaust
directly from the containment to atmosphere and are not associated with
the hydrogen recombiners. Surveillance Requirement 4.6.4.2.a will be
modified to state that the purge blowers being referred to in this
surveillance requirement are the hydrogen recombiner purge blowers.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Specifically, operation of North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Testing of the Hydrogen Recombiner System once per 18 months
will continue to assure that the Hydrogen Recombiner System will be
capable of performing its intended functions. The containment purge
blowers are not part of the Hydrogen Recombiner System and are not
assumed to function during accident conditions. Therefore, these
changes to the Hydrogen Recombiner System Technical Specifications
do not affect the probability or consequences of any previously
analyzed accident.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed Technical Specification changes do not involve any
physical modification of the plant or result in a change in a method
of operation. Testing the Hydrogen Recombiner System once per 18
months will continue to assure that the Hydrogen Recombiner System
will be capable of performing its intended function. Therefore, a
new or different type of accident is not made possible.
(3) Involve a significant reduction in a margin of safety.
The proposed Technical Specification changes do not affect any
safety limits or limiting safety system settings. System operating
parameters are unaffected. The availability of equipment required to
mitigate or assess the consequence of an accident is not reduced.
The containment purge blowers are not part of the Hydrogen
Recombiner System and are not assumed to function during accident
conditions. Testing of the Hydrogen Recombiner System once per 18
months will continue to assure that the Hydrogen Recombiner System
will be capable of performing its intended functions. Safety margins
are, therefore, not decreased.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Mohan C. Thadani, Acting

Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington

Date of amendment request: September 2, 1992
Description of amendment request: The proposed amendment would
revise the technical specifications to give the correct value for the
sodium pentaborate tank low-level alarm.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The change corrects the Technical Specifications to reflect the
correct and more conservative operating capability of the design. In
this instance there is no increase in the probability or
consequences of an accident previously evaluated because no changes
in concentration limits or volume are proposed by this change. The
Technical Specifications are being changed to recognize the more
prudent operating mode of the SLC [standby liquid control] storage
tank in that margin is available, and has always been available,
after a low level alarm. The margin allows corrective action to be
taken prior to exceeding Technical Specification limits. In summary,
a more prudent mode of operating is recognized by this change and
the design requirements of volume and concentration are not changed.
Hence, the accident analyses remains [sic] unaffected by this
change.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The SLC function and reliability are not affected by this
change. No new modes of plant operation are introduced with this
change. Hence, no new or different kind of accident is credible.
3. Does the change involve a significant reduction in a margin
of safety?
No change to the required volume and concentrations are being
proposed by this [modification]. Neither the original design or
accident analysis is affected by this change. A more prudent mode of
operation, that currently exists, is recognized by this proposal.
Therefore, there is no impact to a margin of safety with this
change.

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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, NW., Washington, D.C. 20005-3502
NRC Project Director: Theodore R. Quay Wolf Creek Nuclear Operating
Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey
County, Kansas
Date of amendment request: October 21, 1994 and supplement dated
October 27, 1994
Description of amendment request: This amendment request revises
Technical Specification Surveillance Requirements 4.7.1.2.1.c.2
(operability testing for the turbine-driven auxiliary feedwater (AFW)
pump automatic start feature) and 4.3.2.2 (engineered safety feature
actuation system instrumentation response time testing for the turbine-
driven AFW pump) to correct an inconsistency caused by system
limitations to supply steam to the turbine-driven AFW pump prior to
entry into Mode 3. These specifications are being revised to indicate
that the provisions of Technical Specification 4.0.4 are not applicable
for entry into Mode 3.
In addition, Technical Specification Surveillance Requirement
4.7.1.2.1.c is being revised to delete the requirement to be performed
during shutdown.
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.
WCNOC [Wolf Creek Nuclear Operating Corporation] is proposing to
modify Surveillance Requirements 4.3.2.2 and 4.7.1.2.1.c.2 by adding
an exemption for [from] the provisions of Technical Specification
4.0.4 and deleting the shutdown requirement. Entry into Mode 3 would
allow for appropriate test conditions (e.g., adequate steam pressure
available) to complete the operability testing of the turbine-driven
AFW pump. The acceptance criteria such as response time, or test
frequency, are not revised. Therefore, the surveillance will
continue to verify the operability of the turbine-driven AFW pump.
Additionally, the proposed changes are consistent with the new
improved Standard Technical Specifications for Westinghouse plants
(NUREG-1431)
Considering the above, the proposed changes to Surveillance
Requirements 4.3.2.2 and 4.7.1.2.1.c.2, of the WCGS [Wolf Creek
Generating Station] Technical Specifications, do not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes do not make any physical or operational
changes to existing plant structures, systems, or components. The
proposed changes do not introduce any new failure modes. They simply
allow tests to be performed at appropriate conditions rather than
during shutdown.
Additionally, the proposed changes do not modify the acceptance
criteria for the tests. The purpose of the tests is to ensure that
the turbine-driven AFW pump can perform its intended function.
Thus, the proposed changes do not create the possibility of a
new or different kind of accident from any previously analyzed.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes do not have any adverse impact on the
Updated Safety Analysis Report accident analyses. The applicable
acceptance criteria for the turbine-driven AFW pump will not be
modified by these proposed changes. The proposed changes will permit
the tests to be conducted under the proper conditions, so that the
ability of the turbine-driven AFW pump to perform its intended
safety function can be confirmed.
Based on the above discussions it has been determined that the
requested technical specification revision does not involve a
significant increase in the probability or consequences of an
accident or other adverse condition; or 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 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, NW., Washington, DC 20037
NRC Project Director: Theodore R. Quay

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.

Philadelphia Electric Company, Public Service Electric and Gas
Company, Delmarva Power and Light Company, and Atlantic City
Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of amendment request: October 25, 1994
Brief description of amendment request: The proposed amendment
would clarify the minimum reactor steam pressure required for
Surveillance Requirement (SR) 4.5.C.1(e). The revised SR will require
the licensee to verify that the High Pressure Coolant Injection Pump,
with reactor pressure less than or equal to 175 psig, develop a flow
rate of greater than or equal to 5000 gpm against a system head
corresponding to reactor pressure. The current SR specifies that the
test be performed at 150 psig but does not provide a range of
acceptable pressures.
Date of publication of individual notice in Federal Register: Nov.
7, 1994 (59 FR 55498)
Expiration date of individual notice: December 7, 1994
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.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.

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

Date of application for amendments: June 17, 1994, supplemented by
letter dated September 21, 1994,
Brief description of amendments: The amendments allow removal of
five tables of component lists from the Palo Verde Technical
Specifications (TS) in accordance with NRC Generic Letter (GL) 91-08,
``Removal of Component Lists from Technical Specifications.'' The
affected tables are Table 3.3-9B, Table 3.3-9C, Table 3.6-1, Table 3.8-
2, and Table 3.8-3. These five removed tables will be incorporated into
a new document, which will be administratively controlled according to
the change control provisions of the TS.
Date of issuance: October 31, 1994
Effective date: October 31, 1994, to be implemented no later than
45 days from the date of issuance.
Amendment Nos.: 85, 73, and 57
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37061) The supplemental letter provided certain revised TS pages for
clarification purposes and did not change the original no significant
hazards determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated October 31, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004

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

Date of application for amendments: January 4, 1994
Brief description of amendments: These amendments revise Technical
Specification 3.2.3, ``Azimuthal Power Tilt,'' to change the azimuthal
power tilt limit from less than or equal to 10 percent to less than or
equal to 3 percent when the core operating limit supervisory system is
out of service. The associated TS Bases are similarly changed.
Date of issuance: November 3, 1994
Effective date: November 3, 1994, to be fully implemented no later
than 45 days from the date of issuance
Amendment Nos.: 86, 74, and 58
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22001) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 3, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004

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: May 27, 1993
Brief description of amendment: The amendment revises the heatup
and cooldown curves and the low-temperature overpressure protection
(LTOP) controls. The changes to the LTOP controls support proposed
modifications to allow a variable-setpoint (VLTOP) protection system.
The VLTOP system will increase the allowable operating pressure band in
the LTOP region and increase the flexibility in the use of the reactor
coolant pumps.
Date of issuance: November 1, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 178
Facility Operating License No. DPR-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (59
FR 37064) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 1, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.

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

Date of application for amendment: November 4, 1993, as
supplemented April 27, 1994, and October 10, 1994.
Brief description of amendment: The proposed amendment revises
Technical Specification 6.13.1 to provide use of alarming dosimeters in
high radiation areas. This change includes newly revised 10 CFR Part 20
requirement references and is consistent with NUREG-1413, Standard
Technical Specifications--Westinghouse Plants, Specification 5.11.1.
Date of issuance: November 4, 1994
Effective date: November 4, 1994
Amendment No.: 152
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4935) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 4, 1994. No significant hazards
consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College, Hartsville, South Carolina 29550

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 application for amendment: September 28, 1993, as amended
April 5, 1994.
Brief description of amendment: The amendment revises Technical
Specification 3/4.8.1, ``AC Sources--Operating'', and associated Bases
to be consistent with the new ``Standard Technical Specifications for
Westinghouse Plants'', NUREG-1431, Revision 0.
Date of issuance: November 4, 1994
Effective date: November 4, 1994
Amendment No. 51
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57845) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 4, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Iowa-Illinois Gas and Electric
Company, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power
Station, Units 2 and 3, Grundy County, Illinois; Docket Nos. 50-254
and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock
Island County, Illinois; Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of application for amendments: July 8, 1994
Brief description of amendments: The amendment revises the
operating licenses by adding a license condition that would allow the
commitments made in response to NUREG-0737, ``Clarification of TMI
Action Plan Requirements,'' to be controlled pursuant to the
requirements of 10 CFR 50.59.
Date of issuance: November 3, 1994
Effective date: November 3, 1994
Amendment Nos.: for Dresden, Amendment Nos. 129 and 123; for Quad
Cities, Amendment Nos. 150 and 146; and for Zion, Amendment Nos. 158
and 146.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29, DPR-30,
DPR-39, and DPR-48. The amendments revised the operating licenses.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45021) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 3, 1994. No significant
hazards consideration comments received: No
Local Public Document Room locations: for Dresden, the Morris
Public Library, 604 Liberty Street, Morris, Illinois 60450; for Quad
Cities, the Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois
61021; and for Zion, the Waukegan Public Library, 128 N. County Street,
Waukegan, Illinois 60085.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Power Plant,
Unit 1, Monroe County, Michigan

Date of application for amendment: December 9, 1993 (Reference NRC-
93-0143).
Brief description of amendment: This amendment modified the
Technical Specifications (TS) incorporated in Possession-Only License
No. DPR-9 as Appendix A by modifying the Protected Area definition and
Waste Disposal Surveillances to provide the appropriate 10 CFR Part 20
references in conformance with a revision of 10 CFR Part 20 (56 FR
23360).
Date of issuance: November 3, 1994.
Effective date: This license amendment is effective as of the date
of its issuance and must be fully implemented no later than 30 days
from the date of issuance.
Amendment No.: 10.Possession-Only License No. DPR-9: The amendment
revised the TS.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37070) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 3, 1994.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: November 21, 1991.
Brief description of amendments: The amendments were submitted as a
result of NRC recommendations pertaining to Generic Letter 90-06 for
the power-operated relief valves and block valves and low-temperature
overpressure protection systems.
Date of issuance: October 27, 1994
Effective date: October 27, 1994
Amendment Nos.: 150 and 132
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59748) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated Ocotber 27, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223

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

Date of application of amendments: December 8, 1993, as
supplemented April 20, September 8, 1994, and October 25, 1994.
Brief description of amendments: The amendments revise Technical
Specification 3.4 to address the need to bypass automatic initiation of
the Emergency Feedwater system with the main feedwater pump discharge
pressure is below actuation setpoint during startup and shutdown in
order to prevent inadvertent actuation. The amendments also deleted
operability requirements for the Emergency Condenser Cooling Water
(ECCW) system.
Date of Issuance: October 31, 1994
Effective date: To be implemented within 30 days from the date of
issuance.
Amendment Nos.: 207, 207, and 204
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39584) The April 20, September 8, and October 25, 1994 supplements
provided additional information that did not change the scope of the
December 8, 1994, application and the initial proposed no significant
hazards consideration determination.The Commission's related evaluation
of the amendments is contained in a Safety Evaluation dated October 31,
1994. No significant hazards consideration comments received: No
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691

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

Date of amendment request: February 5, 1993, as supplemented by
letter dated August 1, 1994.
Brief description of amendment: The amendment revised the Technical
Specifications to incorporate a technical review and control process to
supplement the onsite technical review and approval of new procedures
and changes thereto affecting nuclear safety.
Date of issuance: November 4, 1994Effective date: November 4, 1994
Amendment No.: 100
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45022) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 4, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: May 23, 1994
Brief description of amendment: This amendment revises Technical
Specifications Section 3/4.7.1.1, Turbine Cycle, Safety Valves, to
delete a specific reference to the 1994 edition of the ASME Code and
refer to testing in accordance with Technical Specification 4.0.5, the
In-Service Inspection and In-Service Testing Specification.
Date of Issuance: November 1, 1994
Effective Date: November 1, 1994
Amendment No.: 68
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34664) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 1, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

Date of application for amendment: June 30, 1994
Brief description of amendment: The proposed amendment would
clarify the requirement for the audit of conformance to Technical
Specifications, delete the requirement for Safety Committee oversight
of the Emergency Plan and Security Plan and allow designation by the
Plant Superintendent signature authority for procedure approval.
Date of issuance: November 2, 1994
Effective date: Date of issuance and to be implemented within 60
days
Amendment No.: 202
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39591) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 2, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois

Date of application for amendment: August 12, 1994
Brief description of amendment: The amendment modifies Clinton
Power Station Technical Specification 3/4.6.2.2, ``Drywell Bypass
Leakage,'' to allow drywell bypass leakage rate tests to be performed
at intervals as long as five years based on the demonstrated
performance of the drywell structure.
Date of issuance: November 3, 1994
Effective date: November 3, 1994
Amendment No.: 94
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49428) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 3, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.

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

Date of application for amendments: November 12, 1993

Brief description of amendments: The amendments revise the
Technical Specifications for the accumulators to allow extended action
time for improper boron concentration, to provide a consistent action
statement for both units, and to modify the surveillances on the boron
concentration and the isolation valve.

Date of issuance: November 8, 1994

Effective date: November 8, 1994

Amendment Nos.: 184 and 169

Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.

Date of initial notice in Federal Register: December 22, 1993 (58
FR 67848). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 8, 1994.No significant
hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

Date of application for amendments: December 22, 1993
Brief description of amendments: The amendments revise the action
statement in the Technical Specifications for Steam Generator Stop
Valves to be more consistent with NUREG-1431, Standard Technical
Specifications Westinghouse Plants. The proposed changes allow both
greater time for compensatory action as well as operation in Modes 2
and 3 with valves inoperable but closed. A Unit 2 action requirement is
also revised.
Date of issuance: November 8, 1994
Effective date: November 8, 1994
Amendment Nos.: 185 and 170
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4939) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 8, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.

Niagara Mohawk Power Corporation, Docket Nos. 50-220, and 50-410,
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County,
New York

Date of application for amendments: June 9, 1994
Brief description of amendments: The amendments modify paragraph
2.D(4) of
Facility Operating License No. DPR-63 and paragraph 2.E of
Facility Operating License No. NPF-69 to require compliance with
the amended Physical Security Plan. The changes involve the number of
armed security force members that comprise the response force for each
shift at the site.
Date of issuance: October 31, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1--150--Unit 2--58
Facility Operating License Nos. DPR-63 and NPF-69: Amendments
revise the Facility Operating Licenses.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49432) The Commission's related evaluation of the amendments is
contained in a Safeguards Evaluation Report dated October 31, 1994. 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.

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

Date of application for amendment: September 26, 1994
Brief description of amendment: The amendment revises the Technical
Specifications (TS) by adding a footnote to Surveillance Requirement
4.6.1.2.d that defers the performance of Type B and C Containment leak
rate tests to the end of the twelfth re

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3AX94-11123. Public record. Not legal advice.