[Federal Register Volume 61, Number 119 (Wednesday, June 19, 1996)]
[Notices]
[Pages 31171-31192]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10619]



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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from May 24, 1996, through June 7, 1996. The last 
biweekly notice was published on June 5, 1996 (61 FR 28604).

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 July 19, 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

[[Page 31172]]

petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR Part 2. Interested persons should consult a current copy of 
10 CFR 2.714 which is available at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC and at 
the local public document room for the particular facility involved. If 
a request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, 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: April 25, 1996
    Description of amendment request: The proposed amendment would 
change the definition of Operable-Operability, revise Technical 
Specifications (TSs) and associated Bases Section for TSs 3.5.F.1, 
``Core and Containment Cooling systems,'' TSs 3.9.B.1, 3.9.B.2, 
3.9.B.3, 3.9.b.4, ``Auxiliary Electrical System,'' and TSs 3.7.B.1.a, 
c, and e, and 3.7.b.2.a, c, and e, ``Standby Gas Treatment System and 
Control Room High Efficiency Air Filtration System,'' and delete TSs 
4.5.F.1, ``Core and Containment Cooling Systems,'' and 3.7.B.1.f, 
``Standby Gas Treatment System and Control Room High Efficiency Air 
Filtration 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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Operation of PNPS [Pilgrim Nuclear Power Station] in accordance 
with the proposed license amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because of the following:
    Definition of ``Operable-Operability''
    Definitions perform a supporting function for other sections of 
the TS. The definition of ``Operable-Operability'' affects the 
manner

[[Page 31173]]

in which the requirements for a Limiting Condition for Operation 
(LCO) and its associated remedial actions are applied when a support 
system is inoperable. This definition re-affirms the principle that 
a system is operable when it is capable of performing its specified 
function and when all necessary support systems are also capable of 
performing their related support functions. The corollary is that a 
system is inoperable when it is not capable of performing its 
specified function or when a necessary support system is not capable 
of performing its related support function.
    No changes are being made to the plant design, system 
configuration, or method of operation. The proposed change does not 
affect the ability of the AC power sources to perform their required 
safety functions nor affect the ability of the features they support 
to perform their respective safety functions. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    EDG [Emergency Diesel Generator]
    An Individual Plant Examination (IPE) for Internal Events was 
submitted to the NRC in response to Generic Letter 88-20 in 
September 1992. The IPE was used to quantify the overall impact of 
the proposed 14 day allowed outage time on core damage frequency. 
Part III provides the results of a comprehensive Probabilistic 
Safety Assessment (PSA) of the impact of the proposed AOTs [allowed 
outage times] for the EDGs and Startup and Shutdown transformers. As 
shown in Part III, there is not a significant increase in risk due 
to the proposed change. Thus the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The existing specification 3.9.B.1 is being separated into two 
segments (a and b) because of the proposed and different AOTs for 
the Startup and Shutdown transformers. As a result of the PSA, the 
AOT for the Startup transformer (a) is reduced from 7 days to 72 
hours, while the AOT for the Shutdown transformer (b) remains at 7 
days. The reduction of the AOT from 7 days to 3 days is based on the 
relative risk importance of the Startup transformers support to the 
balance of plant systems. Similarly, an additional reduction from 72 
hours to 48 hours is proposed in the AOT for a simultaneous loss of 
both the Startup transformer and an EDG (TS 3.9.B.4.b) based upon 
the Startup transformer's contribution to risk in relation to the 
EDG 14-day AOT risk assessment analysis and that two power sources 
have been removed from the associated bus. The AOT reductions 
represent a measurable decrease in risk as assessed in the PSA. 
Thus, the probability or consequences of an accident previously 
evaluated are not significantly increased.
    The current technical specifications allow one EDG to be out of 
service for three days based on the availability of the SUT [startup 
transformer] and SDT [shutdown transformer] and the fact that each 
EDG carries sufficient engineered safeguards equipment to cover all 
design basis accidents. With one EDG out of service and a Loss of 
Offsite Power (LOOP) condition, the capability to power vital and 
auxiliary system components remains available via the other EDG, and 
for one train of ESF equipment via the SDT for all operating, 
transient and accident conditions. Increasing the EDG AOT to 14 days 
provides flexibility in the maintenance and repair of the EDGs. The 
EDG unavailability will be monitored and trended in accordance with 
the Maintenance Rule. The PSA analyses supports the change to a 14 
day AOT for the EDGs based on an insignificant increase in overall 
risk. Implementation of the proposed change is expected to result in 
less than a one percent increase in the baseline core damage 
frequency (2.84E-05/yr), which is considered to be insignificant 
relative to the underlying uncertainties involved with probabilistic 
safety assessments. Additional conditions are added to the Standby 
Liquid Control, Standby Gas Treatment, and Control Room High 
Efficiency Air Filtration systems requiring the EDG associated with 
these systems to remain operable while in the 14 day EDG AOT. Thus, 
the 14 day EDG AOT does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Eliminating the 4.5.F.1 requirement for daily testing of the 
operable diesel generator when the redundant diesel generator 
becomes inoperable is consistent with the guidance provided in 
Generic Letter 93-05. The change does not affect the ability of the 
emergency diesel generator to perform on demand, and by actually 
lowering the number of demands to demonstrate operability, reduces 
the probability of equipment failure. The redundant EDG will remain 
in service during the entire period of inoperability of the out-of-
service EDG. If a common cause failure cannot be ruled out, the 
redundant EDG will be tested to assure operability. The proposed 
revisions do not involve a significant change to the plant design or 
operation, only to the manner in which remaining equipment is 
confirmed to be operable, which is consistent with NRC guidance. 
Thus operation of PNPS in accordance with the proposed license 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The 3.9.B.1 and 2 requirements to demonstrate both EDGs and 
associated emergency buses operable are deleted. This change is 
based on the NRC guidance provided in item 10.1 of Generic Letter 
93-05, ``Line-Item Technical Specification Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation.'' 
Revising the methods for verifying EDG and emergency bus operability 
does not physically alter the plant or have an affect on the 
probability or consequences of an accident previously evaluated. 
Deleting the testing requirements for an EDG when the other EDG is 
inoperable does not increase the probability or consequences of an 
accident previously evaluated because the reliability program and 
routinely performed TS surveillances continue to provide the added 
assurance sought by the testing. The elimination of this testing 
will serve to improve the overall reliability of the EDGs. Since the 
proposed change does not affect the design or negatively affect the 
performance of the EDGs, the change will not result in a significant 
increase in the consequences or probability of an accident 
previously analyzed.
    SGT [Standby Gas Treatment] and CRHEAF [Control Room High 
Efficiency Air Filtration]
    During normal plant operation, with one SGT or CRHEAF subsystem 
inoperable, the inoperable subsystem must be restored to operable 
status in 7 days. In this condition, the remaining operable SGT or 
CRHEAF subsystem is adequate to perform the required radioactivity 
release control function. However, the overall system reliability is 
reduced because a single failure in the operable subsystem could 
result in the radioactivity release control function not being 
adequately performed. The 7 day completion time is based on 
consideration of such factors as the availability of the operable 
redundant SGT subsystem and the low probability of a DBA [design 
basis accident] occurring during this period.
    If the SGT or CRHEAF subsystem cannot be restored to operable 
status within 7 days when in the Run, Startup, or Hot Shutdown MODE, 
the plant must be brought to a MODE in which the LCO does not apply. 
To achieve this status, the plant must be brought to at least Hot 
Shutdown within 12 hours and to Cold Shutdown within 36 hours. The 
allowed completion times are reasonable, based on operating 
experience, to reach the required plant conditions from full power 
conditions in an orderly manner and without challenging plant 
systems.
    Current TS governing refueling operations restrict fuel movement 
if one train of SGTS or one train of CRHEAF are inoperable. In this 
condition the remaining operable SGT and CRHEAF trains are adequate 
to perform the required radioactivity release control functions. 
However, the overall system reliability is reduced because a single 
failure in the operable train could result in the radioactivity 
release control function of the systems not being adequately 
performed. New requirements are added that require if one train of 
SGT or CRHEAF is inoperable, the redundant train of SGT or CRHEAF 
must be demonstrated to be operable within 2 hours. This 
substantiates the availability of the operable trains. Fuel handling 
is limited only to the following 7 days and if the inoperable train 
is not returned to an operable condition within that time frame, the 
operable SGT train is placed in operation or fuel handling 
activities are suspended. For CRHEAF, after 7 days, the operable 
subsystem is demonstrated operable in accordance with existing 
surveillances on a daily basis. The proposed changes do not modify 
system design, use, or configuration in a manner different from 
their original design and therefore do not involve a significant 
increase in the consequences or probability of an accident 
previously analyzed.
    The revisions to make the SGT and CRHEAF TS sections similar in 
wording are made to enhance usability and alleviate possible 
confusion. These changes are strictly editorial, have no impact, and 
do not alter

[[Page 31174]]

technical content or meaning of the specifications. These editorial 
changes do not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The operation of PNPS in accordance with the proposed license 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated because of the 
following:
    Definition of ``Operable-Operability''
    The revised definition redefines the AC power needs to allow 
either onsite or offsite power available for systems/subsystems to 
be considered operable. This does not compromise the level of safety 
already afforded to such systems/subsystems because the functional 
operability requirements continue to be assured through the 
technical specifications applicable to such systems/subsystems. AC 
power availability continues to be assured through existing and 
proposed surveillances and action statements applicable to AC power 
systems. Reducing the need for both onsite and offsite power sources 
in order to consider operable, the systems/subsystems powered by 
these AC power sources, provides additional operational flexibility 
by allowing redundant systems/subsystems to still be considered 
``operable'' within the requirements of their functional operability 
requirements. No new change or modes of plant operation are 
involved. Therefore, operation in accordance with the revised 
definition does not introduce any new or different kind of accident 
from any accident previously evaluated.
    EDG
    The proposed amendment will extend the action completion/allowed 
outage time for an inoperable emergency diesel generator from 72 
hours to 14 days. The EDGs are designed as backup AC power sources 
for essential safety systems in the event of loss of offsite power. 
The proposed AOT does not change the conditions, operating 
configurations or minimum amount of operating equipment assumed in 
the safety analysis for accident mitigation. The EDGs and AC 
equipment are not accident initiators. No change is being made in 
the manner in which the EDG's provide plant protection. No new modes 
of plant operation are involved. An extended AOT for one EDG does 
not increase the probability of occurrence of a new or different 
kind of accident previously evaluated. The PSA results concluded 
that the risk contribution of the EDG AOT extension is 
insignificant.
    The current Pilgrim Technical Specifications requiring immediate 
and daily testing of the redundant operable EDG is based on the 
assumption that the increased testing provides additional assurance 
that the equipment is available should it be needed. Industry 
experience indicates that repetitive testing can place demands and 
wear on the EDG without necessarily providing additional confidence 
of availability. Also, the new surveillance requires verification 
that offsite power is available and that a common cause failure is 
not present. These actions provide assurance that the required 
emergency buses can be energized with no loss of functions to 
mitigate accident or transient conditions. In addition, Pilgrim has 
implemented an EDG reliability program to maintain reliability of 
EDGs. The proposed change does not introduce any new mode of plant 
operation or new accident precursors, involve any physical 
alterations to plant configurations, or make changes to system set 
points that could initiate a new or different kind of accident. 
Therefore, operation in accordance with the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    The AOT for an inoperable Startup Transformer is reduced from 7 
days to 72 hours based upon the PSA that was performed to 
quantitatively assess the risk impact of the proposed amendment. The 
proposed reduction in AOT improves overall AC power source 
availability because the SUT will potentially be inoperable for 
shorter time periods. Therefore, reducing the AOT does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    SGT and CRHEAF
    The SGT system is designed to filter radioactive materials from 
the secondary containment following a postulated DBA or fuel 
handling accident prior to release to the environment to ensure 
compliance with 10 CFR 100 limits.
    The CRHEAF is designed to filter intake air for the control room 
atmosphere during conditions when normal intake air may be 
contaminated.
    The proposed revisions do not affect the ability of the SGTS or 
CRHEAF to perform their intended function, do not create the 
possibility of a new or different kind of accident from the loss of 
coolant or fuel handling accidents previously analyzed, and do not 
modify system configuration, use, or design. Therefore, operating 
Pilgrim in accordance with this change will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    The revisions to make the SGT and CRHEAF TS sections similar in 
wording are made to enhance usability and alleviate possible 
confusion. These changes are strictly editorial, have no impact, and 
do not alter technical content or meaning of the specifications. 
These editorial changes do not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The operation of PNPS in accordance with the proposed license 
amendment will not involve a significant reduction in a margin of 
safety because of the following:
    Definition of ``Operable-Operability''
    The implementation of the ``Operability'' definition clarifies 
the relationship between AC power supplies and the operability 
status of the equipment requiring AC power. No change is being made 
in which the plant systems relied upon in the safety analyses 
provide plant protection. Plant safety margins are maintained 
through the limitations established in the TS LCOs. Since there will 
be no significant reduction to the physical design or operation of 
the plant there will be no significant reduction to any of these 
margins.
    EDG
    Operation of PNPS in accordance with the proposed license 
amendment will not involve a significant reduction in a margin of 
safety. As shown in Part III [of the application dated April 25, 
1996], incorporation of the proposed change involves an 
insignificant reduction in the margin of safety.
    The proposed changes do not significantly reduce the basis for 
any technical specification related to the establishment of, or the 
maintenance of, a safety margin nor do they require physical 
modifications to the plant. Additional conditions are added to the 
Standby Liquid Control, Standby Gas Treatment, and Control Room High 
Efficiency Air Filtration systems requiring the diesel generator 
associated with the redundant operable trains of these systems to 
remain operable while in the 14 day EDG AOT. Moreover, the PSA 
results showed that the risk contribution of extending the AOT for 
an inoperable EDG is insignificant. The reduction in the AOT for the 
SUT could improve availability, therefore, reducing overall risk. 
Likewise the proposed changes in the deletion of testing have no 
impact on the safety margin.
    As previously stated, implementation of the proposed changes is 
expected to result in an insignificant increase in: (1) power 
unavailability to the emergency buses (given that a loss of offsite 
power has occurred), and (2) core damage frequency. Implementation 
of the proposed changes does not increase the consequences of a 
previously analyzed accident nor significantly reduce a margin of 
safety. Functioning of the EDGs and the manner in which limiting 
conditions of operation are established are unaffected.
    SGT and CRHEAF
    SGT and CRHEAF contribute to the margin of safety by supporting 
the secondary containment system during fuel handling by mitigating 
the consequences of a fuel handling event. Allowing fuel movement to 
continue as established in the LCOs does not involve a significant 
reduction in the margin of safety because the first line of defense, 
the other SGT and CRHEAF trains will be operable. The proposed 
change will allow placing the Operable SGT subsystem in operation, 
or in the case of CRHEAF, conducting daily testing, as an 
alternative to suspending movement of irradiated fuel. This 
alternative is less restrictive than the existing requirement, 
however, the proposed requirements ensure that the remaining 
subsystem is operable, that no failures that could prevent actuation 
have occurred, and that any failure would be readily detected. The 
proposed change does not result in a significant reduction in a 
margin of safety because it allows operations which have the 
potential for releasing radioactive material to the secondary 
containment to continue only if the system designed to mitigate the

[[Page 31175]]

consequences of this release is functioning. Proper operation of 
only one SGT or one CRHEAF subsystem is sufficient to mitigate the 
consequences of any analyzed accident. Therefore, this change does 
not change any of the assumptions in the accident analysis and does 
not involve a significant reduction in a margin of safety.
    The revisions to make the SGT and CRHEAF TS sections similar in 
wording are made to enhance usability and alleviate possible 
confusion. These changes are strictly editorial, have no impact, and 
do not alter technical content or meaning of the specifications. 
These editorial 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: 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

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: April 22, 1996
    Description of amendment request: The licensee is proposing to 
change the technical specifications to reflect a revision to the 
overload cutoff limit on the manipulator crane inside the containment 
at the Haddam Neck Plant. Due to a change in fuel design and supplier, 
the heaviest fuel assembly design starting in Cycle 20 will be the 
Westinghouse-supplied LOPAR design. Therefore, the heaviest combination 
beginning in Cycle 20 will be the Westinghouse LOPAR fuel assembly with 
a full-length rod cluster control assembly (RCCA) inserted. It will now 
be used as the standard for the overload cutoff limit on the 
manipulator crane.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. [The proposed change does not] involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will revise the method of determining the 
overload cutoff limit for the manipulator crane. The actual absolute 
value of the cutoff limit will not be increased and will not affect 
the [probability] of any plant accidents.
    Since there is no actual increase in the absolute overload 
cutoff limit, there will be no adverse effects to the crane, cables, 
or associated hardware. Therefore, there is no impact on the crane's 
ability to perform its intended function. Even though the net 
lifting forces on an individual assembly have increased 25 pounds, 
the limit is within the recommended Westinghouse guidelines with 
respect to fuel handling and will not result in potential damage to 
assembly grids during fuel handling activities.
    As such, CYAPCO [Connecticut Yankee Atomic Power Company] has 
concluded that these changes do not involve an increase in the 
probability or consequences of an accident previously evaluated.
    2. [The proposed change does not] create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The changes conservatively revise the method of determining the 
overload cutoff limit for the manipulator crane. There is no impact 
on the basic functioning of plant systems or equipment. Therefore, 
the change does not create a malfunction that is different from 
those previously evaluated.
    As such, the proposed changes described above do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. [The proposed change does not] involve a significant 
reduction in a margin of safety.
    The proposed revisions in the methodology for determining the 
overload cutoff limit for the manipulator crane is conservative and 
in accordance with vendor standards. The changes do not adversely 
affect any equipment credited in the safety analysis. Also, the 
changes do not adversely affect the probability or consequences of 
any plant accident, including the fuel handling accident or offsite 
doses associated with those accidents.
    As such, the proposed changes have no significant impact on 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: December 14, 1995, as supplemented by 
letter dated May 16, 1996
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) to improve the TS Action 
Statements and Surveillance Requirements for diesel generators in 
accordance with the recommendations and guidance in Generic Letter 93-
05, Generic Letter 94-01, NUREG-1366, and NUREG-1431. The proposed 
amendments would also incorporate technical and administrative changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    Operation of the facilities in accordance with the requested 
amendments will not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Improvements to the LCOs [limiting condition for operation] and 
surveillance requirements for the emergency diesel generators do not 
affect their capability to provide emergency power to plant vital 
instruments and safety related equipment. In fact, these 
improvements make the diesel generators more reliable since they 
significantly reduce the amount of wear and stress due to excessive 
and unnecessary testing. The proposed monthly testing of the diesel 
generator continues to ensure that the system is ready for service 
when needed. The fast starts and fast loadings continue to ensure 
that the timing and loading requirements for engineered safety 
features actuation are met. The proposed changes do not affect any 
of the design basis accident analyses previously evaluated. 
Therefore, these proposed changes do not involve any increase in the 
probability or consequences of any accident previously evaluated. 
The proposed changes are fully consistent with the recommendations 
and guidance contained in GL [Generic Letter] 93-05, GL 94-01, 
NUREG-1366, NUREG-1431, and are compatible with plant operating 
experience.
    Criterion 2
    Operation of the facilities in accordance with the requested 
amendments will not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The 
proposed changes in fact improve the reliability of the diesel 
generators by eliminating unnecessary wear and stress. Improved 
reliability decreases the failure probability which also decreases 
the probability of an accident not previously evaluated. None of the 
requested amendments increase the common mode failure probability 
thus would not increase the chance of both EDG's [emergency diesel

[[Page 31176]]

generators] for a particular nuclear unit being out of service 
simultaneously. The proposed changes are fully consistent with the 
recommendations and guidance contained in GL 93-05, GL 94-01, NUREG-
1366, NUREG-1431, and are compatible with plant operating 
experience.
    Criterion 3
    Operation of the facilities in accordance with the requested 
amendments will not involve a significant reduction in a margin of 
safety. The proposed monthly testing of the diesel generators 
continues to ensure that the system is ready for service when 
needed. The fast starts and fast loadings continue to ensure that 
the timing and loading requirements for engineered safety features 
actuation are met. The proposed changes improve the reliability of 
the diesel generators. Implementation of the Maintenance Rule also 
ensures continued reliability of the diesel generators. No margin of 
safety is decreased as a result of these TS changes.
    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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi and Docket 
No. 40-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: April 18, 1996, as supplemented by 
letter dated June 4, 1996
    Description of amendment request: The licensee has proposed to (1) 
amend Limiting Condition for Operation (LCO) 3.10.6 and Surveillance 
Requirement 3.10.6.3, and (2) add a Surveillance Requirement 3.10.6.4 
of the Technical Specifications (TSs) for the Grand Gulf Nuclear 
Station, Unit 1, and the River Bend Station, Unit 1, to allow another 
method of fuel movement and loading in the core when control rods are 
removed or withdrawn from defueled core cells. Currently, LCO 3.10.6 
allows only fuel loading as part of the approved spiral reloading 
sequence to prevent fuel loading into core cells in which the control 
rod has been removed or withdrawn. This amendment request does not 
withdraw this approved method, revise the frequency of performing the 
surveillance during fuel loading, or alter the method of verifying the 
fuel is being loaded in compliance with the approved method. Grand Gulf 
Unit 1 and River Bend Unit 1 are both General Electric (GE) Boiling 
Water Reactor (BWR)-6 plants, the latest version of the GE design 
series.
    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:
    Entergy Operations, Inc. [(EOI)] propose[d] to change the 
current Grand Gulf Nuclear Station (GGNS) and River Bend Station 
(RBS) Technical Specifications [(TSs)]. The specific proposed change 
is to add an additional method of performing fuel loading into LCO 
3.10.6, ``Multiple Control Rod Withdrawal - Refueling''. The 
proposed change would allow fuel loading [in the core] if a positive 
means of assuring fuel assemblies cannot be loaded into a core cell 
with a withdrawn or removed control rod is in effect. [Currently, 
the TSs for both plants allow fuel assembles to be loaded in 
compliance with an approved spiral reload sequence which is used to 
ensure the reactivity additions are minimized. Spiral loadings 
encompass reloading a core cell on the edge of a continuous fueled 
region.]
    The Commission has provided standards for determining whether a 
no significant hazards consideration exists as stated in 10 CFR 
50.92(c). A proposed amendment to an operating license involves no 
significant hazards consideration if operation of the facility in 
accordance with the proposed amendment would not: (1) involve a 
significant increase in the probability or consequences of an 
accident previously evaluated; 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.
    Entergy Operations, Inc. [EOI] has evaluated the no significant 
hazards consideration in its request for this license amendment and 
determined that no significant hazards consideration results from 
this change. In accordance with 10 CFR 50.91(a), Entergy Operations, 
Inc. [EOI] is providing the analysis of the proposed amendment 
against the three standards in 10 CFR 50.92(c). A description of the 
no significant hazards consideration determination follows:
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The refueling interlocks (i.e., the refueling equipment and one-
rod-out interlocks) allowed to be bypassed by Technical 
Specification [TS] LCO 3.10.6 are explicitly assumed in the analysis 
of the control rod removal error or fuel loading error during 
refueling. This analysis evaluates the consequences of control rod 
withdrawal during refueling. Criticality and, therefore, subsequent 
prompt reactivity excursions are prevented during the insertion of 
fuel, provided all control rods are fully inserted during the fuel 
insertion. The refueling interlocks accomplish this by preventing 
loading fuel into the core with any control rod withdrawn, or by 
preventing withdrawal of a rod from the core during fuel loading.
    LCO 3.10.6 allows multiple control rod withdrawals, control rod 
removals, associated control rod drive (CRD) removal, or any 
combination of these, and the ``full in'' position indication input 
to the refueling interlocks is allowed to be bypassed for each 
withdrawn control rod if all fuel has been removed from the cell. 
This supports the GGNS Updated Final Safety Analyses Report (UFSAR) 
and RBS Updated Safety Analyses Report (USAR) analyses since, with 
no fuel assemblies in the core cell, the associated control rod has 
no reactivity control function and does not need to remain inserted. 
Prior to reloading fuel into the cell, however, the associated 
control rod must be inserted to ensure that an inadvertent 
criticality does not occur, as evaluated in the analysis.
    The Technical Specification [TS] requirements prohibiting fuel 
loading was placed in the Technical Specifications [TSs] for GGNS 
and RBS as part of the originally enforced Technical Specification 
[TS] requirements to resolve NRC concerns identified in IE 
Information Notice No. 83-35, ``Fuel Movement with Control Rods 
Withdrawn at BWRs,'' (IEN 83-35). IEN 83-35 details instances where 
fuel assemblies were loaded into core cells while the control rod 
was withdrawn and discusses that the General Electric Company (GE) 
had issued Service Information Letter (SIL) No. 372.
    SIL No. 372 discusses a potential event where 8 fuel assemblies 
are loaded into 2 [two] adjacent core cells where the control rods 
are withdrawn and no action is taken to recover from the errors. In 
this SIL GE identified that the probability of such an event 
occurring was extremely low but potentially slightly higher than 
10<SUP>-6 probability of the event even further to where it need not 
be considered credible (i.e., below 10<SUP>-6 per reactor year), GE 
recommended that the additional administrative control of 
prohibiting loading fuel with withdrawn rods be enforced.
    The proposed change will only provide an additional way to meet 
the intent of the original GE recommendation. [The currently 
approved method is listed in LCO 3.10.6 and Surveillance Requirement 
3.10.6.3.]. The proposed change will provide the additional 
allowance to perform fuel loading only if an additional positive 
means of assuring fuel assemblies cannot be loaded into a core cell 
with a withdrawn or removed control rod is in effect. The positive 
means will entail a physical barrier such that, even if refueling 
procedures were violated and an attempt was made to load a fuel 
assembly into a core cell with a withdrawn or removed control rod, 
the action would be prevented. This requirement provides sufficient 
additional restrictions to meet the intent of the GE recommendation 
to add additional administrative controls to prevent the postulated 
event from occurring.
    The probability of an inadvertent criticality occurring will 
continue to be precluded by

[[Page 31177]]

the same number of layers of administrative controls [as the 
currently approved method]; therefore, the proposed change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The administrative changes in the Technical Specification [TS] 
requirements do not involve a change in the design of the plant. The 
proposed requirements will continue to ensure that fuel is not 
loaded into a core cell that is associated with a removed or 
withdrawn control rod.
    Therefore, the proposed changes do 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 margin of safety associated with criticality events during 
fuel handling is provided by the event being a non credible event. 
The proposed change will only provide an additional means to meet 
the same intent of ensuring that the event is of such low 
probability as to be considered non credible. The proposed change 
will provide the additional allowance to perform fuel loading only 
if an additional positive means of assuring fuel assemblies cannot 
be loaded into a core cell with a withdrawn or removed control rod 
is in effect. The positive means will entail a physical barrier such 
that even if refueling procedures were violated and an attempt was 
made to load a fuel assembly into a core cell with a withdrawn or 
removed control rod the action would be prevented. This requirement 
provides sufficient additional restrictions to ensure that the event 
is of such low probability as to be considered non credible.
    The probability of an inadvertent criticality occurring will 
continue to be precluded by the same number of layers of 
administrative controls [as the currently approved method]; 
therefore, this change does not reduce the level of safety imposed 
by the current Technical Specification [TS] requirements.
    Therefore, the proposed changes do not cause 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: (1) Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120, for Grand Gulf 
Nuclear Station and (2) Government Documents Department, Louisiana 
State University, Baton Rouge, LA 70803, for River Bend Station.
    Attorney for licensee: (1) Nicholas S. Reynolds, Esquire, Winston 
and Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502, 
for Grand Gulf Nuclear Station and (2) Mark Wetterhahn, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, DC 20005, for River Bend 
Station.
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: May 9, 1996
    Description of amendment request: The amendment request would allow 
allow the licensee to perform the surveillance of the relief mode of 
operation of each of the 20 safety/relief valves (S/RVs) on the 4 main 
steam lines without physically lifting the disk off the seat at power. 
The proposed changes are to Surveillance Requirements (SRs) 3.4.4.3, 
Safety/Relief Valves, 3.5.1.7, Automatic Depressurization System 
Valves, and 3.6.1.6.1, Low-Low Set Valves, of the Technical 
Specifications, and the changes would state that the required operation 
of the valve to verify is that the relief-mode actuator strokes when 
the valve is manually actuated. Each S/RV is a Dikkers, 8 X 10, direct-
acting, spring loaded, safety valve with attached pneumatic actuator 
for relief-mode operation. Eight of the S/RVs use the relief mode to 
perform the Automatic Depressurization System (ADS) function. Also, six 
S/RVs, two of which are also ADS S/RVs, use the relief mode to perform 
the Low-Low Set valve function. The licensee also proposed changes to 
the Bases of the Technical Specifications that are associated with the 
above proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below: The Dikkers S/RV provides 
pressure relief based on the principle of vertically moving the stem 
that attaches directly to the valve disk. The force that provides the 
stem movement is provided by one of two sources; the vessel pressure 
directly against the force of the stem spring (safety mode), or the 
pneumatic actuator arm against the force of the stem spring (relief 
mode). ASME Boiler and Pressure Vessel Code requires testing the safety 
mode of operation once every five year operating cycle. Once a safety 
valve is installed, the safety mode is never tested while the S/RV is 
installed in the plant. The testing of the relief mode of operation for 
a direct-acting S/RV provides verification that the control functions 
of electrical and pneumatic connections have been properly reconnected, 
and that the actuator arm will provide the necessary force to operate 
the S/RV.
    This proposed change provides verification of proper control 
connections by requiring the pneumatic and electrical controls to 
cycle the actuator arm on each S/RV after installation in the 
drywell. The test population of S/RVs removed each outage for safety 
setpoint testing will be tested in the relief mode. This testing 
will demonstrate that the installed S/RVs will function properly in 
the relief mode. The remaining installed S/RVs will continue to be 
tested for proper system function. As presently required by GGNS 
Technical Specifications and administrative procedures, proper 
operation of the solenoid control block will be demonstrated by 
providing an open signal to each S/RV, with a check to verify that 
each solenoid valve repositions. Verification of proper solenoid 
valve operation, in addition to the proper relief-mode operation of 
the test population, provides assurance that the S/RV will perform 
as expected when control air pressure is applied to the solenoid 
valve control block.
    Entergy Operations, Inc. is proposing that the Grand Gulf 
Nuclear Station Operating License be amended to perform the 
surveillance of each safety relief valve (S/RV) relief mode of 
operation without physically lifting the disk off the seat at power.
    During the refueling outage, a sample population of the S/RVs 
will be removed for safety-mode setpoint testing in accordance with 
the GGNS IST program, using ASME Boiler and Pressure Vessel Code, 
Section XI. Each of these removed S/RVs will be tested in the relief 
mode to verify that the pneumatic actuator functions correctly, and 
this test sample will be used to provide assurance that the 
installed S/RV pneumatic actuators will function properly. After the 
test sample of S/RVs has been replaced with recertified spares, and 
S/RV controls have been connected, the upper stem nut that couples 
the valve stem to each newly- installed S/RV's pneumatic actuator 
will be moved up the stem to allow an uncoupled actuation of the 
relief-mode actuator. Control air pressure to each actuator will be 
reduced from normal system pressure to prevent damaging the 
pneumatic relief-mode actuator. The actuator will be remotely 
operated from the control room, as required by current test methods, 
and visual verification will be performed for proper actuator 
response and range of motion. After proper actuator operation has 
been verified, the upper stem nut will be returned to its operating 
stem location. Verification of proper system logic controls and 
function for every installed S/RV will continue to be performed, as 
required by Technical Specifications.
    The commission has provided standards for determining whether a 
no significant hazards consideration exists as stated in 10 CFR 
50.92(c). A proposed amendment to an operating license involves no 
significant hazards if the operation of the facility in accordance 
with the proposed amendment would not: (1) involve a significant 
increase

[[Page 31178]]

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.
    Entergy Operations has evaluated the no significant hazards 
considerations in its request for a license amendment. In accordance 
with 10 CFR 50.91(a), Entergy Operations, Inc. is providing the 
following analysis of the proposed amendment against the three 
standards in 10 CFR 50.92:
    a. No significant increase in the probability or consequences of 
an accident previously evaluated results from this change.
    Each refueling outage, a test sample of the population of S/RVs 
is removed from the plant to perform testing as required by ASME 
Boiler and Pressure Vessel Code, Section XI. These S/RVs will be 
stroked in the relief mode during as-found testing, and are 
therefore verified to operate properly when each S/RV stem is raised 
by the relief-mode pneumatic actuator. This proposed surveillance 
verifies proper S/RV relief-mode operation of all installed S/RVs 
based upon this test sample. This testing, in conjunction with 
replacement of each S/RV prior to the end of its expected service 
life, provides reasonable assurance that the installed S/RVs will 
perform as well as the test population of S/RVs.
    After the S/RVs have been replaced in the plant, and after all 
controls are reconnected, the relief-mode actuator on each newly-
installed S/RV will be uncoupled from the S/RV stem, and stroked. 
This actuator stroke will verify that no damage has occurred to the 
relief-mode actuator during S/RV transportation from its storage 
location to its operating location. The direct coupling of the valve 
stem to disk provides assurance that proper relief actuation will 
occur when the actuator is operated. The safety-mode components are 
completely encased within the valve body and bonnet, which provides 
a rugged structure to prevent damage to these components. The 
remaining installed S/RVs will continue to be tested for proper 
control system function as previously required by Technical 
Specifications. The direct coupling of the S/RV stem to disk 
provides assurance that proper relief-mode actuation will occur when 
the actuator is operated. The safety mode of the GGNS S/RVs is not 
affected by a malfunction of the relief-mode components.
    Blockage of each S/RV discharge line will be prevented by the 
same Foreign Material Exclusion (FME) controls that exist for other 
reactor vessel and support systems. These FME controls, combined 
with the horizontal orientation of the S/RV discharge piping mating 
surfaces, provide reasonable assurance that discharge line blockage 
will not occur.
    Therefore, no significant increase in the probability or 
consequences of an accident previously evaluated results from this 
proposed change.
    b. This change would not create the possibility of a new or 
different kind of accident from any previously analyzed.
    The proposed change demonstrates that each S/RV will perform its 
intended relief-mode function, which is the intent of the present 
surveillance. The relief mode of S/RV operation is demonstrated to 
be operable based upon successful performance of a test population, 
S/RV component service life, and existing Technical Specification 
surveillances. No new failure mechanisms to the relief- mode of 
operation are introduced, as the proposed surveillance verifies 
relief actuator operability. Plant FME controls, combined with the 
horizontal orientation of the S/RV discharge piping mating flange, 
provides reasonable assurance that discharge line blockage will not 
occur. This proposed change does not add any new systems, 
structures, or components, nor does it introduce new S/RV operating 
modes.
    Therefore, this change would not create the possibility of a new 
or different kind of accident from any previously analyzed.
    c. This change would not involve a significant reduction in the 
margin of safety.
    This proposed change will verify that the relief mode of all 
installed S/RVs will operate properly based upon demonstrated relief 
mode performance of a sample of S/RVs. The failure mode of the S/RV 
relief function would require a failure of either the pneumatic 
actuator, lifting linkage, or solenoid block. Each of these items 
has been verified to have a service life exceeding the replacement 
cycle of each S/RV. Therefore, proper operation of a sample 
population of S/RVs provides reasonable assurance that the remaining 
S/RVs would perform identically, within the original margin of 
expected S/RV operability. In addition, each S/RVFEs solenoid block 
and control functions will continue to be tested and cycled each 
refueling outage. The removal of the valve stroke surveillance for 
all S/RVs does not increase the possibility of valve malfunction, 
since valve stroke is verified during the as-found testing of the 
sample population of S/RVs. This proposed surveillance test reduces 
the number of S/RV actuations, and therefore, reduces challenges to 
the system both mechanically and thermally. Also, the proposed 
alternative method of testing reduces the possibility of a stuck-
open S/RV, since this proposed method will not stroke the S/RVs with 
the reactor pressurized during reactor power operations.
    Therefore, this change would not involve a significant reduction 
in the margin of safety.
    Based on the above evaluation, Entergy Operations, Inc. has 
concluded that operation in accordance with the proposed amendment 
involves no significant hazards considerations.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: May 31, 1996
    Description of amendment request: The amendment would provide an 
alternative method to compensate for inoperable refueling equipment 
interlocks. The alternative method would be to insert a control rod 
withdrawal block and verify that all control rods are fully inserted; 
however, the control rods required to be inserted would not apply to 
those control rods withdrawn in accordance with LCO 3.10.6, ``Multiple 
Control Rod Withdrawal -Refueling.'' The amendment would add an 
additional Required Action for Limiting Condition for Operation (LCO) 
3.9.1, ``Refueling Equipment Interlocks,'' of the Technical 
Specifications (TSs) for Grand Gulf Nuclear Station, Unit 1 (GGNS). The 
alternative method then could be used to respond to inoperable 
interlocks instead of only the current method of halting in-vessel fuel 
movement with equipment associated with the inoperable interlock.
    The proposed change does not remove the current Required Action 
method for LCO 3.9.1 and does not change the surveillance requirements 
on the refueling equipment. The licensee has also provided changes to 
the Bases of the TSs for the proposed amendment.
    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. The licensee has proposed the amendment for the TSs for 
both GGNS and River Bend Station (RBS). References made to the RBS TSs 
and to RBS in the licensee's analysis of no significant hazards 
consideration have been removed and replaced by [...]. The licensee's 
analysis is presented below:
    Entergy Operations, Inc. proposes to change the current Grand 
Gulf Nuclear Station (GGNS) [...] Technical Specifications. The 
specific proposed change adds additional acceptable Required Actions 
to the Actions of LCO 3.9.1, ``Refueling Equipment Interlocks,'' 
[for inoperable interlocks]. The additional Required Actions will 
add an alternative [method] to [the current method of] suspending 
fuel movement in the reactor vessel when the refueling interlocks 
are inoperable. The requested alternative is to insert a control rod 
withdrawal block

[[Page 31179]]

immediately and verify all control rods required to be inserted are 
fully inserted. [The control rods required to be inserted would not 
apply to control rods withdrawn in accordance with LCO 3.10.6, 
``Multiple Control Rod Withdrawal--Refueling.'']
    The Commission has provided standards for determining whether a 
no significant hazards consideration exists as stated in 10 CFR 
50.92(c). A proposed amendment to an operating license involves no 
significant hazards consideration if operation of the facility in 
accordance with the proposed amendment would not: (1) involve a 
significant increase in the probability or consequences of an 
accident previously evaluated; 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.
    Entergy Operations, Inc. has evaluated the [criteria for] no 
significant hazards consideration in its request for this license 
amendment and determined that no significant hazards consideration 
results from this change. In accordance with 10 CFR 50.91(a), 
Entergy Operations, Inc. is providing the analysis of the proposed 
amendment against the three standards in 10 CFR 50.92(c). A 
description of the no significant hazards consideration 
determination follows:
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The refueling interlocks are explicitly assumed in the GGNS 
Updated Final Safety Analyses Report (UFSAR) [...] analysis of the 
control rod removal error or fuel loading error during refueling. 
This analysis evaluates the probability and consequences of control 
rod withdrawal during refueling. Criticality and, therefore, 
subsequent prompt reactivity excursions are prevented during the 
insertion of fuel, provided all control rods are fully inserted 
during the fuel insertion. The refueling interlocks accomplish this 
by preventing loading fuel into the core with any control rod 
withdrawn, or by preventing withdrawal of a rod from the core during 
fuel loading.
    When the refueling interlocks are inoperable the current method 
of preventing the insertion of fuel when a control rod is withdrawn 
is to prevent fuel movement. This method is currently required by 
the Technical Specifications. An alternate method to ensure that 
fuel is not loaded into a cell with the control rod withdrawn is to 
prevent control rods from being withdrawn and verify that all 
control rods required to be inserted are fully inserted. The 
proposed actions will require that a control rod block be placed in 
effect thereby ensuring that control rods are not subsequently 
inappropriately withdrawn. Additionally, following placing the 
control rod withdrawal block in effect, the proposed actions will 
require that all required control rods be verified to be fully 
inserted. This verification is in addition to the requirements to 
periodically verify control rod position by other Technical 
Specification requirements. These proposed actions will ensure that 
control rods are not withdrawn and cannot be inappropriately 
withdrawn because an electrical or hydraulic block to control rod 
withdrawal is in place. Like the current requirements the proposed 
actions will ensure that unacceptable operations are blocked (e.g., 
loading fuel into a cell with a control rod withdrawn [would be 
blocked]).
    The proposed additional acceptable Required Actions provide the 
same level of assurance that fuel will not be loaded into a core 
cell with a control rod withdrawn as the current Required Action or 
the Technical Specification Surveillance Requirement.
    Therefore, the proposed change does not significantly increase 
the probability or consequences of an accident previously evaluated.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change in the Technical Specification requirements does not 
involve a change in plant design. The proposed requirements will 
continue to ensure that fuel is not loaded into the core when a 
control rod is withdrawn except following the requirements of LCO 
3.10.6, ``Multiple Control Rod Removal--Refueling,'' which is 
unaffected by this change.
    Therefore, the proposed changes do 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.
    As discussed in the Bases for the affected Technical 
Specification requirements, inadvertent criticality is prevented 
during the insertion of fuel provided all control rods are fully 
inserted during the fuel insertion. The refueling interlocks 
function to support the refueling procedures by preventing control 
rod withdrawal during fuel movement and the inadvertent loading of 
fuel when a control rod is withdrawn.
    The proposed change will allow the refueling interlocks to be 
inoperable and fuel movement to continue only if a control rod 
withdrawal block is in effect and all required control rods are 
verified to be fully inserted. These proposed Required Actions 
provide the same level of protection as the refueling interlocks by 
preventing a configuration which could lead to an inadvertent 
criticality event. The refueling procedures will continue to be 
supported by the proposed required actions because control rods 
cannot be withdrawn and as a result fuel cannot be inadvertently 
loaded when a control rod is withdrawn.
    Therefore, the proposed changes do not cause 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: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: May 31, 1996, as supplemented by letter 
dated May 2, 1996.
    Description of amendment request: The amendment request would 
revise the current reactor vessel material surveillance program 
schedule for GGNS. This is the schedule for withdrawing surveillance 
capsules from the reactor vessel for testing to measure the impact of 
neutron irradiation of the vessel material and is required by Section 
III.B.3 of Appendix H, ``Reactor Vessel Material Surveillance Program 
Requirements,'' of 10 CFR Part 50. The schedule must be approved by the 
Nuclear Regulatory Commission (NRC) before implementation.
    For GGNS, there are three surveillance capsules inside the reactor 
vessel, each of which contains specimens of the reactor vessel 
material. The first capsule was removed from the reactor vessel on May 
7, 1995, during the 7th refueling outage. Because no useful data is 
expected from testing the material specimens in the first capsule, the 
request would allow the first capsule to be placed back into the 
vessel.
    As part of revising the schedule, the licensee is also renumbering 
the three surveillance capsules so that the capsule removed at the 7th 
refueling outage becomes the third capsule when it is placed back in 
the vessel. The proposed change would, however, not extend the time 
that the next capsule (the renumbered first capsule) would be withdrawn 
from the GGNS reactor vessel.
    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:
    Entergy Operations, Inc., proposes to change the withdrawal 
schedule for the reactor vessel material surveillance capsules [and 
renumber the capsules]. The revised schedule for withdrawal of the 
surveillance capsules is withdrawal of the first capsule at 24 
Effective Full Power Years. The withdrawal schedule for the second 
capsule is to be determined at a later date. The third capsule which 
was withdrawn on May 7, 1995 is to be returned to reactor vessel 
during

[[Page 31180]]

the Fall, 1996 outage and retained as a standby. [The current 
schedule for withdrawal of the three capsules is 8 and 24 Effective 
Full Power Years for the first two capsules, and the third capsule 
is a spare with no specific schedule for withdrawal.]
    The Commission has provided standards for determining whether a 
no significant hazards consideration exists as stated in 10 CFR 
50.92(c). A proposed amendment to an operating license involves no 
significant hazards consideration if operation of the facility in 
accordance with the proposed amendment would not: (1) involve a 
significant increase in the probability or consequences of an 
accident previously evaluated; 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.
    In consideration of the October 4, 1995, decision of the Atomic 
Safety and Licensing Board concerning an amendment request from 
Perry Nuclear Power Plant, Entergy Operations, Inc. has evaluated 
the no significant hazards consideration in its request for a change 
to the withdrawal schedule required by 10 CFR 50, Appendix H, and 
determined that no significant hazards consideration results from 
this change. In accordance with 10 CFR 50.91(a), Entergy Operations, 
Inc. is providing the analysis of the proposed amendment against the 
three standards in 10 CFR 50.92(c):
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The change revises the withdrawal schedule for the reactor 
vessel material surveillance capsules and returns a withdrawn 
capsule to the reactor vessel. The capsules [only contain specimens 
of the reactor vessel material and] are not an initiator of any 
previously analyzed accident. The withdrawal or return of the 
surveillance capsule does not effect the probability or consequences 
of any previously analyzed accident. Extending the time for 
withdrawal of the first capsule and returning the withdrawn capsule 
to the vessel do not adversely affect the pressure temperature limit 
curves for the reactor vessel. Regulatory Guide 1.99 [, ``Effects of 
Residual Elements on Predicted Radiation Damage to Reactor Vessel 
Materials,''] is currently used to prepare the pressure temperature 
limit curves and is inherently conservative for boiling water 
reactors (BWRs)[, as GGNS]. The current pressure temperature limit 
curves will continue to be adhered to. Additionally, [GGNS] 
participates in the supplemental test program designed to 
significantly increase the amount of BWR surveillance data. [This 
program has supplemental capsules which were installed in the Cooper 
and Oyster Creek Nuclear Power Plants, which contain the limiting 
GGNS weld and plate vessel material, and which will be withdrawn in 
1996, 2000, and 2002.] This program will be used to complement the 
GGNS surveillance program such that postponement of the capsule 
withdrawals will have minimal impact on the understanding of the 
irradiation effects on the GGNS vessel.
    [The licensee stated in its May 2, 1996, letter that testing of 
the specimens in the removed capsule may not provide useful 
indicators of the damage to the vessel material because the low 
neutron fluence on the vessel and the good material chemistry will 
result in a minimal null-ductility temperature shift. Testing the 
material specimens will destroy them; however, placing the capsule 
back in the vessel will allow the specimens to have more irradiation 
until useful data could be obtained from testing the specimens.]
    Therefore, the proposed change does not significantly increase 
the probability or consequences of an accident previously evaluated.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Returning the withdrawn capsule to the vessel and postponing the 
withdrawal of the first capsule do not contribute to the possibility 
of a new or different kind of accident or [plant] malfunction from 
those previously analyzed [in the Updated Final Safety Analysis 
Report for GGNS]. Failure of the reactor vessel is not a credible 
accident since the vessel itself is a highly reliable component. 
This change does not affect that determination. The potential for 
reactor vessel cracking will be adequately assessed by the proposed 
withdrawal schedule.
    [The licensee stated in its May 2, 1996, letter that testing of 
the specimens in the removed capsule may not be useful indicators of 
the damage to the vessel material because the low neutron fluence on 
the vessel and good material chemistry will result in a minimal 
shift.]
    In addition, the results from the supplemental test program will 
provide indication of the condition of the vessel until the data 
from the first GGNS capsule[, withdrawn and tested,] are available. 
The proposed change provides the same level of confidence in the 
integrity of the vessel. The pressure temperature curves are 
currently controlled by the Technical Specifications and are 
determined using the conservative methodology in Regulatory Guide 
1.99. Therefore, the possibility of failure of the reactor vessel is 
not increased. The proposed change does not involve a change in the 
design of the plant. The current pressure temperature limit curves 
are inherently conservative and will continue to be adhered to.
    Therefore, the proposed changes do 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 current pressure temperature limit curves [for the reactor 
vessel] are inherently conservative and provide sufficient margin to 
ensure the integrity of the reactor vessel. The [proposed] changes 
do not adversely affect these curves. The supplemental test program 
will be used to complement the GGNS surveillance program such that 
postponement of the capsule withdrawal [and testing] will have 
minimal impact on the understanding of irradiation effects on the 
GGNS vessel. The capsules removed in 1996 as part of the 
supplemental program will have a [neutron] fluence higher than the 
25% of the design life fluence used in establishing the original 
GGNS [reactor vessel material surveillance program] schedule; 
therefore, the use of the supplemental test program results will 
meet the intent of the original test schedule.
    Therefore, the proposed changes do not result in a significant 
reduction in the margin of safety.
    Based on the above evaluation, Entergy Operations, Inc. has 
concluded that operation in accordance with the proposed change 
involves no 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: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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: March 21, 1996, and May 13, 1996
    Description of amendment request: The licensee proposed to change 
the Turkey Point Units 3 and 4 Technical Specifications (TS) to 
relocate the requirements of the Radiological Effluent Technical 
Specifications (RETS) to other documents.
    The proposed amendments would relocate the LIMITING CONDITIONS FOR 
OPERATION (LCO) and SURVEILLANCE REQUIREMENTS associated with the RETS 
in accordance with GL 89-01, NUREG-1301, and NUREG-1431, Rev. 1. The 
definition in TS 1.15, ``Members of the Public,'' would be deleted 
since it is already located in 10 CFR Part 20 and has been inserted 
into the Offsite Dose Calculation Manual (ODCM). The definitions for 
the ODCM and Process Control Program (PCP) would be relocated to the 
Administrative Controls section of the TS. TS 3/4.3.3.5 and the 
radioactive gaseous effluent portion of TS 3/4.3.3.6 and associated 
tables, instrumentation operational conditions, remedial actions and 
surveillance requirements would be controlled through the ODCM or PCP 
and associated procedures. Technical

[[Page 31181]]

Specification Administrative Control sections would contain the 
programmatic controls for the ODCM and PCP. The remaining portion of TS 
3.3.3.6 would retain the operational conditions, remedial actions, and 
surveillance requirements for the explosive gas monitor 
instrumentation.
    The procedural details of the current TS on radioactive effluents 
and radiological environmental monitoring would be deleted. Associated 
operational conditions, remedial actions and surveillance requirements 
presently in the Technical Specifications would be controlled through 
the ODCM or PCP.
    Administrative changes to the TS were also proposed due to 
paragraph and section numbering changes and relocations associated with 
the proposed technical changes.
    New sections TS 6.8.4f and 6.8.4g were proposed to provide 
programmatic controls for the Radiological Effluents Controls Program 
and the Radiological Environmental Monitoring Program.
    TS 6.9.1.3 and TS 6.9.1.4 would be simplified and the reporting 
details now contained in these specifications would be relocated to the 
ODCM or PCP with the exception of the requirement to report licensee-
initiated changes to the PCP in the Annual Radioactive Effluent Release 
Report.
    New record retention requirements changes for the ODCM and PCP 
would be added to TS 6.10.3q.
    In summary, as provided in the guidance, the current technical 
content of the specifications which would be transferred to the ODCM or 
the PCP. New programmatic controls for radioactive effluents and 
radioactive effluent monitoring would be added to the TS, as well as 
further clarification to the definitions of the ODCM and PCP. The 
Technical Specification requirements for Gas Decay Tanks and Explosive 
Gas Mixture would be relocated to the Plant Systems section 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.
    (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 changes being proposed are administrative in nature in that 
they relocate Technical Specification requirements associated with 
RETS from the Technical Specifications to the ODCM or PCP. These 
changes are in accordance with the recommendations contained in GL 
89-01, NUREG 1301, and NUREG 1431 Rev. 1. The only change being made 
to existing requirements or commitments are administrative in 
nature. The proposed changes do not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident, nor do they affect 
any assumptions or conditions in any of the accident analyses. Since 
the accident analyses remain bounding, their probability or 
consequences are not adversely affected. Therefore, the probability 
or consequences of an accident previously evaluated are not 
affected.
    (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 changes being proposed are administrative in nature in that 
they relocate Technical Specification requirements associated with 
RETS from the Technical Specifications to the ODCM or PCP. These 
changes are in accordance with the recommendations contained in GL 
89-01, NUREG 1301, and NUREG 1431, Rev. 1. The only change being 
made to existing requirements or commitments are administrative in 
nature. The proposed changes do not involve any change to the 
configuration or method of operation of any plant equipment used to 
mitigate the consequences of an accident.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be 
created.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The changes being proposed are administrative in nature in that 
they relocate Technical Specification requirements associated with 
RETS from the Technical Specifications to the ODCM or PCP. These 
changes are in accordance with the recommendations contained in GL 
89-01, NUREG 1301, and NUREG 1431, Rev. 1. The only change being 
made to existing requirements or commitments are administrative in 
nature. All technical content is preserved. The operating limits and 
functional capabilities of the affected systems, structures, and 
components are unchanged by the proposed amendments.
    Therefore, a significant reduction in a margin of safety would 
not be involved.
    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: May 28, 1996
    Description of amendment request: The licensee proposed to change 
the Turkey Point Units 3 and 4 Technical Specifications (TS) to change 
the licensed qualifications of the Operations Manager. The proposed 
change would delete the qualification option that the Operations Manger 
could have held a Senior Reactor Operator License on a boiling water 
reactor and replace it with an option that this individual could have 
completed the Turkey Point Nuclear Plant Senior Management Operation 
Training Course (i.e., certified at an appropriate simulator for 
equivalent senior operator knowledge level).
    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 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The change being proposed is administrative in nature, addresses 
organizational personnel qualification issues, and does not affect 
assumptions contained in plant safety analyses, the physical design 
and/or operation of the plant, or Technical Specifications that 
preserve safety analysis assumptions.
    The individual Florida Power & Light Company (FPL) chooses to 
fill the position of Operations Manager will have extensive 
educational and management- level nuclear power experience meeting 
the criteria of ANSI N18.1-1971. The Operations Supervisor and 
Nuclear Plant Supervisors maintain SRO licenses on Turkey Point. The 
current Technical Specifications do not require the Operations 
Manager to hold an SRO License at Turkey Point. The current 
Technical Specifications permit the Operations Manager to have held 
an SRO License on another plant. The proposed change will continue 
to require that the Operations Manager has completed the Turkey 
Point Nuclear Plant Senior Management Operations Training Course if 
the incumbent did not previously hold an SRO license. The Turkey 
Point Nuclear Plant Senior Management Operations Training Course 
ensures that the Operations Manager has the training on plant-
specific systems

[[Page 31182]]

and procedures at Turkey Point and a knowledge level equivalent to 
the license requirements for operations management.
    The on-shift Operations' organization is, and will continue to 
be, supervised and directed by the Operations Supervisor, who is 
currently required by Technical Specification 6.2.2.h. to hold an 
SRO License.
    Additionally, the proposed changes do not impact or change, in 
any way, the minimum on-shift manning or qualifications for those 
individuals responsible for the actual licensed operation of the 
facility as required by 10 CFR 50.54(l).
    Based on the above, the proposed changes do not affect the 
probability or consequences of accidents previously analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The change being proposed is administrative in nature, addresses 
personnel qualification issues, does not affect assumptions 
contained in plant safety analyses, the physical design and/or 
operation of the plant, or Technical Specifications that preserve 
safety analysis assumptions.
    The proposed changes address organizational and qualifications 
issues related to the criteria used for assignment of individuals to 
the Operations organization off-shift management chain of command. 
Since the proposed change does not impact or change, in any way, the 
minimum on-shift manning or qualifications for those individuals 
responsible for the actual licensed operation of the facility, 
operation of the facility in accordance with the proposed amendment 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed change addresses organizational and qualification 
issues related to the criteria used for assignment of individuals to 
the Operations organization off-shift management chain of command. 
The proposed change does not impact or change, in any way, the 
minimum on-shift manning or qualifications for those individuals 
responsible for the actual licensed operation of the facility.
    FPL's operating organization at Turkey Point Plant is shown on 
Figure 1-2, Appendix A of the NRC-approved FPL Topical Quality 
Assurance Report (TQAR). Since changes to the TQAR are governed by 
10 CFR Sec. 50.54(a)(3), any changes to the TQAR that reduce 
commitments previously accepted by the NRC require approval by the 
NRC prior to implementation.
    While the Operations Manager is responsible for the plant's 
operating organization, his responsibilities also include management 
of the plant's Health Physics and Chemistry departments. The 
Operations organization is supervised and directed by the Operations 
Supervisor, who is required by Technical Specification 6.2.2.h. to 
hold a Senior Reactor Operator License. The Turkey Point Units 3 and 
4 Technical Specifications do not require that the Operations 
Manager maintain an SRO License (nor even that the incumbent has 
ever held a Senior Reactor Operator License at Turkey Point). The 
Turkey Point Technical Specification 6.3.1, FACILITY STAFF 
QUALIFICATIONS, will ensure that, other than license certification, 
the individual filling the Operations Manager position has the 
requisite education, training, and experience for the management 
position.
    As a result, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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

GPU Nuclear Corporation and Saxton Nuclear Experimental 
Corporation, Docket No. 50-146, Saxton Nuclear Experimental 
Facility (SNEF), Bedford County, Pennsylvania

    Date of amendment request: February 2, 1996, as supplemented on 
February 28, April 24 and May 24, 1996.
    Description of amendment request: The proposed amendment would (1) 
increase the scope of work permitted within the exclusion area at the 
SNEF to include action preparatory to major component and facility 
decommissioning limited to asbestos removal, removal of defunct plant 
electrical services, and installation of decommissioning support 
facilities and systems such as heating, ventilation, and air 
conditioning,
    (2) eliminate administrative access controls requiring that the 
grating covering the auxiliary compartment stairwell and rod room 
door remain locked except for authorized entry, and (3) revise the 
facility layout diagram to allow the exclusion area to consist of, 
at a minimum, the containment vessel, and at a maximum, extend to 
the SNEF outer security fence, and to include on the diagram the 
footprint of the proposed decommissioning support facilities.
    Basis for proposed no significant hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes do not involve a significant hazards 
considerations because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The SNEF ended power operation in May 1972, and the reactor core 
has been removed. In its present condition, the only accidents 
applicable to the site are fire, flooding, and radiological hazard. 
The additional activities associated with the expansion of the 
permissible work scope will not involve a significant increase in 
the probability or consequences of a fire. There is no effect on the 
probability or consequences of flooding nor would there be a 
significant increase in the probability or consequences of an 
offsite radiological hazard. The relocation of administratively 
controlled accesses in accordance with the revised wording and the 
proposed clarification of the facility layout diagram would have no 
affect on analyzed accidents. Activities associated with the 
construction of the decommissioning support facilities and the 
existence of the completed buildings depicted on the revised figure 
will not involve a significant increase in the probability or 
consequences of a fire, flood, or radiological hazard. The proposed 
changes identified by this technical specification change request do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    For the reasons discussed in 1 above, the possibility of a new 
or different kind of accident from any accident previously evaluated 
will not be created by the performance of the activities delineated 
in the proposed revised technical specifications. There is similarly 
no possibility of a new or different kind of accident from any 
accident previously evaluated that would result from relocation of 
administratively controlled accesses within the containment vessel; 
from the flexibility to relocate/modify the exclusion area fence or 
from the identification of the footprint, construction and existence 
of the completed decommissioning support facilities.
    3. Involve a significant reduction in a margin of safety.
    For the reasons discussed in 1 above, none of the proposed 
changes involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the analysis of the licensees and, based 
on this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Saxton Community Library, 911 
Church Street, Saxton, Pennsylvania 16678 Attorney for the Licensee: 
Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge, 
2300 N Street, NW, Washington, DC 20037

[[Page 31183]]

    NRC Project Director: Seymour H. Weiss

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 20, 1996
    Description of amendment request: The proposed amendment would 
revise the Facility Operating License No. NPF-47 and Appendix C to the 
license to reflect the name change from Gulf States Utilities Company 
to Entergy Gulf States, Inc.
    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 significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change documents changing the legal name of the 
company. The proposed change will not affect any other obligations. 
The company will still own all of the same assets, serve the same 
customers, and all existing obligations and commitments will 
continue to be honored.
    Therefore, the proposed change does no significantly increase 
the probability or consequences of an accident previously evaluated.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The administrative changes in the Operating License requirements 
do not involve any change in the design of the plant.
    Therefore, the proposed changes do 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 proposed change is administrative in nature, as described 
above, therefore, this change does not reduce the level of safety 
imposed by any current requirements.
    Therefore, the proposed changes do not cause 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: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: April 25, 1996
    Description of amendment request: The change modifies the 
calibration requirement for the source range monitors and intermediate 
range monitors by noting that the sensors are excluded.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10 CFR 50.92, NNECO has reviewed the proposed change 
and concludes that the change does not involve a significant hazards 
consideration (SHC) since the proposed change satisfies the criteria 
in 10 CFR 50.92(c). That is, the proposed change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    By removing the requirement for sensor calibration the function 
and safety performance of these systems will not be affected. 
Existing surveillances, operator verification of overlap and system 
interlocks ensure correct system performance without sensor 
calibration.
    Therefore, based on the above, the proposed change to the 
Technical Specifications does not involve a significant increase in 
the probability or consequences of any previously analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This change does not cause the source range monitors (SRM) or 
the intermediate range monitors (IRM) to function any differently 
than intended by design and, therefore, does not create the 
possibility of a new or different kind of accident. The Technical 
Specification change deletes a Technical Specification requirement 
which could not literally be complied with for one component and 
that has no effect on the functional performance of the SRMs or 
IRMs.
    Therefore, this change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    This change corrects a Technical Specification requirement which 
could not literally be complied with for one component and that has 
no effect on the functional performance of the SRMs or IRMs. 
Instrument calibrations and functional checks are still performed 
during each refueling outage to assure adequate system performance.
    Therefore, this change has no impact on the margin to safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and Waterford Library, ATTN: Vince Juliano, 49 Rope 
Ferry Road, Waterford, CT 06385.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment requests: February 14, 1996
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant (DCPP), Unit Nos. 1 and 2, to revise 30 TS and add two new 
TS surveillance requirements to support implementation of extended fuel 
cycles at DCPP, Unit Nos. 1 and 2. The specific TS changes proposed 
include those for 9 trip actuating device tests, 12 fluid system 
actuation tests, and 11 miscellaneous tests. Two of the fluid system 
actuation tests are proposed new TS surveillance requirements. The TS 
changes also include the addition of a new frequency notation, ``R24, 
REFUELING INTERVAL,'' to Table 1.1 of the TS. Also, a revision that 
applies to all subsequent TS changes involves revising the Bases 
section of TS 4.0.2 to change the surveillance frequency from an 18-
month surveillance interval to at least once each refueling interval.
    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

[[Page 31184]]

consequences of an accident previously evaluated.
    The surveillance interval notation addition in TS Table 1.1 and 
the updated TS 4.0.2 Bases section are administrative changes that 
do not affect the probability or consequences of accidents.
    The 30 proposed TS surveillance interval increases from 18 to 24 
months do not alter the intent or method by which the inspections, 
tests, or verifications are conducted, do not alter the way any 
structure, system, or component functions, and do not change the 
manner in which the plant is operated. The surveillance, 
maintenance, and operating histories indicate that the equipment 
will continue to perform satisfactorily with longer surveillance 
intervals. Few surveillance and maintenance problems were 
identified. No problems recurred, with the exception of those 
associated with the pressurizer heater emergency breakers, which 
will continue to be surveilled on a quarterly frequency until they 
are replaced.
    There are no known mechanisms that would significantly degrade 
the performance of the evaluated equipment during normal plant 
operation. All potential time-related degradation mechanisms have 
insignificant effects in the timeframe of interest (24 months +25 
percent, or 30 months). Based on the past performance of the 
equipment, the probability or consequences of accidents would not be 
significantly affected by the proposed surveillance interval 
increases.
    The 24-month surveillance intervals for the two new TS proposed 
to verify that the CCW [component cooling water] and ASW [auxiliary 
saltwater] pumps will start automatically are based on an evaluation 
of historical operation, maintenance, and surveillance data for the 
pumps. These historical data are available because the pumps have 
been operated, maintained, and tested on 18- month intervals in 
accordance with procedures since initial plant startup. These new 
surveillances represent additional TS requirements to ensure the CCW 
and ASW pumps start when required. No known degradation mechanisms 
would significantly affect the ability of the pumps to start over 
the timeframe of interest (30 months maximum). Based on the past 
performance of the equipment, these proposed new TS would not affect 
the probability or consequences of accidents.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The surveillance interval notation addition in TS Table 1.1 and 
the updated TS 4.0.2 Bases section are administrative changes that 
do not affect the type of accidents possible.
    For the 30 proposed TS changes involving surveillance interval 
increases from 18 to 24 months, the surveillance and maintenance 
histories indicate that the equipment will continue to effectively 
perform its design function over the longer operating cycles. 
Additionally, the increased surveillance intervals do not result in 
any physical modifications, affect safety function performance or 
the manner in which the plant is operated, or alter the intent or 
method by which surveillance tests are performed. Only a few 
problems have been identified and generally have not recurred. All 
potential time-related degradations have insignificant effects in 
the timeframe of interest. The proposed surveillance interval 
increases would not affect the type of accidents possible.
    The 24-month surveillance intervals for the two new TS proposed 
to verify starting of the CCW and ASW pumps are based on an 
evaluation of historical operation, maintenance, and surveillance 
data. These new TS represent additional requirements to ensure the 
CCW and ASW pumps start when required. No known degradation 
mechanisms would significantly affect the ability of the pumps to 
start over the timeframe of interest. These proposed new TS would 
not affect the type of accidents possible.
    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 change does not involve a significant reduction 
in a margin of safety.
    The surveillance interval notation addition in TS Table 1.1 and 
the updated TS 4.0.2 Bases section are administrative changes that 
do not affect the margin of safety.
    For the 30 proposed TS changes involving 18- to 24-month 
surveillance interval increases, evaluation of historical 
surveillance and maintenance data indicates there have been only a 
few problems experienced with the evaluated equipment.
    There are no indications that potential problems would be cycle-
length dependent or that potential degradation would be significant 
for the timeframe of interest and, therefore, increasing the 
surveillance interval will have little, if any, impact on safety. 
There is no safety analysis impact since these changes will have no 
effect on any safety limit, protection system setpoint, or limiting 
condition for operation, and there are no hardware changes that 
would impact existing safety analysis acceptance criteria. Safety 
margins would not be significantly affected by the proposed 
surveillance interval increases.
    As previously noted, the 24-month surveillance intervals for the 
two new TS are based on an evaluation of historical data, represent 
additional requirements, and are not believed to be significantly 
affected by potential time-dependent degradation. As such, these 
proposed new TS would not affect any margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

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

    Date of amendment requests: May 9, 1996
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 by revising Technical Specifications (TS) 
3/4.3.2, ``Engineered Safety Features Actuation System 
Instrumentation,'' and 3/4.6.2, ``Containment Spray System.'' The 
changes would clarify the description of the initiation signal required 
for operation of the containment spray system at Diablo Canyon Power 
Plant (DCPP) and correctly incorporate changes made in previous license 
amendments. All of the changes are administrative in nature.
    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.
    Revising the description of the containment spray (CS) 
initiating signal clarifies the design of the plant and provides 
uniformity across the Technical Specifications (TS) associated with 
the CS initiation function. The enhanced description does not affect 
system operation or performance, nor the probability of any event 
initiators. The changes do not affect any engineered safety feature 
actuation setpoints or accident mitigation capabilities.
    The administrative changes to TS 3/4.3.2, Table 4.3-2, correct 
the column headings and restore test frequency notation. The changes 
only revise the TS to correspond with previously issued license 
amendments (LAs).
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

[[Page 31185]]

    The administrative changes in the description of the CS 
initiating signal provide uniformity across the TS associated with 
the spray system. There are no design, operation, maintenance, or 
testing changes associated with the administrative changes.
    The administrative changes to TS 3/4.3.2, Table 4.3-2, correct 
the column headings and restore test frequency notation. The changes 
only revise the TS to correspond with previously issued LAs.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The administrative changes in CS signal description are not 
associated with any design, operation, maintenance, or testing 
revisions.
    The administrative changes to TS 3/4.3.2, Table 4.3-2, correct 
the column headings and restore test frequency notation. The changes 
only revise the TS to correspond with previously issued LAs.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

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

    Date of amendment request: May 20, 1996 (TS 373)
    Description of amendment request: The proposed amendment revises 
the technical specifications to incorporate a 24-hour delay in 
implementing the action requirements due to a missed surveillance 
requirement when the action requirements provide a restoration time 
that is less than 24 hours. This change also clarifies that the time 
limit of the action requirements applies from the point in time it is 
identified a surveillance has not been performed and not at the time 
that the allowed surveillance interval was exceeded. The licensee 
claims this amendment is consistent with generic guidance.
    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:
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment to TS definition 1.0.LL is in accordance 
with the guidance of GL 87-09 and NUREG 1433, Revision 1. The 
proposed change will allow BFN to continue operation for an 
additional 24 hours after discovery of a missed surveillance. The 
change being proposed does not affect the precursor for any accident 
or transient analyzed in Chapter 14 of the BFN Updated Final Safety 
Analysis Report. The proposed change does not reflect a revision to 
the physical design and/or operation of the plant. Therefore, 
operation of the facility in accordance with the proposed change 
does not affect the probability or consequences of an accident 
previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to TS definition 1.0.LL is in accordance 
with the guidance of GL 87-09 and NUREG 1433, Revision 1. The 
proposed change will allow the plant to continue operation for an 
additional 24 hours after discovery of a missed surveillance. The 
change being proposed will not change the physical plant or the 
modes of operation defined in the facility license. The change does 
not involve the addition or modification of equipment, nor do they 
alter the design or operation of plant systems. Therefore, operation 
of the facility in accordance with the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment to TS definition 1.0.LL is in accordance 
with the guidance of GL 87-09 and NUREG 1433, Revision 1. The 
proposed change does not affect plant safety analysis or change the 
physical design or operation of the plant. The proposed change will 
allow the plant up to 24 hours to perform a missed surveillance. The 
overall effect is a net gain in plant safety by avoiding unnecessary 
shutdowns and the associate system transients due to missed 
surveillance. Therefore, operation of the facility in accordance 
with the proposed change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: May 8, 1996
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
5.3, ``Reactor,'' and TS 5.4, ``Fuel Storage,'' by removing the 
enrichment limit for reload fuel and imposing fuel storage restrictions 
on the spent fuel storage racks and the new fuel storage racks. The 
revised TS are structured consistent with the Westinghouse Standard 
Technical Specifications and the fuel storage restrictions are based on 
the criticality analyses used to support TS Amendment 92 dated March 7, 
1991.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    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 criticality analysis which was performed in support of 
Technical Specification Amendment 92, dated March 7, 1991, 
demonstrated that adequate margins to criticality can be maintained 
with fuel enrichments up to 49.2 grams of U<SUP>235 per axial 
centimeter stored in the New Fuel Storage Racks and enrichments up 
to 52.3 grams of U<SUP>235 per axial centimeter stored in the Spent 
Fuel Storage Racks.
    The bounding cases of the analysis demonstrated that k<INF>eff 
remains less than 0.95 in the Spent Fuel Storage Racks and the New 
Fuel Storage Racks if flooded with unborated water. The bounding 
cases of the analysis also demonstrated that k<INF>eff remains less 
than 0.98 in the New Fuel Storage Racks if moderated by optimally 
misted moderator. Therefore, the 49.2 grams of U<SUP>235 per axial 
centimeter enrichment is acceptable for storage in the New Fuel 
Storage Racks and 52.3 grams of U<SUP>235 per axial centimeter for 
storage in the Spent Fuel Storage Racks.
    The only other accident that needs to be considered is a fuel 
handling accident. Since the mass of the fuel assembly would not be 
appreciably altered by the increased fuel

[[Page 31186]]

enrichment, the probability of this accident occurring is not 
changed. The consequences of a fuel handling accident also would not 
be affected by the use of higher fuel enrichment since the fission 
product inventories in a fuel assembly are not a significant 
function of initial fuel enrichment. This accident was analyzed in 
the criticality analysis which was performed in support of Technical 
Specification Amendment 92, dated March 7, 1991.
    It should be noted that any changes in the nuclear properties of 
the reactor core that may result from higher fuel enrichments would 
be analyzed in the appropriate reload analysis.
    The administrative relocation of information to licensee 
controlled documents (i.e., USAR) conforms to NRC policy for the 
content of technical specifications and does not increase the 
probability or consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    As discussed above, the only safety issue significantly affected 
by the proposed change is the criticality analysis of the Spent Fuel 
Storage Racks and the New Fuel Storage Racks. Since it has been 
demonstrated that k<SUP>G<approx-eq><SUP>2eff remains below 0.95 and 
0.98, respectively, in those areas, no new or different accident 
would be created through the use of fuel enrichments up to 52.3 
grams of U<SUP>235 per axial centimeter at the Kewaunee Nuclear 
Power Plant. Administrative controls will ensure that only fuel 
enriched to 49.2 grams of U<SUP>235 per axial centimeter or less 
will be placed into the New Fuel Storage Racks.
    The relocation of information to licensee controlled documents 
does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    Since the criticality analyses have shown that increasing the 
allowable weight percent enrichment to 52.3 grams of U<SUP>235 per 
axial centimeter would not increase k<INF>eff above 0.95 in the 
Spent Fuel Storage Racks and increasing the allowable weight percent 
enrichment to 49.2 grams of U<SUP>235 per axial centimeter would not 
increase k<INF>eff above 0.98 in the New Fuel Storage Racks, it is 
concluded that this proposed change would not reduce the margin of 
safety. Any changes in the nuclear properties of the reactor core 
that may result from higher fuel enrichments would be analyzed in 
the appropriate reload analysis to ensure compliance with applicable 
reload considerations and requirements.
    Relocation of information to licensee controlled documents is an 
administrative action and therefore does not reduce 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

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

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

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

    Date of amendment request: May 17, 1996
    Description of amendment request: The proposed amendments would 
modify Technical Specification Section 3/4.4.5, Steam Generators, 3/
4.4.6, Reactor Coolant System Leakage, and associate Bases to allow the 
installation of tube sleeves as an alternative to plugging to repair 
defective steam generator tubes.
    Date of individual notice in the Federal Register: May 29, 1996 (61 
FR 26936)
    Expiration date of individual notice: June 28, 1996
    Local Public Document Room location:  Wharton County Junior 
College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 
77488 Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington
    Date of application for amendment: April 24, 1996
    Brief description of amendment request: The proposed amendment 
would modify Technical Specifications (TSs) 5.3.1 and 6.9.3.2 to 
reflect use of new fuel obtained from ABB/Combustion Engineering, and 
to incorporate staff-approved core reload analysis computer programs 
(codes). Date of individual notice in Federal Register: May 1, 1996 (61 
FR 19326)
    Expiration date of individual notice: May 31, 1996
    Local Public Document Room location:  Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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.

[[Page 31187]]

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

    Date of application for amendments: January 5, 1996, as 
supplemented by letters dated April 19, May 1, and May 10, 1996.
    Brief description of amendments: The amendments revise the 
operating licenses and Technical Specification (TS) Section 1.26 to 
increase the authorized rated thermal power. The amendments also revise 
TS 4.1.1.4, 3.1.3.4, and 3.2.6 (Figure 3.2-1) to lower the allowable 
reactor coolant system cold leg temperature limits for each of the 
three Palo Verde Nuclear Generating Station units, and TS 3.4.2.1 and 
3.4.2.2 to lower the pressurizer safety valve setpoints for Units 1 and 
3 to support the increased power operation. The Unit 2 pressurizer 
safety valve setpoints in TS 3.4.2.1 and 3.4.2.2 were revised in 
Amendment 78, approved March 28, 1995, to the same values being 
requested for Units 1 and 3 in this submittal.
    Date of issuance: May 23, 1996
    Effective date: May 23, 1996, to be implemented for Unit 1 within 
30 days of issuance; to be implemented for Unit 2 within 30 days of 
issuance; to be implemented for Unit 3 within 45 days as of the date of 
issuance, except for the pressurizer safety valve setpoints change 
which are effective prior to startup from Unit 3's sixth refueling 
outage.
    Amendment Nos.: Unit 1 - 108; Unit 2 - 100; Unit 3 - 80
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7544) The April 19, May 1, and May 10, 1996, supplemental letters 
provided additional clarifying information and did not change the 
initial no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated May 23, 1996. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

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

    Date of application for amendment: January 31, 1996.
    Brief description of amendment: This amendment revises the 
Technical Specifications Section 4.4 to allow the use of 10 CFR Part 
50, Appendix J, Option B, Performance-Based Containment Leakage Rate 
Testing.
    Date of issuance: May 28, 1996
    Effective date: May 28, 1996
    Amendment No. 169
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7545) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 28, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550

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: November 15, 1995, as 
supplemented by letters dated March 15, and April 10, 1996
    Brief description of amendments: The amendments revise the 
Technical Specifications and the associated Bases to increase the 
setpoint tolerance of the main steam safety valves (MSSVs) from plus or 
minus 1% to plus or minus 3%, to incorporate a requirement to reset the 
as-left MSSV lift settings to within plus or minus 1% following 
surveillance testing, and to delete two obsolete footnotes.
    Date of issuance: May 31, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 146 and 140
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65676). The March 15 and April 10, 1996 letters provided clarifying 
information that did not change the scope of the November 15, 1995 
application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated May 31, 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

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: January 12, 1995, as 
supplemented by letter dated June 29, 1995
    Brief description of amendments: The amendments revise and clarify 
portions of Technical Specification Section 6.0, ``Administrative 
Controls.''
    Date of issuance: May 30, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 145 and 139
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1995 (60 
FR 58109) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 30, 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

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: April 3, 1996
    Brief description of amendments: The amendments revise the 
Technical Specifications and the associated Bases to provide that if 
neither Train A or Train B of the hydrogen igniter is operable in any 
one containment region, there is an allowance of 7 days to restore one 
hydrogen igniter to operable status, or be in hot shutdown within the 
next 6 hours.
    Date of issuance: June 3, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 147 and 141
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 16, 1996 (61 FR 
16649) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 3, 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

[[Page 31188]]

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

    Date of amendment request: May 19, 1995, as supplemented by letter 
dated December 7, 1995
    Brief description of amendment: The amendment revised the 
recombiner surveillance requirements to conform with the staff guidance 
provided in NUREG-1432, ``Standard Technical Specifications Combustion 
Engineering Plants.''
    Date of issuance: June 5, 1996
    Effective date: June 5, 1996
    Amendment No.: 119
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
180) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 5, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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

    Date of application for amendments: January 4, 1996
    Brief description of amendments: These amendments rectify a 
discrepancy in Technical Specification 3.5.3, and provide assurance 
that administrative controls for High Pressure Safety Injection pumps 
remain effective in the lower operational modes.
    Date of Issuance: May 30, 1996
    Effective Date: May 30, 1996
    Amendment Nos.: 143 and 183
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5813) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 30, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of application for amendments: November 22, 1995
    Brief description of amendments: These amendments upgrade existing 
TS 3/4.4.6.1 for the Reactor Coolant System Leakage Detection Systems 
by adopting the Standard Technical Specifications for Combustion 
Engineering Plants to both St. Lucie Units.
    Date of Issuance: May 30, 1996
    Effective Date: May 30, 1996
    Amendment Nos.: 144 and 84
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1629) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 30, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of application for amendment: March 28, 1996 (TSCR 234)
    Brief description of amendment: The amendment modifies Technical 
Specification pages 3.1-5 and 3.1-16 to indicate 40 percent of the 
rated reactor thermal power as the anticipatory reactor scram bypass 
setpoint on turbine trip or generator load rejection.
    Date of Issuance: June 4, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 184
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18167) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated June 4, 1996 No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753

Houston Lighting & Power Company, City Public Service Board of San 
Antonio Central Power and Light Company, City of Austin, Texas, 
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County, 
Texas

    Date of amendment request: January 22, 1996, as supplemented April 
4 and May 2, 1996
    Brief description of amendment: The amendment modified the steam 
generator tube plugging criteria in TS 3/4.4.5, Steam Generators, the 
allowable primary-to-secondary leakage in TS 3/4.4.6.2, Operational 
Leakage, and the associated Bases. These changes allowed the 
implementation of alternate steam generator tube plugging criteria for 
the tube support plate/tube intersections for Unit 1.
    Date of issuance: May 22, 1996
    Effective date: May 22, 1996
    Amendment No.: 83
    Facility Operating License No. NPF-76. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 16, 1996 (61 FR 
16651) as corrected April 22, 1996 (61 FR 17735). The additional 
information contained in the supplemental letter dated May 2, 1996, was 
clarifying in nature and thus, within the scope of the initial notice 
and did not affect the staff's proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated May 22, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

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

    Date of application for amendment: July 21, 1995, as supplemented 
August 8, 1995 and December 15, 1995
    Brief description of amendment: The amendment made administrative 
changes to various sections of the DAEC Technical Specifications (TS). 
The amendment replaced the surveillance condition when an Emergency 
Service Water pump or loop is inoperable with an OPERABILITY 
verification of the opposite train's Emergency Diesel Generator (EDG). 
The amendment modified the TS to allow credit for demonstration of EDG 
OPERABILITY that occurred within the previous 24 hours. The amendment 
revised the format and language of TS Section 5.5

[[Page 31189]]

to clarify the requirements and state the capacity of the spent fuel 
pool and vault storage in order to remove ambiguities in the wording 
and to be more consistent with the Improved Standard TS guidance. The 
amendment revised the list of Operations Committee responsibilities 
(Section 6.5.1.6) to eliminate Committee review of procedures 
implementing Security and Emergency Plans.
    Date of issuance: June 5, 1996
    Effective date: June 5, 1996
    Amendment No.: 214
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49938) and February 2, 1996 (61 FR 3953) The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
June 5, 1996. No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401

Northern States Power Company, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County, Minnesota

    Date of application for amendments: May 4, 1995, as supplemented 
November 27, 1995, and March 1, 1996
    Brief description of amendments: The amendments revise the 
pressurizer and main steam safety valve lift setting tolerance from 
plus or minus 1 percent to plus or minus 3 percent (as-found setpoint 
only), revise the safety limit curves, reformat Section 2, and correct 
typographical errors.
    Date of issuance: May 21, 1996 Effective date: May 21, 1996, with 
full implementation within 30 days
    Amendment Nos.: Unit 1 - 123, Unit 2 - 116
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47621) The November 27, 1995, and March 1, 1996, letters provided 
clarifying information in response to NRC staff questions. This 
information was within the scope of the original application and did 
not change the staff's initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated May 21, 1996. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

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

    Date of application for amendments: March 13, 1996
    Brief description of amendments: These amendments delete the 
requirement in Technical Specifications (TS) 4.0.5a for NRC written 
approval prior to implementation of relief from ASME Code requirements 
by deleting ``...(g),.except where specific written relief has been 
granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).'' Also, 
the amendments add the ASME Section XI definition of ``Biennially or 
every 2 years - At least once per 731 days,'' in TS 4.0.5b.
    Date of issuance: May 28, 1996
    Effective date: May 28, 1996
    Amendment Nos.: Unit 1 - 112; Unit 2 - 110
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18173) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 28, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendments: April 3, 1996
    Brief description of amendments: These amendments revise the 
combined Technical Specifications (TS) for the Diablo Canyon Nuclear 
Power Plant, Unit Nos. 1 and 2 to revise Technical Specifications 3/
4.7.5, ``Control Room Ventilation System;'' 3/4.7.6, ``Auxiliary 
Building Safeguards Air Filtration System;'' and 3/4.9.12, ``Fuel 
Handling Building Ventilation System'' to clarify the testing 
methodology utilized by PG&E to determine the operability of the 
charcoal and high efficiency particulate air (HEPA) filters in the 
engineering safeguards features (ESF) air handling units at the Diablo 
Canyon Power Plant (DCPP).
    Date of issuance: May 28, 1996
    Effective date: May 28, 1996
    Amendment Nos.: Unit 1 - 113; Unit 2 - 111
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18173) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 28, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: May 8, 1996, as supplemented May 
10, 1996, and May 29, 1996, and June 3, 1996.
    Brief description of amendment: This amendment modifies the 
Technical Specifications to correct several typographical errors that 
were implemented in the Improved Technical Specifications at Ginna 
Station per Amendment No. 61.
    Date of issuance: June 3, 1996
    Effective date: As of date of issuance.
    Amendment No.: 65
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (61 FR 24965, dated May 17, 
1996). That notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice published May 
17, 1996, also provided for a hearing by June 17, 1996, but indicated 
that if a Commission makes a final no significant hazards consideration 
determination, any such hearing would take place after issuance of the 
amendment. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 3, 1996.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

[[Page 31190]]

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

    Date of application for amendment: February 9, 1996 as superseded 
by letter dated March 22, 1996.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 1.7, 4.6.1.1, 3.6.1.3, 4.6.1.3, 6.8.4 and the 
associated Bases section to directly reference Regulatory Guide 1.163, 
``Performance-Based Containment Leak Test Program,'' as required by 10 
CFR 50, Appendix J, Option B for the Type A containment integrated leak 
rate tests and the Type B and C local leak tests.
    Date of issuance: May 28, 1996
    Effective date: May 28, 1996, to be implemented within 30 days from 
the date of issuance.
    Amendment No.: 111
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18174) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 28, 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, et al., Docket Nos. 50-338 and 
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa 
County, Virginia

    Date of application for amendments: January 30, 1996
    Brief description of amendments: The amendments modify the 
Technical Specifications to increase the minimal allowable reactor 
coolant system total flow rate.
    Date of issuance: June 5, 1996
    Effective date: June 5, 1996
    Amendment Nos.: 201 and 182
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7559) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 5, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of application for amendment: April 24, as supplemented by 
letter dated May 29, 1996.
    Brief description of amendment: The amendment would modify the WNP-
2 technical specifications to support Cycle 12 operation, reflect use 
of new fuel obtained from ABB/Combustion Engineering, and incorporate 
staff-approved core reload analysis computer programs (codes). Date of 
issuance: June 4, 1996 Effective date: June 4, 1996, to be implemented 
within 30 days of issuance.
    Amendment No.: 146
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 1, 1996 (61 FR 
19326). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 4, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has

[[Page 31191]]

made a determination based on that assessment, it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By July 19, 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.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, 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 20555-001, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket No. 50-249, Dresden Nuclear 
Power Station, Unit No. 3

    Date of application for amendment: May 22, 1996
    Brief description of amendment: The amendment authorizes, on a one- 
time temporary basis, operation of Dresden, Unit 3, with the structural 
steel members in the Low Pressure Coolant Injection (LPCI) corner rooms 
outside the Updated Final Safety Analysis Report (UFSAR) design 
parameters, but capable of performing their intended safety function. 
Following a reactor scram on May 15, 1996, Commonwealth Edison Company 
(ComEd) performed a Safety Evaluation (SE) in accordance with the 
requirements of 10 CFR 50.59 to determine if the current configuration 
of the corner room structural steel members had reduced the margin of 
safety as described in the UFSAR. The SE determined that the 
configuration does not reduce the margin of safety with respect to the 
stress allowables for the structural steel if subjected to a Safe 
Shutdown Earthquake (SSE). An unreviewed safety question was determined 
to exist because stress allowables for the structural steel subjected 
to an Operating Basis Earthquake (OBE) were found outside the UFSAR 
requirements; however, the current configuration of the corner room 
structural steel members has not

[[Page 31192]]

significantly reduced the margin of safety as described in the UFSAR.
    Date of Issuance: May 31, 1996 Effective date: May 31, 1996
    Amendment No.: 144
    Facility Operating License No. DPR-25. The amendment revised the 
license.
    Press release issued requesting comments as to proposed no 
significant hazards consideration: Yes. Joliet Herald News on May 25, 
1996, and the Morris Daily Herald on May 29, 1996. Comments received: 
No comments were received on the proposed no significant hazards 
consideration determination; however, comments were received concerning 
the licensee's timeliness and decision-making in restoring the UFSAR 
design margin to the structural steel members installed the LPCI corner 
rooms at Dresden Unit 3.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the State of Illinois and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated May 31, 1996.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.
    NRC Project Director: Robert A. Capra
    Dated at Rockville, Maryland, this 12th day of June 1996.
    For the Nuclear Regulatory Commission
John A. Zwolinski,
Deputy Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[Doc. 96-15398 Filed 6-18-96; 8:45 am]
BILLING CODE 7590-01-F