[Federal Register Volume 59, Number 225 (Wednesday, November 23, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-11123]


[[Page Unknown]]

[Federal Register: November 23, 1994]


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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 refueling outage.
    Date of issuance: October 31, 1994
    Effective date: October 31, 1994
    Amendment No.: 181
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (59 FR 52005, October 13, 1994) 
That notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice also provided 
for an opportunity to request a hearing by November 14, 1994, but 
indicated that if the Commission makes a final no significant hazards 
consideration determination any such hearing would take place after 
issuance of the amendment. The Commission's related evaluation of the 
amendment, finding of exigent circumstances, and final determination of 
no significant hazards consideration are contained in a Safety 
Evaluation dated October 31, 1994.
    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.

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 1, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications concerning the Reactor Coolant System Volume.Date of 
issuance: November 8, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 182
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49432). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 8, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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

    Date of application for amendments: February 16, 1994 (Reference 
LAR 94-04)
    Description of amendment request: The proposed amendments revise 
the combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant, Unit Nos. 1 and 2. Specifically, TS 4.2.2, ``Heat Flux Hot 
Channel Factor--FQ(z),'' and 6.9.1.8, ``Core Operating Limits 
Report,'' are revised as follows: (1) The 2-percent FQ(z) penalty 
listed in TS 4.2.2.2.e.1) would be deleted and the statement revised to 
indicate the use of an appropriate factor to be specified in the Core 
Operating Limits Report (COLR). (2) TS 6.9.1.8.b.1 would be changed to 
reference Revision 1 of WCAP 10216-P-A, ``Relaxation of Constant Axial 
Offset Control FQ(z) Surveillance Technical Specification,'' dated 
February 1994.
    Date of issuance: October 31, 1994
    Effective date: 60 days from date of issuance.
    Amendment Nos.: 96 and 95
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17603) 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: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407No significant hazards 
consideration comments received: No.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17603) 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: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

PECO Energy Company, Docket No. 50-171, Peach Bottom Atomic Power 
Station, Unit 1, Peach Bottom, Pennsylvania.

    Date of application for amendment: May 9, 1994.
    Brief description of amendment: This amendment modified Possession-
Only License No. DPR-12 and the Technical Specifications (TS) 
incorporated as Appendix A by changing the name of Philadelphia 
Electric Company to PECO Energy Company, by providing the appropriate 
10 CFR Part 20 references, and by reducing the required frequency for 
performing periodic inspections in the containment vessel below ground 
level for water accumulation.
    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.: 8.Possession-Only License No. DPR-12: The amendment 
revised Possession-Only License No. DPR-12 and the TS.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45030) 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of application for amendments: September 16, 1994
    Brief description of amendments: These amendments extend the 
snubber functional testing interval from 18 months (+/- 25%) to 24 
months (+/- 25%) (plus or minus was published as [greater than or equal 
to] in the initial Federal Register notice), and increase the sample 
plan size from 10 percent to 13.3 percent. The combination of these two 
changes will ensure that the entire population of snubbers will be 
tested in a 15-year period.
    Date of issuance: November 2, 1994
    Effective date: November 2, 1994Amendment Nos. 81 and 42
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 30, 1994 (59 
FR 50019) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 2, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: September 16, 1992, supplemented 
June 27, 1994, and September 26, 1994
    Brief description of amendment: The amendment revised Technical 
Specifications Section 4.6.B (Emergency Power System Periodic Tests--
Station Batteries) to incorporate changes which allow battery testing 
surveillance interval extensions to accommodate operation on a 24-month 
fuel cycle. These changes followed the guidance provided in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle,'' as applicable.
    Date of issuance: November 2, 1994
    Effective date: November 2, 1994
    Amendment No.: 155
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 28, 1992 (57 FR 
48825) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 2, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: July 21, 1994, as supplemented 
September 26, 1994.
    Brief description of amendment: The amendment relocates fire 
protection requirements from the Technical Specifications to the plant 
fire protection program in accordance with the guidance provided in 
Generic Letter (GL) 86-10, ``Implementation of Fire Protection 
Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements 
from the Technical Specifications.'' The amendment also modifies the 
Facility Operating License to incorporate the standard fire protection 
license condition provided in GL 86-10.
    Date of issuance: November 3, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 218
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42345) 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: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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 application for amendments: August 19, 
1994, as supplemented October 4, 1994

    Brief description of amendments: The amendments reduce the minimum 
setpoints and allowable values for the steam generator low and low-low 
level reactor protection system signals.
    Date of issuance: November 4, 1994
    Effective date: November 4, 1994
    Amendment Nos. 159 and 140
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47180) The supplemental letter provides additional information but 
does not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated November 4, 1994. 
No significant hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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 application for amendments: June 13, 1994
    Brief description of amendments: These amendments add a new section 
3.0.6 to the technical specifications and the associated Bases, that 
permits an out-of-service component to be returned to service under 
administrative controls for the purpose of determining operability, and 
make an editorial correction.
    Date of issuance: November 8, 1994
    Effective date: November 8, 1994
    Amendment Nos. 160 and 141
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39590) 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: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendment: December 17, 1993
    Brief description of amendment: The change revises TS 3/4.3.3.6, 
Accident Monitoring Instrumentation, and its associated bases; 
relocates TS 3/4.6.5.1, Hydrogen Monitors, and TS 3/4.3.3.1, Tables 
3.3-6 and 4.3-3, Item 1.c, Reactor Building Area High Range Radiation 
Monitors, into the Accident Monitoring TS.
    Date of issuance: November 7, 1994
    Effective date: November 7, 1994
    Amendment No.: 118
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7699) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 7, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180.

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

    Date of application for amendments: May 16, 1994 (TS 94-03)
    Brief description of amendments: The amendments remove the response 
time limits for the reactor trip and engineered safety feature 
functions from the technical specifications in accordance with Generic 
Letter 93-08.
    Date of issuance: November 9, 1994
    Effective date: November 9, 1994
    Amendment Nos.: 190 and 182
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32236) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated November 9, 1994No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of amendment request: September 19, 1994
    Brief description of amendments: The amendments allow a one-time 
six-month extension for certain emergency diesel generator technical 
specification surveillance requirements and other related surveillance 
requirements. The one-time extension from 18 to 24 months for the 
affected surveillance requirements is applicable only to Unit 2, Train 
A, until completion of the second refueling outage for Unit 2.
    Date of issuance: November 2, 1994
    Effective date: Effective as of its date of issuance, to be 
implemented within 30 days.
    Amendment Nos.:  Unit 1--Amendment No. 31; Unit 2--Amendment No. 17
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 30, 1994 (59 
FR 50024) The October 20, 1994, submittal provided additional 
clarifying information and did not change the initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
November 2, 1994. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of amendment request: September 19, 1994
    Brief description of amendments: The amendments revise the 
technical specifications by eliminating the requirement that the 18-
month surveillance requirements (SRs) for the emergency core cooling, 
containment spray, spray additive, containment isolation valves, 
auxiliary feedwater and component cooling water systems be performed 
``during shutdown'' or ``during REFUELING MODE or COLD SHUTDOWN.'' The 
SRs are still required to be performed on an 18-month surveillance 
interval, but may be performed in any mode in which it is technically 
and operationally acceptable to perform the testing.
    Date of issuance: November 2, 1994
    Effective date: Within 30 days of its date of issuance
    Amendment Nos.:  Unit 1--Amendment No. 32; Unit 2--Amendment No. 18
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 30, 1994 (59 
FR 50022) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 2, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of application for amendment: March 31, 1994
    Brief description of amendment: The amendment revises the Kewaunee 
Nuclear Power Plant Technical Specifications (TS) by incorporating 
operability and surveillance requirements for the recently installed 
Auxiliary Feedwater Pump Low Discharge Pressure Trip instrumentation. 
Surveillance requirements were added to Table TS 4.1-1, ``Minimum 
Frequencies for Checks, Calibrations and Test of Instrument Channels.'' 
TS 3.4, ``Steam and Power Conversions System,'' has been revised to 
explicitly link operability of the associated Auxiliary Feedwater Pump 
Low Discharge Pressure Trip channel to operability of the associated 
auxiliary feedwater pump. In addition, minor format inconsistencies in 
TS 3.4.b.1.A and 3.4.b.1.B were corrected.
    Date of issuance: November 1, 1994
    Effective date: Date of issuance, to be implemented within 30 days
    Amendment No.: 112
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34671) 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: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

    Dated at Rockville, Maryland, this 16th day of November 1994.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II Office of Nuclear Reactor 
Regulation
[Doc. 94-28758 Filed 11-22-94; 8:45 am]
BILLING CODE 7590-01-F