[Federal Register Volume 61, Number 129 (Wednesday, July 3, 1996)]
[Notices]
[Pages 34884-34908]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-16879]


=======================================================================
-----------------------------------------------------------------------

UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from June 8, 1996, through June 21, 1996. The 
last biweekly notice was published on June 19, 1996 (61 FR 31171).

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

[[Page 34885]]

    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: May 1, 1996
    Description of amendment request: The proposed amendment would 
modify Table 3.1.1, ``Reactor Protection System (SCRAM) Instrumentation 
Requirement,'' Table 3.2.C.1, ``Instrumentation that Initiates Rod 
Blocks,'' and Technical Specification 3/4.4, ``Standby Liquid 
Control.''
    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?
    Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1
    The changes to Note 7 to Table 3.1.1 and the addition of Note 6 
to Table 3.2.C.1 are proposed to clarify their requirements, the 
appropriate action to take, and their relationship to plant modes. 
This revised scram and rod block applicability is acceptable because 
control rods withdrawn from a core cell containing no fuel 
assemblies have a negligible impact on the reactivity of the core, 
and, therefore, these features are not required to be operable (i.e. 
provide the capability to scram). Provided all rods otherwise remain 
inserted, the RPS [Reactor Protection System] functions serve no 
purpose and are not required. In this condition, the required 
shutdown margin (Specification 3.3.A.1) and the required one-rod-out 
interlock (Specification 3.10.A) ensure that no event requiring the 
RPS or Rod Block will occur.
    The Actions of Table 3.1.1 for inoperable equipment were 
previously revised in Amendment <greek-i>147 to be consistent with 
the improved STS [Standard Technical Specifications]. Action (A) 
requires fully inserting all insertable control rods in core cells 
containing one or more fuel assemblies. Since Specification 3.10.A 
requires all control rods to be fully inserted during fuel movement, 
the proposed applicable conditions cannot be entered while moving 
fuel. In addition, Specification 3.10.D used for controlling 
multiple control rod removal, requires all control rods in a 3X3 
array centered on the CRDs [Control Rod Drive] being removed to be 
fully inserted and electrically disarmed and all other control rods 
fully inserted. The only possible action is control rod withdrawal, 
which is addressed by Action A.
    Hence operating Pilgrim in accordance with the proposed changes 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Section 3/4.4
    The proposed change involves reformatting, renumbering, and 
rewording of the existing Technical Specifications and Bases along 
with other changes to the Technical Specifications discussed above. 
The reformatting, renumbering, and rewording along with the other 
changes listed involves no technical changes to existing Technical 
Specifications, and does not impact initiators of analyzed events. 
It also does not impact the assumed mitigation of accidents or 
transient events. Therefore, the change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change relocates requirements to other sections of 
the Technical Specifications, to plant procedures, or to the 
Technical Specifications BASES. The procedure change and BASES 
change processes require any changes that reflect plant design as 
described in the FSAR [Final Safety Analysis Report] be evaluated in 
accordance with 10 CFR 50.59. Since any changes will be evaluated 
per 10 CFR 50.59, no increase (significant or insignificant) in the 
probability or consequences of an accident previously evaluated will 
be allowed. Therefore, this change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 34886]]

    The proposed change provides more stringent requirements than 
previously existed in the Technical Specifications. The more 
stringent requirements will not result in operation that will 
increase the probability of initiating an analyzed event. If 
anything the new requirements may decrease the probability or 
consequences of an analyzed event by incorporating the more 
restrictive changes discussed above. The change will not alter 
assumptions relative to mitigation of an accident or transient 
event. The more restrictive requirements will not alter the 
operation of process variables, structures, systems, or components 
as described in the safety analyses. Therefore, the change will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed change deletes the requirements for Standby Liquid 
Control (SLC) System operability during Hot Shutdown, Cold Shutdown, 
and Refueling. The SLC System is not assumed in the initiation of 
any previously evaluated events and therefore the proposed change 
will not increase the probability or consequence of a previously 
analyzed accident. The SLC System is not assumed to operate in the 
mitigation of any previously analyzed accidents which are assumed to 
occur during Hot Shutdown, Cold Shutdown or Refueling. This change 
will not result in operation that will increase the probability of 
initiating an analyzed event. This change will not alter assumptions 
relative to mitigation of an accident or alter the operation of 
process variables, structures, systems, or components as described 
in the safety analyses. Therefore, this change will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change adds an action for both SLC subsystems 
inoperable that delays the requirement to initiate plant shutdown 
immediately and allows time to recover at least one subsystem before 
subjecting the plant to a potentially unnecessary transient. 
Allowing a short period of time to recover one subsystem is 
acceptable because of the large number of independent control rods 
available to shut down the reactor and the diversity of means 
available to cause control rod insertion. This change will not alter 
assumptions relative to mitigation of an accident or alter the 
operation of process variables, structures, systems, or components 
as described in the safety analyses. Therefore, this change will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed change deletes requirements for demonstrating 
operability of the redundant subsystems which eliminates excessive 
and unnecessary testing of safety significant equipment. This is 
consistent with guidance 10.1 of Generic Letter 93-05, ``Line-Item 
Technical Specifications Improvements to Reduce Surveillance 
Requirement for Testing During Power Operations''. The change does 
not affect the ability of the SLC system to perform on demand, and 
by actually lowering the number of demands to demonstrate 
operability, reduces the probability of equipment failure. Since the 
change will not alter assumptions relative to mitigation of an 
accident or alter the operation of process variables, structures, 
systems, or components as described in the safety analyses, the 
change will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change replaces the requirement to verify B-10 
enrichment concentration by test anytime boron is added to the 
solution and each refueling outage with verifying the enrichment 
prior to addition. Since enrichment of the solution in the tank 
cannot change by any other means but chemical addition, ensuring 
that only properly enriched material is available for addition is 
adequate to maintain enrichment at the required level. This change 
will not alter assumptions relative to mitigation of an accident or 
alter the operation of process variables, structures, systems, or 
components as described in the safety analyses. Therefore, this 
change will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1
    The changes to Note 7 to Table 3.1.1, and the addition of Note 6 
to Table 3.2.C.1 are proposed to clarify their requirements, the 
appropriate action to take, and their relationship to plant modes. 
This revised scram and rod block applicability is acceptable because 
control rods withdrawn from a core cell containing no fuel 
assemblies have a negligible impact on the reactivity of the core, 
and, therefore, are not required to be operable. Provided all rods 
otherwise remain inserted, the RPS functions serve no purpose and 
are not required. In this condition, the required shutdown margin 
(Specification 3.3.A.1) and the required one-rod-out interlock 
(Specification 3.10.A) ensure that no event requiring the RPS or Rod 
Block will occur.
    The Actions of Table 3.1.1 for inoperable equipment were 
previously revised in Amendment <greek-i>147 to be consistent with 
the improved STS. Action (A) requires fully inserting all insertable 
control rods in core cells containing one or more fuel assemblies. 
Since Specification 3.1O.A requires all control rods to be fully 
inserted during fuel movement, the proposed applicable conditions 
cannot be entered while moving fuel. In addition, Specification 
3.10.D, used for controlling multiple control rod removal, requires 
all control rods in a 3X3 array centered on the CRDs being removed 
to be fully inserted and electrically disarmed and all other control 
rods fully inserted. The only possible action is control rod 
withdrawal, which is addressed by Action A. Hence, operating Pilgrim 
in accordance with the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Section 3/4.4
    The proposed change involves reformatting, renumbering, and 
rewording of the existing Technical Specifications and Bases along 
with other changes to the Technical Specifications discussed above. 
The reformatting, renumbering, and rewording along with the other 
changes listed involves no technical changes to existing Technical 
Specifications. These changes are administrative and do not impact 
the assumed mitigation of accidents or transient events. Therefore, 
these changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change relocates requirements to other Technical 
Specification sections, to plant procedures, or to the Technical 
Specification BASES. Relocating requirements will not alter the 
plant configuration (no new or different type of equipment will be 
installed) or changes in methods governing normal plant operation. 
Relocating requirements will not impose different requirements and 
adequate control of information will be maintained. Relocating 
requirements will not alter assumptions made in the safety analysis 
and licensing basis. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes make some existing requirements more 
restrictive and add additional requirements to the Technical 
Specifications but will not alter the plant configuration (no new or 
different type of equipment will be installed) or change methods 
governing normal plant operation. These changes do impose different 
requirements, however, they are consistent with assumptions made in 
the safety analyses. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change relaxes the modes of applicability for the 
SLC. Relaxing the applicability will not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed) or changes in methods governing normal plant 
operation. Therefore, this change will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.
    Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1
    This revised scram and rod block applicability is acceptable 
because control rods withdrawn from a core cell containing no fuel 
assemblies have a negligible impact on the reactivity of the core, 
and, therefore, are not required to be operable (provide a scram). 
Provided all rods otherwise remain inserted, the RPS functions serve 
no purpose and are not required. In this condition, the required 
shutdown margin (Specification 3.3.A.1) and the required one-rod-out 
interlock (Specification 3.10.A) ensure that no event requiring the 
RPS or Rod Block will occur.
    The Actions of Table 3.1.1 for inoperable equipment were 
previously revised in

[[Page 34887]]

Amendment <greek-i>147 to be consistent with the improved STS. 
Action (A) requires fully inserting all insertable control rods in 
core cells containing one or more fuel assemblies. Since 
Specification 3.10.A requires all control rods to be fully inserted 
during fuel movement, the proposed applicable conditions cannot be 
entered while moving fuel. In addition, Specification 3.10.D, used 
for controlling multiple control rod removal, requires all control 
rods in a 3X3 array centered on the CRDs being removed to be fully 
inserted and electrically disarmed and all other control rods fully 
inserted. The only possible action is control rod withdrawal, which 
is adequately addressed by Action A.
    Therefore, operating Pilgrim in accordance with the proposed 
changes will not involve a significant reduction in a margin of 
safety.
    Section 3/4.4
    The administrative changes involve no technical changes. These 
proposed changes will not reduce a margin of safety because there is 
no impact on any safety analysis assumptions. Also, because the 
change is administrative in nature, no question of safety is 
involved. Therefore, these changes do not involve a significant 
reduction in a margin of safety. The change relocates requirements 
to other Technical Specification sections, to plant procedures, or 
to the Technical Specification BASES. These changes will not reduce 
a margin of safety since there is no impact on any safety analysis 
assumptions. In addition, the requirements to be transposed are the 
same as the existing Technical Specifications. Since any changes to 
plant procedures and Technical Specification BASES are required to 
be evaluated per 10 CFR 50.59, no reduction (significant or 
insignificant) in a margin of safety will be allowed. Therefore, 
these changes will not involve a significant reduction in a margin 
of safety.
    The addition of new requirements and making existing ones more 
restrictive either increases or does not affect the margin of 
safety. These changes do not impact any safety analysis assumptions. 
As such, no question of safety is involved. Therefore, these changes 
will not involve a significant reduction in a margin of safety.
    The proposed change would remove a backup (in the Hot Shutdown, 
Cold Shutdown, and Refueling Modes) to the available systems for 
reactivity control; however, this backup is not considered in the 
margin of safety when determining the required reactivity for 
shutdown and refueling events. This change will have no impact on 
any safety analysis assumptions. As such, no question of safety is 
involved. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    The SLC system is not assumed to function in any DBA or 
transient and is not the primary success path of a safety sequence 
analysis. It is a backup to the CRD scram function, therefore, 
allowing a short period of time to recover one subsystem will have 
no impact on any safety analysis assumptions. As such, no question 
of safety is involved. Therefore, this change does not involve a 
significant reduction in a margin of safety.
    The change does not alter the requirements for enrichment/ 
concentration of the boron solution necessary to satisfy 10 CFR 
50.62. Since enrichment of the solution in the tank cannot change by 
any other means but chemical addition, this change does not involve 
a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199
    NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

    Date of amendments request: November 15, 1995
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications (TS) to alter the wording of TS 
4.8.2.5.a in accordance with the guidance of Generic Letter (GL) 91-09, 
``Modification of Surveillance Interval For The Electrical Protection 
Assemblies In Power Supplies For The Reactor Protection System.''
    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 amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the proposed change does not alter the design, 
function, or operation of the EPAs [Electrical Protective 
Assemblies]. The proposed amendments modify the surveillance 
requirement for an electrical protective device on the Reactor 
Protection System [RPS]. The RPS-EPA units are designed to protect 
RPS equipment from abnormal operating voltage or frequency. The 
proposed change will preclude the need to test the RPS-EPA units 
during power operation. This will eliminate the potential for 
reactor scrams and Group isolations during performance of the 
surveillance, thus, preventing unwarranted challenges to safety 
systems. The proposed change does not affect any accident precursor 
or initiator. Therefore, the probability of an accident is not 
affected by the proposed change. The proposed amendments do not 
affect the operability of the RPS-EPA units. The proposed change 
does not affect the ability of the Reactor Protection System to 
maintain the integrity of the fuel cladding, protect the reactor 
coolant pressure boundary, or limit the amount of energy released to 
primary containment. Therefore, the consequences of an accident is 
not affected by the proposed change.
    2. The proposed amendments do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated above, these proposed amendments do not alter 
the design, functions, or operation of the EPAs. The RPS relay trip 
logic remains protected from power supplies operating with abnormal 
voltage or frequency. Additionally, the redundancy of this 
protection is not changed.
    Thus, the proposed amendments do not create the possibility of a 
new or different kind of accident.
    3. The proposed amendments do not involve a significant 
reduction in a margin of safety because the benefit to safety by 
reducing the frequency of testing during power operation and 
attendant possible challenges to safety systems more than offsets 
any risk to safety from relaxing the surveillance requirement to 
test the EPAs during power operation. The testing of each EPA 
channel involves a dead-bus transfer and the momentary interruption 
of power results in a half scram and half isolation. Generic Letter 
91-09 notes that many plants have encountered problems with the 
reset of the half trip resulting in inadvertent scrams and group 
isolations that challenge safety systems during power operation. 
Eliminating EPA testing at power operation increases the margin of 
safety by eliminating the potential for trips due to testing that 
challenge safety systems. An insignificant reduction in the margin 
of safety is introduced by increasing the test interval up to a 
maximum of a refuel cycle which will produce a small increase in 
risk that an inoperable EPA would not be detected. The elimination 
of potential challenges to safety systems provides a safety benefit 
that offsets the increased risks of component failure.
    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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Eugene V. Imbro

[[Page 34888]]

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: June 6, 1996
    Description of amendment request: The proposed change would revise 
technical specifications (TS) Section 4.2.3 to allow the licensee to 
defer the ultrasonic inspection of the reactor coolant pump flywheel 
for one operating cycle.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The safety function of the Reactor Coolant Pump (RCP) flywheel 
is to provide a coastdown period during which the RCPs would 
continue to provide reactor coolant flow to the core after a loss of 
power to the RCPs. The maximum loading on the RCP motor flywheel 
results from overspeed following a large break Loss of Coolant 
Accident (LOCA). The estimated maximum obtainable speed in the event 
of a Reactor Coolant System (RCS) piping break was established 
conservatively, and the proposed one-time change does not affect 
that analysis.
    The RCP flywheels have been carefully designed and manufactured 
from high quality steel. Twenty-two inspections have been performed 
at HBRSEP, Unit No. 2 over the past 25 years and no indications have 
been discovered that would affect the integrity of the flywheel. The 
Westinghouse Owners Group (WOG) has performed an extensive study 
documented in WCAP-14535, ``Topical Report on Reactor Coolant Pump 
Flywheel Inspection Elimination,'' that includes an evaluation of 
industry experience, a stress and fracture evaluation, and a risk 
assessment, and has concluded that RCP flywheel inspections may be 
safely eliminated.
    Reduced coastdown times due to a single failed flywheel would 
not place the plant in an unanalyzed condition since a locked rotor 
(i.e., an instantaneous coastdown) is analyzed in the Updated Final 
Safety Analysis Report (UFSAR). The proposed change also does not 
increase the amount of radioactive material available for release or 
modify any systems used for mitigation of releases during an 
accident. Therefore, the proposed change does not involve an 
increase in the probability of consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change will not change the design, configuration, 
or method of operation of the plant. Therefore, the proposed change 
will not create the possibility of a new kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The RCP flywheels have been carefully designed and manufactured 
from high quality steel. Twenty-two inspections have been performed 
at HBRSEP, Unit No. 2 over the past 25 years and no indications have 
been discovered that would affect the integrity of the flywheel. The 
Westinghouse Owners Group (WOG) has performed an extensive study 
documented in WCAP-14535, ``Topical Report on Reactor Coolant Pump 
Flywheel Inspection Elimination,'' that includes an evaluation of 
industry experience, a stress and fracture evaluation, and a risk 
assessment, and has concluded that RCP flywheel inspections may be 
safety eliminated. The proposed change would only result in a one-
time deferral of the scheduled inspection for one operating cycle. 
In consideration of the historical integrity of the HBRSEP, Unit No. 
2 RCP flywheels, the industry experience, the results of the WOG 
study, and the deferral of the risk of RCP flywheel damage during 
disassembly and inspection, we conclude that a one operating cycle 
deferral of the scheduled RCP flywheel inspection will not result in 
a 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: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Eugene V. Imbro

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: May 31, 1996
    Description of amendment request: The proposed amendment would 
change the plant Technical Specifications (TS) Table 3.3-7, Seismic 
Monitoring Instrumentation, and TS Table 4.3-4, Seismic Monitoring 
Instrumentation Surveillance Requirements, to correct the location 
described for one of the three Triaxial Peak Accelerograph Recorders.
    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.
    These recorders are passive components which serve only a 
recording function. They can neither initiate an accident nor serve 
to mitigate accident consequences. The proposed change serves only 
to correct the location, commensurate with design documents, for one 
of the three recorders described in the Technical Specifications. 
Accordingly, this change is administrative in nature. 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 correction is an administrative change to correct 
the location of a recorder currently described in the Technical 
Specifications. No physical alterations to plant equipment are being 
made, and there will be no changes that alter how 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.
    Technical Specification Bases 3/4.3.3.3 specify the acceptance 
level for seismic instrumentation as ``consistency'' with the 
recommendations of Regulatory Guide 1.12. Since the regulatory guide 
states only that one recorder should be provided at a ``selected 
location on the reactor piping,'' it is not material whether it is 
installed on Loop 1 versus Loop 2. Therefore, the proposed change 
does not affect a margin of safety as defined in the Bases to the 
Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Eugene V. Imbro

[[Page 34889]]

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

    Dates of amendment request: December 18, 1995, May 3 and June 11, 
1996
    Description of amendment request: The licensee proposed to change 
the Turkey Point Units 3 and 4 Technical Specifications (TS) to uprate 
the core thermal output of Turkey Point Units 3 and 4 from 2200 MWt to 
2300 MWt. The proposed TS changes were divided into eight groups. The 
submittal included a ``No Significant Hazards'' evaluation for each of 
the eight groups. The groupings are as follows:
    TS changes associated with the uprated power level, the revised 
core safety limits, revised DNB [departure from nucleate boiling] 
parameters, Engineered Safety Features Actuation System (ESFAS) and 
reactor trip setpoint changes, and Reactor Coolant Pump (RCP) Breaker 
Position Trip, were evaluated together. The safety of these proposed 
changes were verified by the accident analyses that were completed in 
support of the uprated power.
    TS changes associated with reducing the SI [safety injection] pump 
discharge head requirement and increasing usable volume requirements 
for the Demineralized Water Storage Tank (DWST) and the Condensate 
Storage Tank (CST) were addressed together.
    TS changes associated with pressurizer and main steam safety valve 
(MSSV) setpoint tolerance increases were assessed together.
    TS changes associated with operation at reduced power with 
inoperable MSSVs were assessed separately.
    TS changes associated with the service period for heatup and 
cooldown pressure-temperature limit curves were assessed together.
    The Surveillance Requirement change for the emergency containment 
cooling [ECC] unit operability was handled separately since this was a 
design change that required extensive evaluations.
    TS change associated with the methyl iodide removal efficiency in 
the Control Room Emergency Ventilation System was assessed separately.
    All LOCA [loss-of-coolant accident] related changes dealing with 
the peaking factor increase, COLR [core operating limit report] 
changes, Evaluation Model references, and relocation of peaking factors 
from the TS and subsequent inclusion in the COLR were included in one 
``No Significant Hazards'' evaluation. All of the items are closely 
related since the LOCA analysis is performed to ensure peaking factor 
acceptability.
    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.
    LICENSE CONDITION, RATED THERMAL POWER, CORE SAFETY LIMITS, 
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS, ESFAS 
INSTRUMENTATION TRIP SETPOINTS, DNB PARAMETERS AND RCP BREAKER 
POSITION TRIP
    (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 an increase in the 
probability or consequences of an accident previously evaluated 
because operation with these revised values will not cause any 
design or analysis acceptance criteria to be exceeded. The 
structural and functional integrity of all plant systems are 
unaffected. The overtemperature Delta T and overpower Delta T 
reactor trip functions as well as ESFAS functions are part of the 
accident mitigation response and are not accident initiators. All 
proposed changes have been assessed and no design and analysis 
acceptance criteria have been exceeded. Therefore the probability of 
occurrence previously evaluated is not affected.
    The proposed changes do not affect the integrity of the fission 
product barriers utilized for mitigation of dose consequences as a 
result of an accident. Dose consequences were reviewed and 
reanalyzed (as needed) and found acceptable. Therefore, the 
probability or consequences of an accident previously evaluated are 
not significantly increased.
    (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 
because their effects do not affect accident initiation sequences. 
All new operating configurations have been evaluated and no new 
limiting single failures have been identified. In addition, no new 
failure modes have been identified. Therefore, it is concluded that 
no new or different kind of accident from any accident previously 
evaluated has been created as a result of these revisions.
    (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 do not involve a reduction in a margin of 
safety because the margin of safety associated with these parameters 
as verified by the results of the accident analyses, are within 
acceptable limits. All transients impacted have been analyzed and 
have met the applicable accident analyses acceptance criteria (e.g., 
DNBR [departure from nucleate boiling ratio], RCS [reactor coolant 
system] pressure, secondary side pressure, etc.). The margin of 
safety required for each affected safety analysis is maintained. The 
adequacy of the revised Technical Specifications values has been 
confirmed such that there is no reduction in the margin of safety. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    AVAILABLE VOLUME CHANGE FOR CONDENSATE STORAGE TANK (CST) AND 
DEMINERALIZED WATER STORAGE TANK (DWST), AND REDUCED SAFETY 
INJECTION (SI) PUMP DISCHARGE HEAD REQUIREMENT.
    (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 revised tank volumes and SI head requirements have been 
evaluated with respect to system performance and analysis impacts. 
All accident analysis acceptance criteria continue to be met. The 
design function of all affected systems have been reviewed and all 
system design criteria continue to be met. The structural and 
functional integrity of the affected systems are unaffected. These 
changes are not initiators for any accident and therefore the 
probability of occurrence of an accident previously evaluated has 
not increased.
    The proposed changes do not affect the integrity of the fission 
product barriers for mitigation of dose consequences. All dose 
consequences remain well within the 10 CFR 100 limits. Therefore 
there is no 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 revised tank volumes and SI head requirements do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated because these modifications do not 
affect accident initiation sequences. No new operating configuration 
is being imposed by the adjustments that would create a new failure 
scenario. In addition, no new failure modes or limiting single 
failures have been identified. Therefore, it is concluded that no 
new or different kind of accident from any accident previously 
evaluated have been created as a result of these revisions.
    (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 do not involve a reduction in a margin of 
safety because the margin of safety associated with these 
parameters, as verified by the results of the accident analyses and 
system evaluations, are within acceptance limits. The margin of 
safety required for each affected safety analysis is maintained. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    PRESSURIZER AND MAIN STEAM SAFETY VALVE SETPOINT TOLERANCES

[[Page 34890]]

    (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 revised tolerances for main steam safety valves and 
pressurizer safety valves do not involve an increase in the 
probability or consequences of an accident previously evaluated 
because operation with these revised values will not cause any 
design or analytical acceptance criteria, such as those applicable 
to primary and secondary side pressures to be exceeded. The 
structural and functional integrity of the valves are unaffected by 
this proposed change. The tolerance changes do not initiate or cause 
initiation of any transient. Therefore, the probability of 
occurrence previously evaluated is not affected.
    The changes do not affect the integrity of the fission product 
barriers utilized for dose consequence mitigation. Therefore, the 
probability or consequences of an accident previously evaluated is 
not increased.
    (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 revised valve tolerances do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated because the tolerances do not affect accident initiation 
sequences. No new operating configuration is being imposed by the 
tolerances that would create a new failure scenario. In addition, no 
new failure modes or limiting single failures have been identified. 
Therefore, it is concluded that no new or different kind of accident 
from any accident previously evaluated have been created as a result 
of these revisions.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The changes to valve tolerances do not involve a reduction in a 
margin of safety because the margin of safety associated with the 
MSSVs and the pressurizer safety valves, as verified by the results 
of the accident analyses and valve evaluations, are within 
acceptable limits. Transients impacted by this change have been 
analyzed and have met the applicable accident analyses acceptance 
criteria, such as those applicable to primary and secondary side 
pressure. The margin of safety required for each affected safety 
analysis is maintained. This conclusion is not changed by the valve 
tolerances for the main steam safety valves and the pressurizer 
safety valves. Therefore, the changes do not involve a significant 
reduction in the margin of safety.
    OPERATION AT REDUCED POWER WITH INOPERABLE MAIN STEAM SAFETY 
VALVES
    (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 not exceed 110 percent of the 
design pressure 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.
    (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 as specified in TS Table 3.7-1 will improve the 
capability of the MSSVs to perform their pressure relief function to 
ensure the secondary side steam pressure does not exceed 110 percent 
of design pressure following a Loss of Load/Turbine Trip event. 
Therefore, since the function of the MSSVs is improved by the 
proposed changes, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    (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 that 110 percent of secondary side steam 
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.
    SERVICE PERIOD FOR HEATUP AND COOLDOWN PRESSURE-TEMPERATURE 
LIMIT CURVES
    (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.
    Calculation of the service period for the heatup and cooldown 
curves does not involve an increase in the probability or 
consequences of an accident previously evaluated because the 
calculations were completed to verify the adequacy of the existing 
curves and to determine an appropriate service period. The use of 
approved methods and the acceptable results have shown that no 
design or analysis criteria are changed. The structural and 
functional integrity of the reactor vessel has been verified.
    No fission product barriers or inputs to dose analyses are 
adversely affected by these calculations and reverification of the 
existing heatup/cooldown curves. Therefore, the probability or 
consequences of an accident previously evaluated are not increased.
    (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 revised service period does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated because the recalculation of an acceptable service period 
does not affect accident initiation sequences. No new operating 
configuration is being imposed by the calculations that would create 
a new failure scenario. In addition, no new failure modes or 
limiting single failures have been identified. Therefore, the types 
of accidents defined in the UFSAR continue to represent the credible 
spectrum of events to be analyzed which determine safe plant 
operation. Therefore, it is concluded that no new or different kind 
of accident from any accident previously evaluated have been created 
as a result of these revisions.
    (3) Operation of the facility in accordance with the proposed 
license amendments would not involve a significant reduction in a 
margin of safety.
    Calculations were performed to determine the service period 
appropriate for the existing curves. The changes to service period 
do not involve a reduction in a margin of safety because the margin 
of safety associated with the heatup/cooldown curves, as verified by 
the results of the analyses, are unchanged. Therefore, the proposed 
change to the service period does not involve a significant 
reduction in the margin of safety.
    MODIFICATION TO SURVEILLANCE REQUIREMENT FOR EMERGENCY 
CONTAINMENT COOLING SYSTEM
    (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 purpose of the ECC units is to help mitigate the 
consequences of an accident (i.e., to help maintain the containment 
pressure and temperature within their design

[[Page 34891]]

values following a design basis accident). The ECC units do not 
operate during normal operation of the plant. Failure of the ECC 
units would not initiate a plant transient or accident. Therefore, 
the proposed change involving the ECC units would not affect the 
probability of occurrence of an accident previously evaluated.
    Evaluations demonstrate that, with two ECC units operating 
during a LOCA or MSLB [main steamline break], the containment 
pressure and temperature will be maintained within their design 
values. These evaluations also demonstrate that, with two ECC units 
operating during a LOCA or MSLB, the temperature of the CCWS 
[component cooling water system] will be maintained within its 
design temperature. Therefore, the proposed change involving the ECC 
units would not affect the 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 purpose of the ECC units is to mitigate design basis 
accidents, and failure of the ECC units would not cause a plant 
transient or accident. Furthermore, a single failure of an ECC unit 
during a LOCA or MSLB would not lead to a new or different kind of 
accident. Although the revised Technical Specifications require two 
ECC units to start automatically on a LOCA signal, they would also 
require that all three ECC units be operable. On a single failure of 
an operating ECC unit, there would be sufficient time to start the 
standby ECC unit to accomplish the design function of the ECC 
system. Therefore, the proposed amendment 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 change in the actuation logic of the ECC units 
would not cause either the containment pressure and temperature or 
the CCWS temperature to exceed their design values. While the energy 
released into containment and subsequently transferred to the CCWS 
will increase as a result of the thermal uprate, this increase is 
insignificant and will not result in either the containment or CCWS 
exceeding a design limit. Therefore, the proposed change would not 
affect the margin of safety.
    CONTROL ROOM EMERGENCY VENTILATION SYSTEM
    (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 change does not affect the integrity of the fission 
product barriers utilized for mitigation of dose consequences as a 
result of an accident. Only the iodide removal efficiency of the 
control room emergency ventilation system is increased, and this 
change is in the conservative direction.
    To assure consistency between testing efficiency and analysis 
assumptions for post-accident control room doses, the methyl iodide 
removal efficiency required to be demonstrated by laboratory test, 
is being increased from 90% to 99%. This increase in testing 
efficiency is consistent with the recommendations set by the NRC 
staff in Regulatory Guide 1.52 to support analysis efficiencies for 
elemental iodine and methyl iodide removal of 95%, respectively. 
Testing performed to verify methyl iodide removal efficiency will be 
performed under conditions representative of the control room 
environment.
    Since this change in removal efficiency is in the conservative 
direction, plant safety will not be adversely impacted.
    (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 change to the control room emergency ventilation 
system iodide removal efficiency does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated because operation of the control room emergency 
ventilation system is not identified in any accident initiation 
sequence. The system is provided to minimize operator exposure to 
airborne radioactivity released as a result of an accident. The new 
operating configuration has been evaluated and no new limiting 
single failures have been identified as a result of the proposed 
modification. Therefore, it is concluded that no new or different 
kind of accidents from any accident previously evaluated have been 
created as a result of these revisions.
    (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 do not involve a reduction in the margin of 
safety because the margin of safety associated with this change is 
in the conservative direction. Thus, plant safety will not be 
adversely impacted and the margin of safety required for the 
affected safety analysis is maintained. The adequacy of the revised 
Technical Specification values to maintain the plant in a safe 
operating condition has been confirmed, since the testing will be 
done to a more conservative criteria (i.e., 99% efficiency). 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    RELOCATION OF F<INF>Q(Z) [HEAT FLUX HOT CHANNEL FACTOR] AND F 
Delta H [NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR] LIMITS FROM 
TECHNICAL SPECIFICATIONS TO CORE OPERATING LIMITS REPORT AND 
EDITORIAL CORRECTIONS
    (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 relocation of the values for F<INF>Q and F Delta H from the 
Technical Specifications to the Core Operating Limits Report is 
administrative in nature and has no impact on the probability or 
consequences of any Design Bases Event (DBE) occurrence which was 
previously evaluated. The determination of the F<INF>Q and F Delta H 
limits will be performed using methodology approved by the NRC and 
poses no significant increase in the probability or consequences of 
any accident previously evaluated.
    The changes being proposed as editorial in nature do not affect 
assumptions contained in the safety analyses, the physical design 
and/or operation of the plant, nor do they affect Technical 
Specifications that preserve safety analysis assumptions. Therefore, 
these proposed changes do not affect the probability or consequences 
of accidents previously analyzed.
    (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 relocation of the F<INF>Q and F Delta H limits from the 
Technical Specifications to the Core Operating Limits Report is 
administrative in nature and has no impact, nor does it contribute 
in any way to the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The determination of the F<INF>Q and F Delta H limits will be 
performed using NRC-approved methodology and are submitted to the 
NRC as a revision to the COLR to allow the NRC staff to trend 
peaking factors. The Technical Specifications will continue to 
require operation within the required core operating limits and 
appropriate actions will be taken if the F<INF>Q and F Delta H 
limits are exceeded. Therefore, the proposed amendments does not in 
any way create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The editorial changes proposed are administrative in nature and 
do not affect assumptions contained in plant safety analyses, the 
physical design and/or operation of the facility, nor do they affect 
Technical Specifications that preserve safety analysis assumptions. 
Therefore, these changes do not create the possibility of a new or 
different kind of accident.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The relocation of the F<INF>Q and F Delta H limits from the 
Technical Specifications to the Core Operating Limits Report is 
administrative in nature and has no impact on the margin of safety. 
The determination of the F<INF>Q and F Delta H limits will be 
performed using methodology approved by the NRC and does not 
constitute a significant reduction in the margin of safety.
    The supporting Technical Specification values are defined by the 
accident analyses which are performed to conservatively bound the 
operating conditions defined by the Technical Specifications. 
Performance of analysis and evaluation have confirmed that the 
operating envelope defined by the Technical Specifications continues 
to be bounded by the analytical basis, which in no case exceeds the 
acceptance limits. Therefore, the margin of safety provided in the 
analyses in accordance with the acceptance limits is maintained and 
not significantly reduced.

[[Page 34892]]

    The changes being proposed as editorial in nature do not relate 
to or modify the safety margins defined in, and maintained by the 
Technical Specifications. Therefore, the proposed changes which 
correct administrative errors and clarify existing Technical 
Specification requirements do not involve any 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: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: Frederick J. Hebdon

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

    Dates of amendment request: April 19, 1996, May 10, 1996, and May 
28, 1996
    Description of amendment request: The licensee proposed to change 
the Turkey Point Units 3 and 4 Technical Specifications (TS) to address 
frequency extension for actions required on a periodic basis, delete 
the separate notification requirement for an inoperable startup 
transformer, and allow the operating RHR loop to be removed from 
operation during refueling operations under certain conditions.
    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 amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed amendments are purely administrative in nature. 
These amendments will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because they do not affect assumptions contained in plant safety 
analyses, the physical design and/or operation of the plant, nor do 
they affect Technical Specifications that preserve safety analysis 
assumptions. Therefore, the proposed changes do not affect the 
probability or consequences of accidents previously analyzed.
    (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 use of the modified specifications can not create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
the administrative changes and clarifications, since the proposed 
changes do not involve the addition or modification of equipment nor 
do they alter the design or operation of affected plant systems, 
structures, or components.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are unchanged by the proposed 
amendments. The modified specifications which correct administrative 
errors and clarify existing Technical Specification requirements do 
not significantly reduce any of 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 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: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: Frederick J. Hebdon

Gulf States Utilities Company, Cajun Electric Power Cooperative, 
and Entergy Operations, Inc., Docket No. 50-458, River Bend 
Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: May 30, 1996
    Description of amendment request: The proposed amendment would 
revise the technical specifications surveillance requirement (SR) 
3.8.3.4 to specify a 5-start pressure for the air receivers associated 
with the Division III, High Pressure Core Spray emergency 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. Does the change involve a significant increase in the 
probability or consequence of an accident previously evaluated?
    The purpose of the proposed Technical Specification change is to 
establish consistency between the basis for the air start pressure 
required for the Division I and II diesels and the value required 
for the Division III diesel. The value of 160 psig currently 
specified in SR 3.8.3.4 is representative of a 5-start value for the 
Division I and II diesels, however, this value is not representative 
of a 5-start for the Division III diesel. While the 160 psig value 
does serve to satisfy the requirements of 10 CFR 50.36 with regard 
to maintaining the lowest functional level required for the Division 
III diesel to perform its design safety function, the current value 
does not serve to maintain the design margin utilized when sizing 
the air receivers for the purpose of satisfying the Standard Review 
Plan guidance contained in section 9.5.6 (NUREG-0800 Revision 2).
    The proposed value fully complies with the guidance provided in 
NUREG-0800 and is more conservative than the value currently 
included in the Technical Specifications. The proposed value is well 
within the capability of the air system's design and will not 
subject the air system to excessive pressures or undue cycling of 
the system's compressors. The proposed change has no effect on the 
probability of an accident as diesel generators have no bearing on 
the initiation of any analyzed event. In addition, the capability of 
the Division III diesel to perform its design basis function (i.e., 
starting, accelerating to rated speed and voltage, and connecting to 
its respective bus within 13 seconds) is not affected by this 
change. The ability of the diesel to support the mitigation of 
analyzed accidents is not affected and hence the consequences of any 
analyzed event are not affected. Therefore, the proposed change does 
not increase the probability or the consequences of previously 
analyzed accidents.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not introduce any new failure modes. 
All of the affected components remain within their applicable design 
limits. In addition, the environmental qualification of any plant 
equipment is not adversely affected by the proposed change. Since 
the performance of this system is not adversely affected by this 
change and the design margins of this system are not challenged in a 
manner differently than previously analyzed, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change raises the required starting air pressure 
for the Division III above that currently required by the Technical 
Specifications to establish consistency between the basis of the 
Division III value with the value used for the Division I and II 
diesels. Issuance of the proposed change will establish a 5 start 
air receiver pressure for each of the three safety-related diesels 
at

[[Page 34893]]

River Bend. While the proposed value is slightly less than the 5 
start value discussed in River Bend's SER, the proposed value is 
supported by the River Bend site-specific test data and does not 
adversely affect existing analyses or system performance. Therefore, 
the proposed change does not result in a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005
    NRC Project Director: William D. Beckner

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 6, 1996, as supplemented by letters 
dated June 7 and 9, 1996
    Description of amendment request: The proposed amendment would 
revise the technical specification Limited Safety System Setting for 
the MINIMUM CRITICAL POWER RATIO (MCPR) for dual recirculation loop 
operation and for single recirculation loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The purpose of the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) is to provide statistical confidence that less than 0.1% of 
the fuel rods in a core would experience transition boiling during 
the most limiting analyzed Anticipated Operational Occurrence 
(transient). While transition boiling in a BWR does not in and of 
itself signal the onset of fuel cladding failure, this criterion has 
been selected as a conservative and convenient parameter for the 
evaluation of fuel designs. Therefore, while this safety limit does 
not provide any control over either the probability or consequences 
of any accident previously evaluated, it does ensure that evaluated 
transients remain within NRC-approved criteria. Revision of the 
SLMCPR will establish in the CNS Technical Specifications a valid 
limit, based on the NRC approved GESTAR II methodology using cycle-
specific inputs. This change will result in the input of more 
restrictive core operating limits into the plant process computer, 
ensuring that CNS will be operated within the constraints of the new 
SLMCPR limits of 1.07 for dual recirculation loop operation, and 
1.08 for single recirculation loop operation. No plant hardware 
modifications are associated with this change. Therefore, since this 
proposed change will not change the physical configuration of the 
plant, nor result in operational changes which invalidate 
assumptions used in any CNS accident analysis, this change does not 
involve an increase in the probability or consequences of any 
accident previously evaluated.
    2. Does the proposed License Amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    This change revises the SLMCPR values in the CNS Technical 
Specifications in accordance with a cycle specific analysis 
performed for the remainder of the current cycle. The SLMCPR ensures 
that less than 0.1% of the fuel rods in a core would experience 
transition boiling during the most limiting Anticipated Operational 
Occurrence. Increasing the SLMCPR from 1.06 to 1.07 for dual 
recirculation loop operation and from 1.07 to 1.08 for single 
recirculation loop operation will ensure that the specified 
statistical confidence will be met for all analyzed transients. This 
change does not involve any plant hardware changes. The only 
operational changes will be the institution of appropriate thermal 
restrictions on reactor core operation in accordance with the SLMCPR 
changes. Therefore, this proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change create a significant reduction in 
the margin of safety?
    This change will establish in the CNS Technical Specifications, 
SLMCPR values that ensure the margin of safety to the NRC approved 
Anticipated Operational Occurrence evaluation acceptance criteria 
will be met. Increasing the SLMCPR institutes more restrictive 
thermal limitations on core operation. The change of the SLMCPR from 
1.06 to 1.07 for dual recirculation loop operation, and from 1.07 to 
1.08 for single loop operation will ensure that the acceptance 
criteria for evaluated transients will continue to be met, and that 
the appropriate limit is reflected in the CNS Technical 
Specifications. Therefore, this proposed change does not create a 
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: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499
    NRC Project Director: William D. Beckner

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

    Date of amendment request: May 15, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.3.2, ``Isolation Actuation 
Instrumentation,'' to establish a range of allowable and trip setpoints 
for high temperature (varying as a function of ambient temperature) in 
the Main Steam Line Tunnel Lead Enclosure Area. Specifically, a new TS 
Figure 3.3.2-1 would be added to provide a curve of allowable 
temperature values and a curve of trip temperature setpoints, both 
plotted over a range of ambient temperatures. The new Figure would be 
referenced by Table 3.3.2-2 at item 1.d.3 (High Temperature Main Steam 
Line Tunnel Lead Enclosure Trip Function) by a new footnote stating:
    The trip setpoint and allowable value for a channel may be 
established based on Figure 3.3.2-1, if:
    a. The actual ambient temperature readings for all operable 
channels in the Lead Enclosure Area are equal to or greater than the 
ambient temperature used as the basis for the setpoint, and
    b. The absence of steam leaks in the Main Steam Line Tunnel Lead 
Enclosure Area is verified by visual inspection prior to increasing 
a channel setpoint, and
    c. A surveillance is implemented in accordance with Note (d) of 
Table 4.3.2.1-1.
    Similarly, TS Surveillance Table 4.3.2.1-1 would be supplemented at 
item 1.d.3 (High Temperature Main Steam Line Tunnel Lead Enclosure) 
with a new footnote stating:
    (d) In addition to the normal shift channel check, if a channel 
setpoint has been established using Figure 3.3.2-1, then once per 
shift, the actual ambient temperature reading for all operable 
channels in the Lead Enclosure Area shall be verified to be equal to 
or greater than the ambient temperature used as the basis for the 
setpoint.
    Basis for proposed no significant hazards consideration 
determination: The main steam tunnel high temperature isolation 
actuation instrumentation is part of the Leak Detection System (LDS). 
It is used to detect leakage early at 25 gallons per minute (gpm) and 
initiate signals to automatically close the Main Steam Isolation Valves 
before a pipe break could occur. The existing temperature setpoints for 
the tunnel lead enclosure are based upon transient analyses for steam 
leaks in the steam tunnel utilizing

[[Page 34894]]

winter temperatures as an initial condition. The licensee finds that a 
change is needed because actual temperatures in the tunnel, especially 
during the summer, are approaching the setpoints when steam leakage is 
not occurring. Under the present conditions, a minor disturbance in the 
turbine building ventilation system could cause an unwarranted 
isolation actuation at full power with resulting Main Steam Isolation 
Valve closure and reactor scram.
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:
    1. The operation of NMP2 [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 LDS instrumentation in the main steam line tunnel isolates 
the Main Steam Isolation Valves upon sensing a steam leak of 25 gpm. 
For an elevated ambient temperature in the Lead Enclosure area, a 
setpoint established using the proposed Figure 3.3.2-1 ensures that 
the Main Steam Isolation Valves continue to receive an isolation 
signal upon sensing a steam leak of 25 gpm. Verifying the absence of 
any steam leak in the area prior to raising any temperature 
instrument setpoint ensures that the ability to sense a 25 gpm leak 
is not compromised by an increased ambient temperature resulting 
from a smaller steam leak. The periodic surveillance to verify the 
actual ambient temperature ensures the continued validity of the 
ambient temperature used for the setpoint basis, and provides 
sufficient advance indication to take appropriate compensatory 
action. Accordingly, this change will not involve a significant 
increase in the consequences of any accident previously evaluated.
    Furthermore, the LDS function provides a mitigation action for a 
postulated main steam line pipe leak which could lead to a pipe 
break. This function does not affect any accident precursors, and 
the proposed change does not affect the function of the LDS system. 
Accordingly, this change will not involve a significant increase in 
the probability of any accident previously evaluated.
    2. The operation of NMP2 in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any previously evaluated.
    The qualification of safety-related equipment in the main steam 
lead enclosure is evaluated using actual temperatures and component 
qualified life is adjusted accordingly. The temperature elements are 
the only safety-related equipment affected by this change, 
therefore, the instrumentation response to previously evaluated 
accidents will not be adversely affected. This change will not 
affect the performance of safety related structures. Accordingly, 
the design capabilities of those structures, systems and components 
affected by the proposed change are not challenged in a manner not 
previously evaluated so as to create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The operation of NMP2 in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety.
    The proposed change provides a range of setpoints and allowable 
values for the Main Steam Line Tunnel Lead Enclosure temperatures. 
The calculation of the allowable values and trip setpoints was 
performed using the same methodologies as previously employed. For 
an elevated ambient temperature in the Lead Enclosure area, a 
setpoint established using the proposed Figure 3.3.2-1 ensures that 
the Main Steam Isolation Valves receive an isolation signal upon 
sensing a steam leak of 25 gpm, resulting in a main steam line 
isolation prior to a pipe break. Therefore, the proposed change 
provides the same level of protection against a main steam line 
break as the existing setpoint values. The proposed setpoints will 
provide increased scram avoidance, and thereby reduce unnecessary 
challenges to the plant shutdown systems. Accordingly, the proposed 
change does not result in 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: Jocelyn A. Mitchell, Acting Director

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

    Date of application for amendments: March 25, 1996
    Description of amendment request: These amendments revise the 
safety limit minimum critcal power ratios (SLMCPRs) to support use of 
GE-13 fuel at Peach Bottom Atomic Power Station.
    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 TS [technical specification] changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The derivation of the revised GE13 SLMCPRs for incorporation 
into the TS, and its use to determine cycle-specific thermal limits, 
have been performed using USNRC [U.S. Nuclear Regulatory 
Commission]-approved methods within the existing fuel licensing 
criteria as discussed in NEDE-32198P, ``GE13 Compliance With 
Amendment 22 of NEDE-24011-P-A (GESTAR II),'' and cannot increase 
the probability or severity of an accident.
    The basis of the SLMCPRs calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid boiling transition if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling and fuel damage in the event of a 
postulated accident. The fuel licensing acceptance criteria for the 
SLMCPRs calculation apply to the GE13 fuel in the same manner that 
they have applied to previous fuel designs. The probability of fuel 
damage is not increased. Therefore, the proposed TS changes do not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2) The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR for the GE13 fuel design is a Technical Specification 
numerical value, designed to ensure that transition boiling does not 
occur in 99.9% of all fuel rods in the core during the limiting 
postulated accident. It cannot create the possibility of any new 
type of accident. The new SLMCPRs are calculated using USNRC-
approved methods and have the same calculational basis as the SLMCPR 
for other GE fuel designs previously used at PBAPS, Units 2 and 3. 
Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident, from any accident previously 
evaluated.
    3) The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS Bases will remain the 
same. The new SLMCPRs are calculated using USNRC-approved methods 
which are in accordance with the current fuel licensing criteria. 
The SLMCPRs for the GE13 fuel remain high enough to ensure that 
greater than 99.9% of all fuel rods in the core will avoid boiling 
transition if the limit is not violated, thereby preserving the fuel 
cladding integrity. Therefore, the proposed TS changes do 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: Government Publications

[[Page 34895]]

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

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit 
No. 2, York County, Pennsylvania

    Date of application for amendment: June 13, 1996
    Description of amendment request: The proposed amendment to the 
Technical Specifications (TS) will permit a one time performance of 
Surveillance Requirement 3.3.1.1.12, for the Average Power Range 
Monitor Flow Biased High Scram function, with a delayed entry into its 
associated TS Conditions and Required Actions for up to 6 hours 
provided core flow is maintained at or above 82 percent. This change 
would be in effect until the end of refueling outage 2R11, currently 
scheduled for early October 1996.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    i) The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The APRM system provides monitoring and accident mitigation 
functions to limit peak flux in the core during startup and run 
modes. This proposed TS change for delaying entry into Conditions 
and Required Actions associated with SR 3.3.1.1.12 for the APRM flow 
bias function will have no impact on the APRM system or any system 
that interfaces with it. No pressure boundary interfaces or process 
control parameters will be challenged.
    This change does not affect the operation of any equipment. 
Delaying entry into Conditions and Required Actions associated with 
SR 3.3.1.1.12 does not affect either the initiator of any accident 
previously evaluated or any equipment required to mitigate the 
consequences of an accident, or the isotopic inventory in the fuel. 
Thus, the change does not increase either the probability or the 
consequences of accidents previously evaluated.
    ii) The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Because there is no direct pressure boundary interface or 
process control function associated with the APRM system or its 
interfacing electronics, the possibility of a new or different type 
of accident than any previously evaluated will not be created. 
Although the flow bias instrument loop does employ flow transmitters 
to measure recirculation drive flow, delaying entry into Conditions 
and Required Actions associated with SR 3.3.1.1.12 will have no 
impact on their pressure boundary function. Also, failure of the 
sensing line associated with these transmitters has already been 
accounted for in the initial plant design by including excess flow 
check valves for sensing line break isolation.
    The proposed change does not introduce a new mode of plant 
operation and does not involve the installation of any new equipment 
or modifications to the plant. Therefore, it does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    iii)The proposed change does not involve a significant reduction 
in a margin of safety.
    The APRM flow biased high scram function is not specifically 
credited in the safety analysis. However, it is intended to provide 
an additional margin of protection from transient induced fuel 
damage during operation where recirculation flow is reduced to below 
the minimum required for rated power operation.
    The margin of safety associated with this change refers to the 
margin inherent in the accident analyses that takes credit for the 
clamped high flux scram only (i.e., margin between scramming at 120% 
peak flux and the peak flux necessary for fuel damage). The current 
reactor operating state (end of cycle coast down extended core flow) 
dictates that only the 120% flux trip be enforced. This trip remains 
functional during the APRM flow biased high scram calibration.
    Currently, the Conditions and Required Actions associated with 
SR 3.3.1.1.12 permit a one hour delay prior to entry because it 
minimizes risk while allowing time for restoration or tripping of 
channels by operations personnel. Because the APRM flow biased 
function is not enforced during end of cycle, coast down, extended 
core flow conditions, extending entry in associated Conditions and 
Required Actions from one to six hours has no impact on the margin 
associated with the clamped high flux scram. In the event core flow 
drops below 82%, the flow point below which APRM setpoints 
automatically become flow biased, the associated Conditions and 
Required Actions will be entered.
    Therefore, extending entry into associated Conditions and 
Required Actions associated with SR 3.3.1.1.12, provided core flow 
remains at or above 82%, from one to six hours does not reduce 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 16, 1996
    Description of amendment request: The proposed amendment to the 
James A. FitzPatrick Technical Specifications (TSs) proposes to delete 
the requirement for the Plant Operating Review Committee (PORC) to 
review the fire protection program and implementing procedures. This 
proposal will reduce the administrative burden on the committee while 
making PORC's responsibilities more consistent with the other 
responsibilities described in Section 6.1.5.6 of the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes delete the Plant Operating Review Committee 
(PORC) review of the fire protection program and implementing 
procedures, and deleted fire protection inspection and audit 
requirements that are redundant to those performed under the 
cognizance of the Safety Review Committee (SRC). The changes do not 
introduce any new modes of plant operation, make any physical 
changes, or alter any operational setpoints. Therefore, the changes 
do not degrade the performance of any safety system assumed to 
function in the accident analysis. Consequently, there is no effect 
on the probability or consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated.
    No physical changes to the plant or changes to equipment 
operating procedures are proposed. The changes are administrative 
and will not have any direct affect on equipment important to 
safety. Therefore the changes cannot create the possibility of a new 
or different kind of accident.

[[Page 34896]]

    3. Involve a significant reduction in the margin of safety.
    Adequacy of the fire protection program and implementing 
procedures is assured by the fire protection license condition, the 
procedure review and approval process implemented by Amendment 222, 
the provisions of 10 CFR 50.59, and inspections and audits performed 
under the cognizance of the SRC. Consequently, deleting PORC's 
responsibility for review of the fire protection program and 
implementing procedures, and deleting the inspection and audit 
requirements contained in Specification 6.14.A and 6.14.B will not 
degrade the fire protection program. Therefore, the proposed changes 
do not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019
    NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

    Date of amendment request: May 30, 1996
    Description of amendment request: The proposed amendment would 
revise Minimum Critical Power Ratio Safety Limit and associated basis. 
The changes are required to support introduction of General Electric 
Company supplied, GE12, 10x10 fuel into the Cycle 13 reactor core.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    A change in the SLMCPR [Safety Limit Minimum Critical Power 
Ratio] does not affect initiation of any accident. Operation in 
accordance with the revised SLMCPR ensures the consequences of 
previously analyzed accidents are not changed.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The SLMCPR establishes a performance limit for the fuel. 
Therefore changing the limit will not initiate any accident.
    3. Involve a significant margin of safety because:
    The analyses performed to determine the revised SLMCPR assure 
maintenance of the same margin of safety as presently exists for the 
prevention of onset of transition boiling.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019
    NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

    Date of amendment request: May 30, 1996
    Description of amendment request: The proposed amendment would 
revise Anticipated Transient Without Scram (ATWS) Recirculation Pump 
Trip Reactor Pressure - High setpoint when either zero or one Safety 
Relief Valves are out-of service.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    A change in the ATWS Recirculation Pump Trip Reactor Pressure - 
High setpoint does not affect initiation of any accident. Operation 
in accordance with the revised setpoints ensures the consequences of 
previously analyzed accidents are not changed.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    RPV [reactor pressure vessel] pressure following an ATWS with 
MSIV [main steam isolation valve] closure event (worst case 
transient for RPV pressurization) remains within acceptable limits 
with the revised setpoint. Therefore changing the setpoint will not 
lead to a new type of accident.
    3. Involve a significant reduction in a margin of safety 
because:
    The analyses performed to determine the revised ATWS 
Recirculation Pump Trip Reactor Pressure - High setpoint assure 
maintenance of the same margin of safety as presently exists for 
limiting RPV pressure following an ATWS with MSIV closure (limiting 
transient).
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019
    NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

    Date of amendment request: May 30, 1996
    Description of amendment request: The proposed amendment would 
eliminate selected response time testing requirements. The affected 
Technical Specifications (TS) are TS 4.1.A, ``Surveillance 
Requirements, Reactor Protection System,'' and TS 4.2.A, ``Surveillance 
Requirements, Instrumentation, Primary Containment Isolation 
Functions.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The purpose of the proposed TS change is to eliminate response 
time testing requirements for selected sensors in the RPS [reactor 
protection system] and Primary Containment Isolation System. The 
BWROG [Boiling Water Reactor Owners Group] has completed an 
evaluation which demonstrates that response time testing is 
redundant to the other TS required testing. These other tests

[[Page 34897]]

in conjunction with actions taken in response to NRC Bulletin 90-01, 
``Loss of Fill-Oil in Transmitters
    Manufactured by Rosemount,'' and Supplement 1 to Bulletin 90-01, 
are sufficient to identify failure modes or degradation in 
instrument response time and ensure operation of the associated 
systems within acceptable limits. Furthermore, failure modes 
detected by response time testing are detectable by other TS 
required testing. This evaluation was documented in Reference 1 [See 
application dated May 30, 1996]. NYPA [New York Power Authority] has 
confirmed the applicability of this evaluation to the FitzPatrick 
Plant. In addition, NYPA will complete the actions identified in the 
NRC staff's safety evaluation of NEDO-32291-A.
    Because of the continued application of other existing TS 
required tests such as channel calibrations, channel checks, channel 
functional tests, and logic system functional tests, the response 
time of these systems will be maintained within the acceptance 
limits assumed in plant safety analyses and required for successful 
mitigation of an initiating event. The proposed changes do not 
affect the capability of the associated systems to perform their 
intended function within their required response time, nor do the 
proposed changes themselves affect the operation of any equipment. 
Therefore, the proposed 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 those previously evaluated because:
    The proposed changes do not affect the ability of the systems to 
perform their intended function within the acceptance limits assumed 
in plant safety analyses and required for successful mitigation of 
an initiating event. No new failure modes are introduced by the 
changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The current TS required response time test limits are based on 
the maximum allowable values assumed in the plant safety analyses. 
These analyses conservatively establish the margin of safety. As 
described above, the proposed changes do not affect the capability 
of the associated systems to perform their intended function within 
the allowed response time used as the basis for the plant safety 
analysis. Plant and system response to an initiating event will 
remain in compliance within the assumptions of the safety analyses, 
and therefore the margin of safety is not affected.
    Further, although not explicitly evaluated, the proposed changes 
will provide an improvement to plant safety and operation by 
reducing the time safety systems are unavailable, reducing safety 
systems actuations, reducing plant shutdown risk, limiting radiation 
exposure to plant personnel, and eliminating the diversion of key 
personnel to conduct unnecessary testing. Therefore, the overall 
effect of the changes should increase the margin the safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019
    NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

    Date of amendment request: March 6, 1996, as supplemented by letter 
dated May 30, 1996
    Description of amendment request: The proposed change to Hope Creek 
Technical Specification (TS) 3.8.1, ``A.C. Sources - Operating'', would 
decrease the minimum fuel oil storage capacity of the Emergency Diesel 
Generator Fuel Oil Storage Tanks, from 48,800 to 44,800 gallons. In 
addition, footnote ** is deleted from TS 3.8.1.1.b.2. The proposed 
change would also add an Action Statement to address remedial action 
when a fuel oil transfer pump becomes inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    TANK LEVEL
    Amendment 59 provides an allowance for transferring fuel oil 
from a pair of storage tanks associated with an inoperable 
[Emergency Diesel Generator] EDG to another pair of storage tanks in 
order to demonstrate compliance with PSE&G's commitment to 
Regulatory Guide 1.137. The proposed change is consistent with that 
transfer strategy and extends this allowance to include using fuel 
oil in operable EDG storage tanks in order to reduce the amount of 
stored fuel oil. Transfer from operable EDG storage tanks is, 
actually, less complex than transferring from an inoperable EDG 
storage tank since power to the transfer pumps would be available.
    The low level alarm setpoint is the only physical change to be 
made. No change is being made to the EDGs, to the fuel oil storage 
tanks, or to the fuel oil transfer system and since EDG fuel oil 
supply is associated with mitigating the consequences of an 
accident, there is no change in the probability of any accident 
analyzed in the [Updated Final Safety Analysis Report] UFSAR.
    Since the proposed change still ensures the minimum fuel oil 
storage capacity meets the existing licensing basis and since off-
site replacement oil is expected to be available within 60 hours 
there is no change in the consequences of an accident previously 
evaluated.
    TRANSFER PUMP ACTION STATEMENT
    Since no change is being made to the EDGs, to the fuel oil 
storage tanks or to the fuel oil transfer system, and since EDG fuel 
oil supply is associated with mitigating the consequences of an 
accident, there is no change in the probability of any accident 
analyzed in the UFSAR.
    The proposed change provides compensatory action in the event a 
single fuel oil transfer pump is inoperable without having to 
immediately declare the EDG inoperable. The change ensures the 
affected EDG remains fully capable of functioning as assumed in the 
safety analyses, therefore, there is no significant impact on the 
consequences of an accident previously evaluated.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    TANK LEVEL AND TRANSFER PUMP ACTION STATEMENT
    The proposed changes will result in a setpoint change to the low 
level alarm. No other physical changes to the EDGs, to the fuel oil 
storage tanks, or to the fuel oil transfer system will result from 
the proposed changes. Operation including the proposed changes will 
not impair the diesel generators from performing as provided in the 
design basis. In addition, EDG fuel oil supply is associated with 
mitigating accident consequences, not accident prevention. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Will not involve significant reduction in a margin of safety.
    TANK LEVEL
    The margin of safety is provided by the on-site storage of an 
adequate supply of diesel fuel oil to ensure uninterrupted EDG 
operation for seven days. Although the proposed change may result in 
a reduction of stored fuel oil, the new minimum continues to provide 
for an on-site seven day supply of diesel fuel oil.
    TRANSFER PUMP ACTION STATEMENT
    The margin of safety is provided by the ability of the fuel oil 
transfer pumps to supply an adequate flow of the stored fuel to each 
EDG day tank. The proposed change continues to provide 100% capacity 
to the EDG day tank for a minimum of three days with no operator 
action. With the proposed action, adequate transfer capability is

[[Page 34898]]

provided for a minimum of seven days fuel oil supply at which time 
refilling of the tanks would provide an indefinite supply. With both 
transfer pumps on a single EDG inoperable, the remaining three EDGs 
would provide adequate power for safe shutdown. Transfer of fuel oil 
from the storage tanks with inoperable transfer pumps can still be 
effected using temporary hoses.
    Since the proposed changes do not involve the addition of plant 
equipment, are consistent with the intent of the existing Technical 
Specifications, are consistent with allowances for fuel oil 
transfers approved in Amendment 59, meets the intent of Regulatory 
Guide 1.137, and are consistent with the design basis of the diesel 
generators and the accident analysis, no action proposed by this 
request will occur that will involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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: May 10, 1996
    Description of amendment request: The proposed amendments would 
change Technical Specification Sections, 1.0, 2.0, 3/4 1.0, 3/4 2.0, 
5.0 and 6.0. These changes support the Margin Recovery Program (MRP) 
and support increased steam generator tube plugging, improved fuel 
reliability, reduced fuel costs, longer fuel cycles, reduced spent fuel 
storage, and enhanced reactor safety. These changes incorporate the 
results of the revised safety analyses (margin recovery) and the 
establishment of a Core Operating Limits 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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The accidents potentially affected by the parameters and 
assumptions associated with the MRP have been evaluated/ analyzed 
and all design standards and applicable safety criteria are met. The 
consideration of these changes does not result in a situation where 
the design, material, or construction standards that were applicable 
prior to the change have been altered. Therefore, the changes 
occurring with the MRP will not result in any additional challenges 
to plant equipment that could increase the probability of any 
previously evaluated accident.
    The changes associated with the MRP do not affect plant systems 
such that their function in the control of radiological consequences 
is adversely affected. The safety evaluation documents that the 
design standards and applicable safety criteria limits continue to 
be met and therefore fission barrier integrity is not challenged. 
The MRP changes have been shown not to adversely affect the response 
of the plant to postulated accident scenarios. In all cases, the 
calculated doses are within the regulatory criteria and therefore do 
not constitute an increase in consequences. These changes will, 
therefore, not affect the mitigation of the radiological 
consequences of any accident described in the Updated Final Safety 
Analysis Report (UFSAR).
    Based on the above, it is concluded that the probability or 
consequences of an accident previously evaluated is not 
significantly increased by the proposed changes.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The possibility for a new or difference[t] type of accident from 
any accident previously evaluated is not created since the changes 
associated with the MRP do not result in a change to the design 
basis of any plant component or system. The evaluation of the 
effects of the MRP changes shows that all design standards and 
applicable safety criteria limits are met. These changes therefore 
do not cause the initiation of a new accident nor create any new 
failure mechanisms. Component integrity is not challenged. The 
changes do not result in any event previously deemed incredible 
being made credible. The MRP changes will not result in more adverse 
conditions and will not result in any increase in the challenges to 
safety systems.
    Therefore, the consideration of the MRP as described in the 
safety evaluation does not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is maintained by assuring compliance with 
acceptance limits reviewed and approved by the NRC. Since all of the 
appropriate acceptance criteria for the various analyses and 
evaluations have been met, by definition there has not been a 
reduction in any margin of safety.
    Therefore, the margin of safety as defined in the Bases to the 
Salem Unit 1 and 2 Technical Specifications has not been 
significantly reduced.
    Based on the above, PSE&G has determined that the proposed 
changes do not involve a significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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

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

    Date of amendment request: April 16, 1996
    Description of amendment request: The proposed amendment would 
revise the Virgil C. Summer Nuclear Station, Unit 1 (VCSNS), Technical 
Specifications (TS) to implement the amended regulation to 10 CFR Part 
50, Appendix J, Option B (new rule), to provide a performance-based 
option for leakage-rate testing of containment. The proposed amendment 
will revise the VCSNS TS 3/4.6 ``Containment Systems,'' TS Bases 3/4.6, 
and TS 6.8 ``Administrative Controls - Programs and Procedures,'' to 
adopt the implementation requirements of 10 CFR Part 50, Appendix J, 
Option B. The proposed amendment utilizes the guidelines (guidelines) 
provided in ``Option B'' of Regulatory Guide (RG) 1.163 ``Performance-
Based Containment Leak-Test Program, September 1995,'' and NEI 94-01, 
``Industry Guideline for Implementing Performance-Based Option of 10 
CFR 50, Appendix J, July 26, 1995.'' The licensee has stated that the 
proposed amendment is within these prescribed guidelines and does not 
propose any deviations to the established methods which would impact 
already approved analyses/justifications and established review 
process.
    The proposed change will remove the prescriptive TS requirements 
for the performance of containment leakage testing and allow leakage 
testing to be conducted as determined appropriate through the 
performance-based or risk-based alternatives described in the VCSNS 
Containment Leakage Rate Testing Program developed in accordance with 
RG 1.163 and NEI 94-01. Since the requirements of Appendix J to 10 CFR 
Part 50 will continue to

[[Page 34899]]

apply, the type of testing will not change. The proposed request does 
not modify any plant equipment or systems.
    The requirements of Appendix J will continue to govern the type of 
test, testing methodology, and acceptance criteria for Type A, B, and C 
testing. The performance-based testing of Option B eliminates or 
modifies prescriptive regulatory requirements for which the burden is 
marginal to safety for which the reviews and analyses have been 
presented in NUREG-1493, ``Performance-Based Containment Leak-Test 
Program, Final Report, September 1995.''
    Earlier leakage testing performed at VCSNS has demonstrated low 
overall containment leakage and supports the implementation of Option 
B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The probability or consequences of an accident previously 
evaluated is not significantly increased.
    There is no increase in the probability of an accident since 
there is no work that would affect containment integrity. The 
testing of containment isolation valves (CIVs) and other containment 
penetration sealing devices is not postulated as an accident 
precursor or initiating event.
    Type A testing is capable of determining the total leakage from 
both local leakage paths and gross containment leakage paths. Our 
Type B and C testing has consistently provided accurate leakage 
rates for valves and penetrations.
    Administrative controls govern maintenance and testing such that 
there is very low probability that unacceptable maintenance or 
alignments can occur. Prior to and following maintenance on CIVs and 
penetrations, a local leak rate test (LLRT) is required to be 
performed. As a result, Type A testing is not required to accurately 
quantify the leakage through containment penetrations.
    Any specific exemptions to the requirements of Appendix J will 
require approval by the NRC before implementation.
    Therefore, this proposed change does not involve a significant 
increase in the possibility or consequences of an accident 
previously evaluated.
    2. The possibility of an accident or a malfunction of a 
different type than any previously evaluated is not created.
    The proposed request does not involve any physical changes to 
the plant, affect the operation of the plant, or change testing 
methods or acceptance criteria. The history of containment testing 
verifies that containment integrity has been maintained.
    The frequency changes allowed by implementation of Option B will 
not significantly decrease the level of confidence in the ability of 
the reactor building to limit offsite doses to allowable values. No 
accident or malfunction can be the result of the allowed changes to 
test schedule or frequency.
    Since the proposed request will not directly impact equipment, 
procedures or operations, the changes will not create the 
possibility of any new or different kind of accident from any 
previously evaluated.
    3. The margin of safety has not been significantly reduced.
    The reason for performing containment leakage rate testing is to 
assure that the leakage paths are identified, and that any accident 
release will be restricted to those paths assumed in the safety 
analysis. The purpose for the schedule is to assure that containment 
integrity is verified on a periodic basis.
    Implementation of Option B to provide flexibility in the 
scheduled requirements does not mean that containment integrity will 
be compromised. The historical leakage rate test results for VCSNS 
and for the nuclear industry support extension of testing 
frequencies and demonstrate that structural integrity has been 
maintained.
    Therefore, the margin of safety has not been significantly 
reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, SC 29218
    NRC Project Director: Eugene V. Imbro

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph 
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: April 22, 1996
    Description of amendment request: The amendment would revise the 
Technical Specifications to implement the L* Tubesheet Region Plugging 
Criterion, which would allow a steam generator tube to remain in 
service with bands of axial degradation in the tubesheet region 
provided sufficient non-degraded tubing remains to satisfy regulatory 
guidance concerning structural and leakage integrity.
    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 Farley Nuclear Plant Unit 2 steam generators 
in accordance with the proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The supporting technical evaluations of the subject criteria 
demonstrate that the presence of the tubesheet enhances the tube 
integrity in the region of the hardroll by precluding tube 
deformation beyond its initial expanded outside diameter. The 
resistance to both tube rupture and tube collapse is strengthened by 
the presence of the tubesheet in that region. The result of the 
hardroll of the tube into the tubesheet is an interference fit 
between the tube and the tubesheet. Tube rupture [cannot] occur 
because the contact between the tube and tubesheet does not permit 
sufficient movement of tube material. In a similar manner, the 
tubesheet does not permit sufficient movement of tube material to 
permit buckling collapse of the tube during postulated LOCA [loss-
of-coolant accident] loadings.
    The type of degradation for which the L* criterion has been 
developed (cracking with an axial or near axial orientation) has 
been found not to significantly reduce the axial strength of a tube. 
An evaluation including analysis and testing has been done to 
determine the strength reduction for axial loads with simulated 
axial and near axial cracks. This evaluation provides the basis for 
the acceptance criteria for tube degradation subject to the L* 
criterion.
    The SRE [sound roll expansion] L* length is sufficient to 
preclude significant leakage from tube degradation located below the 
L* length. The existing Technical Specification leak rate 
requirements and accident analysis assumptions remain unchanged in 
the unlikely event that significant leakage from this region does 
occur. Any leakage from the tube within the tubesheet at any 
elevation in the tubesheet is fully bounded by the existing steam 
generator tube rupture analysis included in the Farley Nuclear Plant 
Final Safety Analysis Report. A conservative leakage allowance for 
each L* tube is provided to determine the impact of L* criterion 
upon offsite doses in the event of a postulated double ended 
guillotine break of the main steam line outside of containment, but 
upstream of the main steam line isolation valves. Since Farley Unit 
2 has implemented the Interim Plugging Criteria (IPC) for ODSCC at 
the tube support plates, projected steam line break (SLB) leakage at 
the end of the next successive operating cycle must be evaluated. 
Per Generic Letter 95-05, plants implementing the IPC can utilize 
SLB leakage limits higher than the originally assumed 1.0 gpm 
primary to secondary leakage value provided an analysis of offsite 
doses consistent with the Standard Review Plan methodology is 
performed. This analysis performed for the Farley Unit 2 plant 
indicates that primary to secondary leakage of 11.2 gpm in the 
faulted loop (0.1 gpm in the intact loops) will result in offsite 
doses at the site boundary of less than 10% of the 10 CFR [Part] 100 
guidelines. The total projected SLB leakage from all leakage sources 
must remain below this value. Per attachment 4 addressing the L* 
methodology,

[[Page 34900]]

the number of tube ends to which L* criterion can be applied is 
limited to 600 per steam generator. Using a bounding SLB leakage 
allowance per L* tube, the SLB leakage component from 600 L* tube 
ends will be less than 0.33 gpm in the faulted loop. The proposed 
alternate plugging criterion does not adversely impact any other 
previously evaluated design basis accident. As the current Unit 2 
IPC SLB leakage has been calculated to be less than 2 gpm in the 
faulted loop, [an] SLB leakage margin of over 9 gpm is provided for 
this cycle.
    As noted above, tube rupture and pullout is not expected for 
tubes using the L* criterion. In addition to the L* length, a 
minimum length of SRE below the identified degradation must be 
established. The aggregate L* distance of SRE provides the 
structural integrity to prevent tube pullout. Conservatively, it is 
assumed that the degraded band length does not provide any support 
in resisting tube pullout.
    Therefore SNC [Southern Nuclear Operating Company, Inc.] 
concludes that Operation of the Farley Nuclear Plant Unit 2 steam 
generators in accordance with the proposed license amendment does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the proposed L* criterion does not introduce 
any significant changes to the plant design basis. Use of the 
criterion does not provide a mechanism to result in an accident 
initiated outside of the region of the tubesheet expansion. The 
structural integrity of L* tubes will be maintained during all plant 
conditions. Any hypothetical accident as a result of any tube 
degradation in the expanded portion of the tube would be bounded by 
the existing tube rupture accident analysis. If it is postulated 
that a circumferential separation of an L* tube were to occur below 
the PLRL [pullout load reaction length], tube structural and leakage 
integrity will be maintained during all plant conditions. 
Verification of the L* distance of non-degraded tube roll expansion 
prevents the postulated separated tube from lifting out of the 
tubesheet during all plant conditions. Verification of the L* 
criterion prevents tube displacement of any magnitude, and 
therefore, postulated axial cracks existing a minimum of 0.5 inch 
from either the bottom of the roll transition or top of tubesheet, 
whichever is lower, from migrating out of the tubesheet.
    Therefore, SNC concludes that the proposed license amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    The use of the L* criterion has been concluded to maintain the 
integrity of the tube bundle commensurate with the requirements of 
draft Regulatory Guide 1.121 under normal and postulated accident 
conditions. The safety factors used in the verification of the 
strength of the degraded tube are consistent with the safety factors 
in the ASME [American Society of Mechanical Engineers] Boiler and 
Pressure Vessel Code used in steam generator design. The L* length 
has been verified by testing to be greater than the length of roll 
expansion required to preclude significant leakage during normal and 
postulated accident conditions. The leak testing acceptance criteria 
are based on the primary to secondary leakage limit in the Technical 
Specifications and the leakage assumptions used in the FSAR accident 
analyses. The L* distance provides for structural integrity during 
all plant conditions.
    Implementation of the L* criterion will decrease the number of 
tubes which must be taken out of service with tube plugs or repaired 
with sleeves. Both plugs and sleeves reduce the RCS [reactor coolant 
system] flow margin, thus implementation of the L* criterion will 
maintain the margin of flow that would otherwise be reduced in the 
event of increased plugging or sleeving.
    Therefore, SNC, concludes based on the above, it is concluded 
that the proposed change does not result in a significant reduction 
in a loss of margin with respect to plant safety as defined in the 
Final Safety Analysis Report [FSAR] or the bases of the FNP [Farley 
Nuclear Plant] technical specifications.
    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: Herbert N. Berkow

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

    Date of application request: May 29, 1996
    Description of amendment request: The application requests staff 
review and approval of a modification to the facility, as described in 
the safety analysis report, that involves an unreviewed safety 
question. The modification will reduce the single failure trip 
potential for the main feedwater control and bypass valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Callaway safety analysis assumes the MFC&BVs [main feedwater 
control and bypass valves] close during certain events in order to 
terminate fluid inventory addition to faulted steam generators and 
thereby preclude the diversion of auxiliary feedwater to the main 
feedwater system. This feature is necessary because each feedwater 
line at Callaway is equipped with only one MFIV [main feedwater 
isolation valve]. It should be noted that the safety analysis simply 
requires the valves to close and does not prescribe a mechanism for 
accomplishing that action.
    The following are accidents that credit feedwater isolation or 
AFW [auxiliary feedwater] addition. There is no impact by the 
proposed modification on the consequences of each accident.
    <bullet>Feedwater System Malfunctions That Result In An Increase 
In Feedwater Flow
    <bullet>Inadvertent Opening Of A Steam Generator Relief or 
Safety Valve
    <bullet>Steam System Piping Failure
    <bullet>Loss of Nonemergency AC Power to the Station Auxiliaries
    <bullet>Loss of Normal Feedwater Flow
    <bullet>Feedwater System Pipe Break
    <bullet>Decrease in Reactor Coolant Inventory
    The modification will not change the radiological consequences 
of FSAR [final safety analysis report] Chapter 15 accidents because 
the feedwater isolation function (and NSSS [nuclear steam supply 
system] break response) has not changed. Therefore, there will be no 
increase in the consequences of an accident evaluated previously in 
the FSAR.
    An analysis was performed to quantify the impact of the proposed 
modification on the probability of MFCV [main feedwater control 
valve] failure (closure) during normal plant operation. Comparison 
of this failure probability for the existing design (1.20E-1 per 
year) versus the proposed design (6.99E-2 per year) indicates that 
the percentage reduction in the system failure probability at power 
is 41.75%. Thus, the proposed design results in a reduction in the 
probability of inadvertent MFCV failures at power and hence, a 
reduction in the probability of a reactor trip and subsequent 
challenges to other safety systems.
    While this modification reduces the probability of a reactor 
trip, it slightly increases the unavailability of the feedwater 
isolation function. This is because the original design required 
actuation of only one FWIS [feedwater isolation system] train to 
close the MFC&BVs, whereas the new design requires actuation of both 
trains. The impact of the modification on the probability of 
incurring a feedwater isolation failure was therefore quantified, 
utilizing PRA [probabilistic risk assessment] techniques. Fault 
trees were developed for both the new and existing designs. Failure 
probabilities for each event were then obtained from the IPE 
[individual plant examination] and utilized to calculate failure 
probabilities for the feedwater isolation safety function. This 
calculation considered hardware failures

[[Page 34901]]

only, i.e., failure of an MFIV to close after receiving an actuation 
signal. The failure probability of feedwater isolation, based on the 
proposed design, was determined to be 6.1E-5 per demand (1 event 
every 16,400 demands). The existing design was found to have a 
failure probability of 2.8E-5 per demand (1 event every 35,700 
demands). Therefore, this modification will not significantly 
increase the probability or consequences of an accident evaluated 
previously in the FSAR.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The modification maintains the present de-energize-to-actuate 
configuration of the MFC&BV trip solenoid valves.
    Thus, the proposed modification does not create the possibility 
of an accident of a different type than any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Credit is taken in the accident analyses for the MFIVs to close 
on demand for feedwater isolation. Because of this, the MFIVs have 
been incorporated into the Callaway Technical Specifications. Action 
Statements and surveillance requirements have been developed to 
assure the availability of the valves when needed.
    The MFC&BVs are not addressed by any of the Callaway Technical 
Specifications or their bases. Therefore, this modification will not 
involve a significant reduction in the margin of safety as defined 
in the basis for any technical specification.
    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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: William H. Bateman

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two 
Creeks, Manitowoc County, Wisconsin

    Date of amendment request: May 29, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 15.4.4, ``Containment 
Tests,'' to incorporate the provisions of 10 CFR Part 50, Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors,'' Option B. Revisions would also be made to TS Sections 15.1, 
``Definitions,'' 15.3.6, ``Containment System,'' and 15.6, 
``Administrative Controls,'' to support the proposed changes to Section 
15.4.4.
    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 this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change does not involve a change to structures, 
systems, or components which would affect the probability or 
consequences of an accident previously evaluated in the PBNP [Point 
Beach Nuclear Plant] Final Safety Analyses Report (FSAR). 
Furthermore, containment leakage rate testing is not an initiator of 
any accident. The proposed change simply provides a mechanism within 
the Technical Specifications for implementing a performance-based 
method of determining the frequency for leakage rate testing which 
has been approved by the NRC. The proposed change does not affect 
reactor operations or accident analysis and has no significant 
radiological consequences. Therefore, this change will not create a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that contribute to 
initiation of any accidents. This change involves a potential 
reduction of Type A, B, and C test frequency. Except for the method 
of defining the test frequency, the methods for performing the 
actual tests are not changed. No new accident modes are created by 
extending the testing intervals. No safety-related equipment or 
safety functions are altered as a result of this change. Extending 
the test frequency has no influence on, nor does it contribute to, 
the possibility of a new or different kind of accident or 
malfunction from those previously analyzed. Therefore, the proposed 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The proposed change potentially affects only the frequency of 
Type A, B, and C testing. Except for the method of defining test 
frequency, the methods for performing the actual tests are not 
changed. The proposed change is based on NRC accepted provisions and 
maintains necessary levels of system and component reliability 
affecting containment integrity. Evaluation of the performance-based 
approach to leakage rate testing, as documented in NUREG-1493, 
concludes that the impact on public health and safety due to revised 
testing intervals is negligible. Furthermore, the proposed change 
will not reduce the availability of systems associated with 
containment integrity when they are required to mitigate accident 
conditions. Therefore, the proposed change will not create a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: June 4, 1996
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant Technical Specifications (TS) 
by reducing the surveillance test frequencies for the radiation 
monitoring system (Table TS 4.1-1) and the control rods (Table TS 4.1-
3) in accordance with the guidance of Generic Letter 93-05, ``Line-Item 
Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation,'' dated September 27, 
1993.
    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:
    Table TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations 
and Test of Instrument Channels,'' Item 19
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to determine that no significant hazards 
exist. The proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The radiation monitors are not accident initiators; therefore, 
they cannot increase the probability of an accident occurring. The 
reliability of the radiation monitors is not expected to decrease 
due to the decreased surveillance frequency; therefore, this change 
does not increase the consequences of an accident.

[[Page 34902]]

    The addition of comment (a) to the Check, Calibrate, and Test 
columns is merely a clarification of the existing information in the 
table and does not change the intent of the Technical 
Specifications.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change revises only the testing frequency and does 
not revise the test method or operational performance of the 
radiation monitors. The radiation monitors are not accident 
initiators; therefore, they cannot create a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    Quarterly testing of the radiation monitoring system channels 
will continue to verify operability of the monitors. Decreasing the 
test surveillance frequency is not expected to decrease the 
reliability of the radiation monitors. This change is acceptable in 
accordance with Generic Letter 93-05 and NUREG-1366, ``Improvements 
to Technical Specifications Surveillance Requirements.''
    Table TS 4.1-3, ``Minimum Frequencies for Equipment Tests,'' 
Item 1
    The proposed change in test frequency for control rod exercising 
was reviewed in accordance with the provisions of 10 CFR 50.92 to 
determine that no significant hazards exist. It has been determined 
that the proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change revises only the testing frequency for 
control rod exercising. The control rod exercise surveillance 
procedure will continue to be conducted, on a quarterly basis, to 
ensure that the equipment remains operable. The reduced frequency of 
control rod exercising reduces the probability of an inadvertent 
reactor trip occurring during testing due to a dropped control rod. 
Surveillance procedure SP 49-075 is conducted to verify rod 
movement. In accordance with NUREG-1366, the frequency of a stuck 
control rod occurring is very low. This condition is most often 
discovered during reactor startup or during low power physics 
testing. The reduction in control rod exercising is, therefore, 
considered acceptable and is not expected to affect the probability 
of a stuck control rod occurring.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change revises only the testing frequency and does 
not revise the test method or the design of the control rod system. 
Therefore, a new or different kind of accident will not be created 
by this change.
    3. Involve a significant reduction in the margin of safety.
    Quarterly control rod exercising will continue to verify 
movement of the control rods. No adverse consequences are expected 
to occur due to decreasing the test frequency. This change is 
acceptable in accordance with Generic Letter 93-05 and NUREG-1366.
    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: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: June 10, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification 4.2.b, ``Steam Generator Tubes,'' and 
its associated basis, by allowing the use of Westinghouse laser-welded 
sleeves to repair defective steam generator tubes. A description of the 
sleeving repair process and supporting technical justification are 
contained in WCAP-13088, Revision 3, ``Westinghouse Series 44 and 51 
Steam Generator Generic Sleeving Report.'' WCAP-13088, and a non-
proprietary version (WCAP-13089), were submitted to the Nuclear 
Regulatory Commission on April 13, 1995, to support a similar TS 
amendment request for the DC Cook Nuclear Power Plant, Unit 1.
    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 KNPP [Kewaunee Nuclear Power Plant] in 
accordance with the proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The LWS [laser-welded sleeve] configuration has been designed 
and analyzed in accordance with the requirements of the ASME 
[American Society of Mechanical Engineers] Code. Fatigue and stress 
analyses of the sleeved tube assemblies produced acceptable results; 
i.e., the applied stresses and fatigue usage for the sleeve and weld 
are bounded by the limits established in the ASME Code. ASME Code 
minimum material property values are used for the structural and 
plugging limit analysis. Ultrasonic inspection is used to verify 
that minimum weld fusion zone thicknesses are produced. Mechanical 
testing of 7/8'' tubesheet sleeves installed in roll expanded tubes 
has shown that the individual joint structural strength of Alloy 690 
LWSs provides margin to acceptance limits. These acceptance limits 
bound the most limiting loadings (3 times normal operating pressure 
differential) recommended by RG [Regulatory Guide] 1.121. Therefore, 
each individual joint provides for structural integrity exceeding RG 
recommendations. A hypothetical loss of integrity of one of the 
joints will not result in a loss of structural integrity for the 
sleeve. Leakage testing for 3/4'' and 7/8'' full length tubesheet 
sleeves has demonstrated that unacceptable levels of primary-to-
secondary leakage are not expected during all plant conditions for 
non-welded tubesheet sleeve lower joints. The welded joint produces 
a hermetic seal, and therefore will not leak under any plant 
conditions. Laser welded sleeves will not contribute to the current 
SLB [steam-line break] primary-to-secondary leakage limit of 34 gpm 
in the faulted loop. The 34 gpm leakage limit was calculated in 
accordance with the standard review plan methodology to support 
implementation of the voltage-based repair criteria for tube support 
plate intersections.
    The sleeve minimum acceptable wall thickness (used for 
developing the depth based plugging limit for the sleeve) is 
determined using the guidance of RG 1.121 and the pressure stress 
equation of Section III of the ASME Code. With respect to the design 
of the sleeve for KNPP, the limiting requirement of the RG which 
applies to part throughwall degradation is that the minimum 
acceptable wall must maintain a factor of safety consistent with the 
analysis conditions as defined by the ASME Code. A bounding set of 
design and transient loading input conditions was used for the 
minimum wall thickness evaluation in the generic evaluation. 
Evaluation of the minimum acceptable wall thickness for normal, 
upset and postulated accident condition loading per the ASME Code 
indicates the limiting condition is established for the normal 
operating conditions, and the minimum acceptable wall thickness for 
this case bounds the upset and faulted condition values.
    According to RG recommendations, an allowance for non-
destructive evaluation (NDE) uncertainty and operational growth of 
existing tube wall degradation indications within the sleeve must be 
accounted for when determining the sleeve plugging limit. A 
conservative tube wall degradation growth rate per cycle and an NDE 
uncertainty has been assumed for determining the sleeve TS plugging 
limit. The sleeve wall degradation extent determined by NDE, which 
would require plugging sleeved tubes, is developed using the 
guidance of RG 1.121 and is defined in WCAP-13088 [non-proprietary 
WCAP-13089] to be 25% throughwall (plugging limit = 100% - 
structural limit + NDE uncertainty + growth) for KNPP.
    The hypothetical consequences of failure of the sleeve joint 
would be bounded by the current SG [steam generator] tube rupture 
analysis included in the KNPP Updated Safety Analysis Report. Due to 
the slight reduction in diameter caused by the sleeve wall 
thickness, primary coolant release rates would be slightly less than 
assumed for the SG tube rupture analysis (depending on break 
location), and therefore, would result

[[Page 34903]]

in lower total primary fluid mass release to the secondary system.
    The proposed TS change to use Alloy 690 LWSs does not adversely 
impact any other previously evaluated design basis accidents or the 
results of LOCA [loss of coolant accident] and non-LOCA accident 
analyses for the current TS minimum reactor coolant system flow 
rate. The results of the analyses and testing, as well as plant 
operating experience, demonstrates that the sleeve assembly is an 
acceptable means of maintaining tube integrity. Plugging limit 
criteria are established using the guidance of RG 1.121. 
Furthermore, per RG 1.83 recommendations, the sleeved tube will be 
monitored through periodic inspections with present NDE techniques. 
These measures demonstrate that installation of sleeves spanning 
degraded areas of the tube will restore the tube to a condition 
consistent with its original design basis.
    Corrosion testing of free span LWS joint has indicated that the 
corrosion resistance (relative to roll transitions) can be increased 
by greater than a factor of ten with the application of a PWHT [post 
weld heat treatment] step. Estimations of joint susceptibility based 
on expected far field stresses after heat treatment using the 
expected original tube-to-tubesheet hydraulic expansion transition 
residual stresses and actual time to crack in these transitions at 
KNPP indicate that LWS joint lifetime should exceed the current 
plant license. Consistent with other license amendments addressing 
LWS, all free span laser welds will receive a PWHT; therefore, rapid 
corrosion degradation of the free span joint is not expected. 
Recently performed corrosion testing of LWS joints in locked tube 
conditions indicates that with PWHT the stress corrosion cracking 
resistance and initiation potential in the parent tube weld region 
is greatly enhanced. Similar test results and conclusions would be 
expected for KNPP. The Model 51 SG tube span between the top of the 
tubesheet and the first support plate is such that even lower PWHT 
residual stresses would be expected. Also, the weld placement within 
the hydraulically expanded area and sleeve installation sequence 
have been optimized to provide for some level of heat treatment at 
the upper transition above the weld and lower far field residual 
stress levels. While no parent tube degradation has been detected at 
this elevation, or any other elevation in a laser welded sleeve 
assembly, the relocation of the weld serves to provide further 
resistance to PWSCC [primary water stress corrosion cracking] at 
this elevation. The suggested target PWHT temperature has also been 
optimized in that this temperature provides for adequate PWHT while 
maintaining the parent tube far field stresses.
    Approximately 19,500 LWSs have been installed in the U.S. Of 
this number, over 300 which have up to 3 cycles of operation were 
inspected in 1995 using the CECCO-5 probe. No degradation of the 
sleeves or the parent tube was detected. Operating experience in 
Europe has shown good performance of the LWS joint for up to 5 
cycles of operation. In 1994, approximately 11,200 LWSs were 
installed in the Doel-4 Plant. After one year of operation, all in-
service sleeves were inspected using the +point probe. No service 
induced corrosion was detected. In 1995, approximately 18,600 LWSs 
were installed in two different U.S. plants. Due to their limited 
operational time, these sleeves have not been inspected.
    Conformance of sleeve design with the applicable sections of the 
ASME Code and results of the leakage and mechanical tests support 
the conclusion that installation of LWSs will not increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed license amendment request does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    Installation of LWSs will not introduce significant or adverse 
changes to the plant design basis and does not represent a potential 
to affect any other plant component. Stress and fatigue analysis of 
the repair has shown that the ASME Code and RG 1.121 criteria are 
not exceeded. Installation of LWSs maintains overall tube bundle 
structural and leakage integrity at a level consistent to that of 
the originally supplied tubing during all plant conditions; stresses 
are bounded by the Code and the tubing is leaktight. Sleeving of 
tubes does not provide a mechanism resulting in an accident outside 
of the area affected by the sleeves. Any hypothetical accident as a 
result of potential tube or sleeve degradation in the repaired 
portion of the tube is bounded by the existing tube rupture accident 
analysis. Since the sleeve design does not affect any component or 
location of the tube outside of the immediate area repaired, in 
addition to the fact that the installation of sleeves and the impact 
on current plugging level analyses is accounted for, the possibility 
that laser welded sleeving creates a new or different type of 
accident is not supported.
    Installation of LWSs will reduce the potential for primary-to-
secondary leakage during postulated steam line break while not 
significantly impacting primary coolant flow area in the event of a 
LOCA. By effectively isolating degraded areas of the tube through 
repair, the potential for steam line break leakage is reduced.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety.
    The LWS repair of degraded SG tubes as identified in WCAP-13088 
[non-proprietary WCAP-13089] has been shown by analysis to
    restore the integrity of the tube bundle consistent with its 
original design basis conditions; i.e., tube/sleeve operational and 
faulted conditions stresses and cumulative fatigue usage are bounded 
by the ASME Code requirements and the repaired tubes are leaktight. 
The safety factors used in the design of sleeves for the repair of 
degraded tubes are consistent with the safety factors in the ASME 
Code used in SG design. The design of the LWS lower joint for 7/8'' 
tube sleeves has been verified by testing to sufficiently preclude 
leakage during normal and postulated accident conditions. The 
portions of the installed sleeve assembly which represents the 
reactor coolant pressure boundary will be monitored for the 
initiation and progression of sleeve/tube wall degradation, thus 
satisfying the requirements of RG 1.83. The portion of the tube 
bridged by the sleeve joints is effectively removed from the 
pressure boundary, and the sleeve then forms the new pressure 
boundary. The areas of the sleeved tube assembly which require 
inspection are defined in WCAP-13088 [non-proprietary WCAP-13089]. 
Since the installed sleeves represent a portion of the pressure 
boundary, a baseline inspection of these areas is required prior to 
operation with sleeves installed.
    The effect of sleeving on the design transients and accident 
analyses has been reviewed based on the installation of sleeves up 
to the level of SG tube plugging coincident with the minimum reactor 
coolant flow rate. The installation of sleeves is evaluated as the 
equivalent of some level of SG tube plugging. This is based on 
determining the minimum reactor coolant flow for the LOCA 
evaluation. Information provided in WCAP-13088 [non-proprietary 
WCAP-13089] describes the method to determine the flow equivalent 
for all combinations of tubesheet and tube support plate sleeves. 
Therefore, installation of LWSs will not result in 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: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two 
Creeks, Manitowoc County, Wisconsin

    Date of amendment request: June 4, 1996 (VPNPD-96-035)
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 15.2.3, ``Limiting Safety 
System Settings and Protective Instrumentation,'' and Section 15.5.3, 
``Design Features - Reactor,'' to incorporate changes associated with 
the operation of Point Beach Nuclear Plant (PBNP), Unit 2, with 
replacement steam generators.
    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 this facility under the proposed Technical 
Specifications will not

[[Page 34904]]

create a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed changes do not involve a change to structures, 
systems, or components which would affect the probability or 
consequences of an accident previously evaluated in the PBNP Final 
Safety Analyses Report (FSAR). The proposed setpoints maintain the 
margin to safe operation of Unit 2 with the replacement steam 
generators. In order to maintain one set of safety analyses for both 
units, the analyses for operation of Unit 2 with the replacement 
steam generators were performed to encompass the operation of Unit 
1. Therefore, the proposed changes apply to the operation of both 
units and maintain the margin of safety for each. The proposed 
change to the description of nominal RCS [reactor coolant system] 
volume is an administrative change and has no effect on plant 
operation. Therefore, the probability or consequences of a 
previously evaluated accident are not significantly increased as a 
result of these changes.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve a change to the plant 
design. The proposed setpoints maintain the margin to safe operation 
of Unit 2 with the replacement steam generators. In order to 
maintain one set of safety analyses for both units, the analyses for 
operation of Unit 2 with the replacement steam generators were 
performed to encompass the operation of Unit 1. Therefore, the 
proposed changes apply to the operation of both units and maintain 
the margin of safety for each. These changes do not affect any of 
the parameters or conditions that contribute to initiation of any 
accidents. The proposed change to the description of nominal RCS 
volume is an administrative change and has no effect on plant 
operation or initiation of any accidents. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The proposed setpoints maintain the margin to safe operation of 
Unit 2 with the replacement steam generators. In order to maintain 
one set of safety analyses for both units, the analyses for 
operation of Unit 2 with replacement steam generators were performed 
to encompass the operation of Unit 1. Therefore, the proposed 
changes apply to the operation of both units and maintain the margin 
of safety for each. The proposed change to the description of 
nominal RCS volume is an administrative change and has no effect on 
plant operation. Therefore, the proposed changes will not create a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two 
Creeks, Manitowoc County, Wisconsin

    Date of amendment request: June 4, 1996 (VPNPD-96-036)
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 15.2.1, ``Safety Limit, 
Reactor Core,'' 15.2.3, ``Limiting Safety System Settings, Protective 
Instrumentation,'' and Section 15.3.1.G, ``Operational Limitations,'' 
to maintain safety margin for Unit 2 with replacement steam generators.
    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 this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change does not involve a change to structures, 
systems, or components which would affect the probability or 
consequences of an accident previously evaluated in the PBNP [Point 
Beach Nuclear Plant] Final Safety Analyses Report (FSAR). The 
proposed changes maintain the margin to safe operation for Unit 2 
with the replacement steam generators. In order to maintain one set 
of safety analyses for both units, the analyses for operation of 
Unit 2 with the replacement steam generators were performed to 
encompass the operation of Unit 1. Therefore, the proposed changes 
apply to the operation of both units and maintain the margin of 
safety for each. The proposed changes do not change, degrade, or 
preclude the prevention or mitigation of the consequences of any 
accident described in the FSAR. Therefore, the probability or 
consequences of a previously evaluated accident are not 
significantly increased as a result of these changes.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve a change to the plant 
design. The proposed changes maintain the margin to safe operation 
for Unit 2 with the replacement steam generators. In order to 
maintain one set of safety analyses for both units, the analyses for 
operation of Unit 2 with the replacement steam generators were 
performed to encompass the operation of Unit 1. Therefore, the 
proposed changes apply to the operation of both units and maintain 
the margin of safety for each. These changes do not affect any of 
the parameters or conditions that contribute to initiation of any 
accidents. In addition, the safety functions of safety-related 
systems and components, which are related to accident mitigation, 
have not been altered. Therefore, the proposed changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The proposed changes maintain the margin to safe operation for 
Unit 2 with the replacement steam generators. In order to maintain 
one set of safety analyses for both units, the analyses for 
operation of Unit 2 with replacement steam generators were performed 
to encompass the operation of Unit 1. Therefore, the proposed 
changes apply to the operation of both units and maintain the margin 
of safety for each. The proposed changes have no affect on the 
availability, operability, or performance of the safety-related 
systems and components described in the Technical Specifications. 
Therefore, the proposed changes will not create a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

[[Page 34905]]

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.

Florida Power and Light Company, et al., Docket No. 50-335, St. 
Lucie Plant, Unit No. 1, St. Lucie County, Florida

    Date of amendment request: June 1, 1996
    Description of amendment request: Revise Technical Specifications 
to reflect reduced reactor coolant system flows resulting from 
increased percentage of plugged steam generator tubes.
    Date of publication of individual notice in the Federal Register: 
June 7, 1996 (61 FR 29140)
    Expiration date of individual notice: June 24, 1996
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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: May 31, 1996
    Brief description of amendment request: The amendments (1) revise 
the Reactor Vessel Level Indication System (RVLIS) Action Statements to 
facilitate actions necessary for channel testing to be performed in 
Mode 3, (2) revise the Channel Calibration definition to better account 
for temperature detector channel calibration methodology, and (3) 
delete a requirement to install a jumper in the Auxiliary Feedwater 
actuation logic since a design change will result in the jumper 
function being performed by a relay.
    Date of publication of individual notice in Federal Register: June 
17, 1996 (61 FR 30641)
    Expiration date of individual notice: July 17, 1996
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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.

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

    Date of application for amendments: May 28, 1996, as supplemented 
by letters dated May 31 and June 5, 1996
    Brief description of amendments: These amendments authorize the 
licensee to revise applicable Updated Final Safety Analysis Report 
sections to reflect the installation of a variable flow controller for 
the service water inlet control valves for the containment air coolers 
that is not within the current licensing basis of Calvert Cliffs 
Nuclear Power Plant Units No. 1 and No. 2. These amendments are being 
issued pursuant to the requirements of 10 CFR 50.59(c) because the 
review by Baltimore Gas and Electric Company identified the changes as 
an unreviewed safety question. No changes to the Technical 
Specifications are required by these amendments.
    Date of issuance: June 17, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 215 and 192
    Facility Operating License Nos. DPR-53 and DPR-69: The amendments 
revised the Updated Final Safety Analysis Report. Public comments 
requested as to proposed no significant hazards consideration: Yes (61 
FR 27371 dated May 31, 1996). The 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 July 1, 1996, 
but indicated that if the Commission makes a final no significant 
hazards consideration determination, any such hearing would take place 
after issuance of the amendments. The May 31 and June 5, 1996, letters 
provided additional information that did not change the scope of the 
May 28, 1996, application.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and a final no significant hazards consideration 
determination are contained in a Safety Evaluation dated June 17, 1996.
    Local Public Document Room location:  Calvert County Library, 
Prince Frederick, Maryland 20678

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: October 24, 1994, as 
supplemented August 31, 1995 and February 8, 1996. The August 31, 1995 
and February 8, 1996, letters provide clarification information. The 
new information changed the scope of the October 24, 1994, letter and 
was re-noticed on May 8, 1996, but did not change the initial no 
significant hazards consideration determination.
    Brief description of amendment: The proposed amendment would revise 
the TS to allow the relocation of TS 3/4.11.2.6, Gas Storage Tanks; and 
the associated Bases in the TS to licensee-controlled documents.
    Date of issuance: June 12, 1996
    Effective date: June 12, 1996
    Amendment No.: 64
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60379). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated

[[Page 34906]]

June 12, 1996. 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, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: December 21, 1995
    Brief description of amendments: The amendments delete the 
requirement to place the reactor mode switch in the Shutdown position 
if a stuck open safety/relief valve can not be closed within 2 minutes. 
The operator will still be required to scram the reactor if suppression 
pool average water temperature reaches 110 degrees Fahrenheit or 
greater. The amendment also includes editorial changes to the index 
pages.
    Date of issuance: June 18, 1996
    Effective date: Immediately, to be implemented within 60 days
    Amendment Nos.: 113 and 98
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20844) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 18, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348

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

    Date of application for amendments: July 18, 1994, as supplemented 
by letter dated October 9, 1994
    Brief description of amendments: The amendments revise the current 
combined Technical Specifications (TS) for Units 1 and 2 by separating 
them into individual volumes for Unit 1 and Unit 2. In addition to the 
changes required by the TS split, some administrative and editorial 
changes were made, such as the correction of typographical errors and 
the deletion of unnecessary blank pages.
    Date of issuance: June 12, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 148 and 142
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47166) The October 9, 1995 and June 6, 1996, letters provided 
clarifying information that did not change the scope of the July 18, 
1994, application and the initial proposed no significant hazards 
consideration. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 12, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: February 22, 1996
    Brief description of amendment: The amendment increased the safety 
function lift setpoint tolerances for the safety and relief valves that 
are listed in Surveillance Requirement 3.4.4.1 (Page 3.4-10) of the 
Technical Specifications (TSs) for the Grand Gulf Nuclear Station, Unit 
1. The tolerances were increased from the current plus or minus 1 
percent of the safety function (i.e., safety relief valve) lift 
setpoint to plus or minus 3 percent.
    Date of issuance: June 12, 1996
    Effective date: June 12, 1996
    Amendment No: 123
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13524) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 12, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120

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: May 1, 1996
    Brief description of amendment: The amendment revises the Operating 
License and Technical Specifications (TS) to implement 10 CFR Part 50, 
Appendix J - Option B, by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program.'' Specifically, 
changes have been made to paragraph 2.D of the Operating License; TS 
Section 1.1, ``Definitions;'' TS 3.6.1.1, ``Primary Containment;'' TS 
3.6.1.1, ``Primary Containment Air Locks;'' TS 3.6.1.3, ``Primary 
Containment Isolation Valves (PCIVs);'' and TS Section 5.5, ``Programs 
and Manuals.''
    Date of issuance: June 21, 1996
    Effective date: June 21, 1996
    Amendment No.: 105
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1996 (61 FR 
25708) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 21, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

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

    Date of application for amendment: January 25, 1996
    Brief description of amendment: The amendment revises Technical 
Specification 3/4.3.3, Emergency Core Cooling System Actuation 
Instrumentation, to more clearly define when, during shutdown and 
refueling, the Loss of Voltage and Degraded Voltage relays for the Loss 
of Power actuation trip functions are required to be operable.
    Date of issuance: June 10, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 72
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20851) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 10, 1996. 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

[[Page 34907]]

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: January 5, 1996, as supplemented 
on May 31, 1996
    Brief description of amendment: The amendment implements the 
guidance of Generic Letter 93-08 by relocating Tables 3.3-2, ``Reactor 
Protective Instrumentation Response Times'' and 3.3-5, ``Engineered 
Safety Features Response Times'' from the Technical Specifications to 
the Millstone Unit No. 2 Technical Requirements Manual (TRM). In 
accordance with Generic Letter 93-08, the Limiting Conditions for 
Operations for Technical Specifications 3.3.1.1, 3.3.2.1, and 3.7.1.6 
are revised to eliminate their references to the aforementioned tables. 
The amendment also revises Bases 3/4.3.1 and 3/4.3.2 to reference that 
the instrument response times are located in the TRM and that these 
tables in the TRM are now controlled under 10 CFR 50.59. The amendment 
also removes a cycle-specific note from Tables 3.3-3 and 3.3-4.
    Date of issuance: June 10, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 198
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5816) The May 31, 1996, letter provided additional information that 
did not change the scope of the January 5, 1996, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 10, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360 and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, Connecticut 06385

PECO Energy 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 application for amendments: February 15, 1996
    Brief description of amendments: The amendments change the 
Technical Specifications to implement 10 CFR Part 50, Appendix J, 
Option B, by creating Technical Specification Section 5.5.12, ``Primary 
Containment Leakage Rate Testing Program,'' which refers to Regulatory 
Guide 1.163, ``Performance-Based Containment Leakage-Test Program.''
    Date of issuance: June 18, 1996
    Effective date: Both units, as of date of issuance, to be 
implemented by June 28, 1996.
    Amendments Nos.: 214 and 219
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13531) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 18, 1996. 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: April 25, 1996
    Brief description of amendments: The amendments relocate Technical 
Specification Traversing In-Core Probe System Limiting Condition for 
Operation 3/4.3.7.7 and its Bases 3/4.3.7.7, to the Limerick Generating 
Station Technical Requirements Manual, and modify Note (f) of TS Table 
4.3.1.1-1.
    Date of issuance: June 11, 1996
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos.: 117 and 79
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20840) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 11, 1996 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: March 14, 1996
    Brief description of amendment: The proposed changes would allow a 
one-time extension of the intervals for the steam generator tube 
inspection that is due in July 1996.
    Date of issuance: June 19, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 166
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20854) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 19, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610

Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of application for amendment: February 28, as supplemented 
April 15, and June 3, 1996.
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications (TS) to increase the surveillance 
intervals for ice bed weight sampling and flow passage inspection from 
9 months to 18 months. The TS would also be changed to provide an 
increased ice sublimation allowance, associated with the increased 
surveillance interval, by increasing the minimum total ice weight from 
2,360,875 pounds to 2,403,800 pounds (1214 pounds/basket to 1236 
pounds/basket).
    Date of issuance: June 13, 1996
    Effective date: June 13, 1996
    Amendment No.: 2
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15998) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 13, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402

[[Page 34908]]

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

    Date of application for amendment: April 26, 1996
    Brief description of amendment: The amendment corrected minor 
technical and administrative errors in the Improved Technical 
Specifications prior to its implementation.
    Date of issuance: June 18, 1996
    Effective date: June 18, 1996
    Amendment No.: 85
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 9, 1996 (61 FR 
21213) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 18, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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 requests: April 25 (TXX-94119) and August 12, 
1994 (TXX-94216), as supplemented by letters dated February 15 (TXX-
96055), March 7 (TXX-96078), and April 11, 1996 (TXX-96111).
    Brief description of amendments: These amendments modified the 
Administrative Controls specifications, relocate/remove requirements 
that are adequately controlled by existing regulations other than 10 
CFR 50.36 and the technical specifications. Guidance on the proposed 
changes was developed by NRC and provided in the Standard Technical 
Specifications for Westinghouse Plants, NUREG-1431. The changes also 
update unit staff qualification requirements to Regulatory Guide 1.8, 
Revision 2.
    Date of issuance: June 12, 1996
    Effective date: June 12, 1996, to be implemented witnin 60 days.
    Amendment Nos.: Unit 1 - Amendment No. 50; Unit 2 - Amendment No. 
36
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39599) and September 28, 1994 (59 FR 49439). The additional information 
contained in the supplemental letters dated February 15, March 7, and 
April 11, 1996, were clarifying in nature and thus, within the scope of 
the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated June 12, 1996. 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: March 12, 1996 (TXX-96008)
    Brief description of amendments: The amendments revised the 
Technical Specifications to reflect the approval for the licensee to 
use of the new Containment Leakage Rate Testing Program as required by 
10 CFR Part 50, Appendix J, Option B for Comanche Peak Steam Electric 
Station, Units 1 and 2. Implementation of the new performance based 
leakage rate testing program will be based on the guidance provided by 
Regulatory Guide 1.163, September 1995.
    Date of issuance: June 13, 1996
    Effective date: June 13, 1996, to be implemented within 60 days
    Amendment Nos.: 51 and 37
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15999) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 13, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019

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

    Date of application for amendment: September 9, 1994, as superseded 
by letter dated July 25, 1995, and subsequently supplemented by letters 
dated February 28, 1996, and April 9, 1996.
    Brief description of amendment: The amendment would revise TS 3/
4.8.1 and its associated Bases to improve the overall emergency diesel 
generator reliability and availability.
    Date of issuance: June 17, 1996
    Effective date: June 17, 1996, to be implemented within 30 days of 
the date of issuance.
    Amendment No.: 112
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45188) The February 28, 1996, and April 9, 1996 supplemental letters 
provided additional clarifying information and did not change the 
staff's original no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated June 17, 1996. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: April 15, 1996
    Brief description of amendments: These amendments would revise the 
Technical Specifications to indicate that the quadrant power tilt ratio 
requirements are applicable only at power levels greater than 50% of 
rated core power.
    Date of issuance: June 7, 1996
    Effective date: June 7, 1996
    Amendment Nos.: 210 and 210
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20860) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 7, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185
    Dated at Rockville, Maryland, this 26th day of June 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[Doc. 96-16879 Filed 7-2-96; 8:45 am]
BILLING CODE 7590-01-F