[Federal Register Volume 62, Number 10 (Wednesday, January 15, 1997)]
[Notices]
[Pages 2185-2197]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10115]


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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 December 20, 1996, through January 3, 1997. 
The last biweekly notice was published on January 2, 1997 (62 FR 121).

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 February 14, 1997, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.

[[Page 2186]]

    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.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of amendment request: November 4, 1996, as supplemented on 
December 4, 1996.
    Description of amendment request: The proposed amendment would 
permit Byron, Unit 1, and Braidwood, Unit 1, to remove sheathing filler 
grease in the tendon sheathing for up to 35 tendons in advance of the 
steam generator replacement outages.
    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 prestressing tendons are passive components that form part 
of the containment structure. As passive components, there are no 
tendon failure modes that could act as accident initiators or 
precursors.
    Consequently, the proposed change to remove a portion of the 
tendon sheathing filler grease will not increase the probability of 
an accident previously evaluated.
    The tendons, in their passive role, function to limit the 
consequences of accidents previously evaluated, and their continued 
integrity is important to the ability of the containment to mitigate 
design basis accidents. Structural degradation of the containment is 
a predictable process that can be monitored by a comprehensive 
containment tendon monitoring program as required by Technical 
Specification Surveillance Requirement 4.6.1.6. The monitoring 
program is based on proposed Revision 3 of Regulatory Guide 1.35, 
``Inservice Surveillance of Ungrouted Tendons in Prestressed 
Concrete Containment Structures,'' April 1979.
    The tendon surveillances conducted at both Byron and Braidwood 
have consistently shown that structural integrity of the tendon 
system has been maintained, including adequate corrosion protection 
for the tendon wires and end anchorage components, and there has 
been no evidence of grease leakage from the tendon sheathings. While 
a number of below-grade hoop tendons have shown signs of water 
intrusion, the tendons that will have grease removed are above-grade 
and are not expected to experience water intrusion.
    A review of domestic nuclear facility experience found cases 
where large grease voids existed for periods longer than requested 
under the proposed change without resultant corrosion in those 
tendon systems. A case where tendon wires removed from a 
decommissioned plant were exposed to an environment more severe than 
expected in a sealed tendon sheath did not show signs of corrosion. 
These experiences demonstrate the effectiveness of the initial 
corrosion protection systems applied to the tendons and the 
effectiveness of partial grease protection in the tendon sheathing.
    Based on the above cases, it can be concluded that the removal 
of the filler grease (grease voids greater than 5 percent) from the 
tendon sheathing in up to thirty-five tendons for a limited period 
will not adversely affect the integrity of the tendons or the 
capability of the tendon system to fulfill its design basis 
function.
    The removal process will only remove the grease not directly 
adhering to the tendons. The grease remaining will be adequate to 
protect the tendons during the relatively short period of partial 
grease removal. Therefore, no changes in the tendon properties would 
be expected, and the consequences of design basis accidents 
previously evaluated will not be affected by the proposed change.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

[[Page 2187]]

    The proposed change only affects the tendon sheathing filler 
grease void limits of TSSR 4.6.1.6. No new equipment is being 
installed and no existing equipment is being modified. Operation 
with a grease void in excess of current requirements does not alter 
system configurations such that any new or different accidents can 
be initiated. Therefore, no new or different accident initiators or 
precursors are being introduced, and the proposed change will not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety applicable to the proposed change is 
defined by the difference between the design pressure of the 
containment and the point at which the containment would actually 
fail. The design pressure of the containment is 50 psi. As a result 
of conservatism inherent in the design techniques and in the 
material selections made for the Byron and Braidwood containments, a 
substantial margin to failure exists in the containment. This margin 
is discussed in Subsection 3.8.1.8 of the Updated Final Safety 
Analysis Report. It is noted therein that the ultimate capacity of 
the concrete shell is 125 psi, corresponding to the initiation of 
yield in the hoop post-tensioning tendons in conjunction with 
yielding of the reinforcement near the mid-height of the containment 
wall.
    It is also noted in Subsection 3.8.1.8 that the ultimate 
capacity of a containment electrical penetration is 108 psi. While 
this value is substantially greater than the 50 psi required of the 
design, it is lower than the 125 psi at which failure of the 
containment wall section would be predicted. Therefore, tendon 
strength is not the limiting factor in the margin of safety inherent 
in the containment.
    As previously discussed, no degradation of the tendons is 
expected to occur as a result of the proposed TS change. Further, 
the tendon strength is not the limiting factor in the containment 
ultimate capacity, which is substantially greater than the 
requirement placed on the containment design by the plant design 
basis. Therefore, the proposed change will not reduce the margin of 
safety designed into Byron and Braidwood.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois

    Date of application for amendment request: December 6, 1996
    Description of amendment request: The proposed amendment would 
allow a single control rod to be moved when the plant is in HOT 
SHUTDOWN and COLD SHUTDOWN condition provided the one-rod-out interlock 
is OPERABLE and the reactor mode switch is in the refuel position.
    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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    This revision would allow a single control rod to be withdrawn 
under control of the reactor mode switch position one-rod-out 
interlock in OPERATIONAL MODES 3 or 4. This interlock is explicitly 
assumed in the safety analysis for control rod removal error during 
refueling. A prompt reactivity excursion could potentially result in 
fuel failure. The one-rod-out interlock, together with the 
requirements for adequate SHUTDOWN MARGIN (SDM), provides protection 
against prompt reactivity excursions by preventing withdrawal of 
more than one control rod and ensuring the core remains subcritical 
with any one control rod withdrawn. The addition of surveillance 
requirements for the one-rod-out interlock will assure the interlock 
is OPERABLE prior to withdrawal of a control rod in OPERATIONAL 
MODES 3 and 4. Although this change will increase the frequency of 
single control rod withdrawals in OPERATIONAL MODES 3 and 4, the 
probability of previously analyzed accidents, including control rod 
withdrawal error, is not affected because the same actions are 
required, although they are now conducted in different OPERATIONAL 
MODES.
    The consequences of previously analyzed accidents in OPERATIONAL 
MODES 3 and 4 are not affected by this proposed change. The SDM 
requirements of TS 3.3.A assure the reactor is maintained 
subcritical when all control rods are fully inserted, without 
crediting the single control rod having the highest reactivity worth 
which is assumed to be fully withdrawn. The one-rod-out interlock of 
the reactor mode switch Refuel position permits only a single 
control rod to be withdrawn. The proposed change will not affect the 
potential for attaining criticality in OPERATIONAL MODES 3 and 4 or 
effect the initial conditions assumed in any design basis accident 
analysis.
    Based on this, the probability or consequences of any accident 
previously evaluated is not increased by the proposed changes.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    Single control rods can be withdrawn to permit control rod 
recoupling in OPERATIONAL MODES 3 and 4 under existing TS. The 
proposed change will merely expand this allowance to other control 
rod maintenance and testing activities performed in OPERATIONAL 
MODES 3 and 4. The revision to Specification 3/4.10.A provides 
additional assurance that the one-rod-out interlock is OPERABLE in 
OPERATIONAL MODES 3 and 4.
    The additional control rod maintenance and testing activities 
which could be performed in OPERATIONAL MODES 3 and 4 are permitted 
by the existing TS in OPERATIONAL MODES 1, 2 and 5. Examples of 
activities which could be performed include venting of control rods 
following a reactor scram or control rod drive system outage, normal 
control rod insertion/withdrawal timing and adjustment, control rod 
scram time testing and control rod friction testing.
    Based on this, the proposed changes do not create the 
possibility of a new or different kind of accident from those 
previously evaluated.
    Specification 3/4.10.A is revised to ensure the one-rod-out 
interlock is OPERABLE, enhancing the assurance that the plant will 
prevent the withdrawal of more than one control rod in the manner 
currently assumed. Expanding the applicability of this existing 
requirement to OPERATIONAL MODES 3 and 4 similarly does not create 
the possibility of a new or different kind of accident from those 
previously evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    The TS currently permit single control rod withdrawal for the 
purpose of control rod recoupling when in OPERATIONAL MODES 3 or 4 
if the one-rod-out interlock is OPERABLE. This change merely allows 
additional activities for which a single control rod may be 
withdrawn in OPERATIONAL MODES 3 or 4, with the same restriction 
that the one-rod-out interlock is OPERABLE.
    While the TS currently allow limited control rod withdrawal in 
OPERATIONAL MODES 3 and 4 provided the one-rod-out interlock is 
OPERABLE, no explicit surveillance requirements for the one-rod-out 
interlock exist while in OPERATIONAL MODES 3 or 4. The proposed 
changes to the Applicability statement in TS 3/4.10.A will result in 
applicability of the Surveillance Requirements for the one-rod-out 
interlock whenever control rod withdrawal is performed in 
OPERATIONAL MODES 3 and 4.
    Together, the OPERABILITY requirements for the one-rod-out 
interlock and the SDM requirements of TS 3.3.A will continue to 
ensure that the reactor will be maintained subcritical during single 
control rod withdrawals. Therefore, this change will not involve a 
significant reduction in the margin of safety.

[[Page 2188]]

    As described, the proposed amendment for Dresden and Quad Cities 
Stations will not reduce the availability of systems required to 
mitigate accident conditions. Neither are new or significantly 
different modes of operation proposed. Therefore, the proposed 
changes do not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: October 2, 1996
    Description of amendment request: Relocation of Radiological 
Effluent Technical Specifications for Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1 - Does not Involve a Significant Increase in the 
probability or Consequences of an Accident Previously Evaluated.
    The proposed changes are considered administrative in nature. 
These changes alter only the location of programmatic controls and 
procedural details relative to radioactive effluents, radiological 
environmental monitoring, solid radioactive wastes, and associated 
reporting requirements. Compliance with applicable regulatory 
requirements will continue to be maintained. In addition, the 
proposed changes do not alter the conditions and assumptions in any 
of the Safety Analysis Report (SAR) accident analyses. Since the SAR 
accident analyses remain bounding, the radiological consequences 
previously evaluated are not adversely affected by the proposed 
changes.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed changes do not involve any changes to the 
configuration or method of operation any plant equipment. The 
proposed changes are considered administrative in nature. 
Accordingly, no new failure modes have been defined for any plant 
system or component important to safety nor has any new limiting 
single failure have been identified as a result of the proposed 
changes. Also, there will be no change in types or increase in the 
amounts of any radioactive effluents released offsite.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in Margin 
of Safety.
    The proposed changes do not involve nay actual change in the 
methodology used in the control of radioactive effluents, solid 
radioactive wastes, or radiological environmental monitoring. These 
changes are considered administrative in nature and provide for the 
relocation of procedural details outside the Technical 
Specifications. This change adds appropriate administrative controls 
in the Technical Specifications to provide continued assurance of 
compliance with applicable regulatory requirements.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. I21Therefore, based upon the reasoning 
presented above and the previous discussion of the amendment 
request, Entergy Operations has determined that the requested change 
does not involve 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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: October 2, 1996
    Description of amendment request: Relocation of Selected Technical 
Specifications Instrumentation Requirements Allowed by Generic Letter 
95-10
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1 - Does not Involve a Significant Increase in the 
probability or Consequences of an Accident Previously Evaluated.
    The [Nuclear Regulatory Commission] NRC issued Generic Letter 
(GL) 95-10 to allow licensees to relocate certain instrumentation 
requirements to licensee controlled documents or programs. The staff 
has concluded that the specifications listed in the GL were not 
required to be included in the technical specifications as required 
by 10 CFR 50.36. The staff concluded that the instrumentation 
addressed in these specifications are not related to dominant 
contributors to plant risk.
    The specifications included in this amendment request are being 
relocated to the Technical Requirements Manual (TRM). Once in the 
TRM, future changes to these requirements will be controlled under 
10 CFR 50.59. By controlling future changes under 10 CFR 50.59, NRC 
review and approval will be requested for changes exceeding the 
regulatory threshold of an unreviewed safety question.
    This amendment request does not remove or modify any of the 
instrumentation requirements for either unit. This amendment request 
does not affect any of the accident initiators, conditions or 
assumptions for any of the accidents previously evaluated. 
Therefore, this change does not involve a significant increase in 
the probability of any accident previously evaluated.
    This amendment request is administrative in nature and does not 
affect any system or component functional requirements. This change 
does not affect the operation of the plant or affect any component 
that is used to mitigate the consequences of any accident. 
Therefore, this change does not involve a significant increase in 
the consequences of any accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The relocation of existing requirements from the technical 
specifications to other licensee controlled documents is considered 
administrative in nature. This change does not modify or remove any 
plant instrumentation requirements. This proposed change will not 
affect any plant system or structure, nor will it affect any system 
functional or operability requirements. Consequently, no new failure 
modes are introduced as a result of this change. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in Margin 
of Safety.
    The proposed amendment request represents a relocation of a 
portion of the information previously located in each unit's 
technical specification instrumentation section to other licensee 
controlled documents that ate controlled under 10 CFR 50.59. The 
proposed change is administrative in nature because the 
instrumentation requirements for the facility remain the same.

[[Page 2189]]

The proposed change does not represent a change in the configuration 
or operation of the plant. Therefore, this change does not involve a 
significant reduction in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve 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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: December 2, 1996
    Description of amendment request: The proposed Technical 
Specification (TS) Change Request will permit the use of 10 CFR Part 50 
Appendix J, Option B, Performance-Based Containment Leakage Testing for 
Type A, B and C leak rate testing. TSs 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, 
4.6.1.6 and 4.6.1.7 are revised and Section 6.15 is added establishing 
the Containment Leakage Rate Testing Program. The Bases are revised to 
reflect this change. Minor editorial changes are included in this 
request. Waterford Steam Electric Station is planning to have a 
Containment Leakage Rate Testing Program in place prior to the next 
scheduled refueling outage. This program will be in accordance with the 
guidelines contained in Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Test Program,'' dated September 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change will not affect the assumptions, design 
parameters, or results of any accident previously evaluated. The 
proposed change does not add or modify any existing equipment. The 
proposed changes will result in increased intervals between 
containment leakage tests determined through a performance based 
approach. The intervals between such tests are not related to 
conditions which cause accidents. The proposed changes do not 
involve a change to the plant design or operation. Therefore, this 
change does not involve a significant increase in the probability of 
any accident previously evaluated.
    NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' 
contributed to the technical bases for Option B of 10 CFR 50 
Appendix J. NUREG-1493 contains a detailed evaluation of the 
expected leakage from containment and the associated consequences. 
The increased risk due to lengthening of the intervals between 
containment leakage tests was also evaluated and found acceptable. 
Using a statistical approach, NUREG-1493 determined the increase in 
the expected dose to the public from extending the testing frequency 
is extremely small. It also concluded that a small increase is 
justifiable due to the benefits which accrue from the interval 
extension. The primary benefit is in the reduction in occupational 
exposure. The reduction in the occupational exposure is a real 
reduction, while the small increase to the public is statistically 
derived using conservative assumptions. Therefore, this change does 
not involve a significant increase in the consequences of any 
accident previously evaluated.
    The proposed change does not involve modifications to any 
existing equipment. The proposed change will not affect the 
operation of the plant or the manner in which the plant is operated. 
The reduced testing frequency will not affect the testing 
methodology. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not change the performance methodology 
of the containment leakage rate testing program. However, the 
proposed change does affect the frequency of containment leakage 
rate testing. With an increased frequency between tests, the 
proposed change does increase the probability that a increase in 
leakage could go undetected for a longer period of time. Operational 
experience has demonstrated the leak tightness of the containment 
buildings has been significantly below the allowable leakage limit.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to containment 
leakage rates. The limitation on containment leakage rate is 
designed to ensure the total leakage volume will not exceed the 
value assumed in our accident analysis. The margin of safety for the 
offsite dose consequences of postulated accidents directly related 
to containment leakage is maintained by meeting the 1.0 La 
acceptance criteria. The proposed change maintains the 1.0 La 
acceptance criteria. Therefore, the proposed change will not involve 
a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Dates of amendment request: December 9, 1996
    Description of amendment request: The licensee proposed to modify 
specifications for selected cycle-specific reactor physics parameters 
to refer to the St. Lucie Unit 1 Core Operating Limits Report (COLR) 
for limiting values. Minor administrative changes are also included. 
The proposed Technical Specification (TS) changes utilized the guidance 
provided in Generic Letter 88-16 and are intended to be consistent with 
the Standard Technical Specifications for Combustion Engineering Plants 
(NUREG-1432, Revision 1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment relocates the calculated values of 
selected cycle-specific reactor physics parameter limits from the TS 
to the COLR, and includes minor editorial changes which do not alter 
the intent of stated requirements. The amendment is administrative 
in nature and has no impact on any plant configuration or system 
performance relied upon to mitigate the
    consequences of an accident. Parameter limits specified in the 
COLR for this amendment are not changed from the values presently 
required by Technical Specifications. Future changes to the 
calculated values of such limits may only be made using NRC approved 
methodologies, must be consistent with all applicable safety 
analysis limits, and are controlled by the 10 CFR 50.59 process. 
Assumptions used for accident initiators and/or safety analysis 
acceptance criteria are not changed by this amendment. Therefore, 
operation of the facility in accordance with the proposed amendment 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

[[Page 2190]]

    (2) Operation of the facility in accordance with the proposed
    amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment relocates the calculated values of cycle 
specific reactor physics limiting parameters to the COLR and will 
not change the physical plant or the modes of operation defined in 
the facility license. The changes do not involve the addition of new 
equipment or the modification of existing equipment, nor do they 
alter the design configuration of St. Lucie plant systems. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed

    amendment would not involve a significant reduction in a margin 
of safety.
    The cycle specific parameter limits being relocated to the COLR 
by this amendment have not been changed from the values presently 
required by the TS, and a requirement to operate the plant within 
the bounds of the limits specified in the COLR is retained in the 
individual specifications. Future changes to the calculated values 
of these limits by the licensee may only be developed using NRC-
approved methodologies, must remain consistent with all plant safety 
analysis limits addressed in the Final Safety Analysis Report 
(FSAR), and are further controlled by the 10 CFR 50.59 process. As 
discussed in Generic Letter 88-16, the administrative controls 
established for the values of cycle specific parameters using the 
guidance of that letter assure conformance with 10 CFR 50.36. Safety 
analysis acceptance criteria are not being altered by this 
amendment. Therefore, 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 
thisreview, it appears that the three standards of 50.92(c) are 
satisfied.Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light, 
11770 US Highway 1, North Palm Beach, Florida 33408
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: December 20, 1996
    Description of amendment request: The licensee proposed to delete a 
footnote associated with TS 2.1.1, ``Reactor Core Safety Limits,'' 
which requires reactor thermal power to be limited to 90% of 2700 
Megawatts thermal for Cycle 14 operation beyond 7000 Effective Full 
Power Hours [EFPH]. The thermal power limit was required pending 
completion of a Small Break Loss of Coolant Accident (SBLOCA) 
reanalysis that demonstrated acceptable results using input assumptions 
corresponding to an increased number of steam generator tubes being 
plugged. The SBLOCA reanalysis was completed and included with the 
submittal.
    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 proposed change will allow full Cycle 14 operation at 100% 
of rated power (2700 MWth), by deleting the requirement to
    derate to 90% of rated power prior to exceeding 7000 EFPH. This 
restriction was imposed in the NRC transmittal letter for License 
Amendment 145 for SBLOCA considerations when considering the 
increased SGTP [steam generator tube plugging]
    level of 30% plus or minus 7%. All Final Safety Analysis Report 
(FSAR) events, other than SBLOCA were evaluated at 100% of rated 
thermal power and showed no significant increases in the probability 
or consequences of accidents previously evaluated.
    The SBLOCA was reanalyzed to demonstrate continued compliance 
with 10 CFR 50.46 criteria. There is no impact of the proposed 
change on any FSAR accident initiator. The plant configuration and 
systems remain unchanged.
    Therefore, 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.
    (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.
    This proposed amendment removes the requirement in the Technical 
Specifications to derate to 90% of 2700 MWth for Cycle 14 operation 
beyond 7000 EFPH. There will be no change to the modes of operation 
of the plant. The plant configuration and the design functions of 
all the safety systems remain unchanged.
    The proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. The changes do 
not involve the addition of new equipment or the modification of 
existing equipment, nor do they alter the design of St. Lucie plant 
systems. Therefore, operation of the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The impact of the proposed change on available margin to the 
acceptance criteria for Specified Acceptable Fuel Design Limits 
(SAFDL), primary and secondary over-pressurization, peak containment 
pressure, potential radioactive releases, 10 CFR 50.46 requirements 
for the large break LOCA, and existing limiting conditions for 
operation has been evaluated and addressed in the reduced RCS 
[reactor coolant system] flow operating license Amendment No. 145. A 
requirement to derate to 90% of 2700 MWth was imposed based on the 
SBLOCA analysis. The small break LOCA analysis with 30% plus or 
minus 7% SGTP
    supported operation up to 7000 EFPH at 100% of rated thermal 
power. A reanalysis of SBLOCA with the limiting end-of-cycle 
conditions at 100% of rated power, demonstrates continued compliance 
with 10 CFR 50.46 criteria.
    Therefore, 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 thisreview, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light, 
11770 US Highway 1, North Palm Beach, Florida 33408
    NRC Project Director: Frederick J. Hebdon

GPU Nuclear Corporation, Docket No. 50-289, Three Mile Island, Unit 
1, Dauphine County, Pennsylvania

    Date of amendment request: December 3, 1996
    Description of amendment request: This amendment will incorporate 
certain improvements from the Standard Technical Specifications for B&W 
Plants (NUREG-1430).
    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:
    GPU Nuclear has determined that this Technical Specification 
Change Request involves no significant hazards consideration as 
defined in 10 CFR 50.92 because:
    1. Operation of the facility in accordance with the proposed 
amendment would not

[[Page 2191]]

involve a significant increase in the probability of occurrence or 
the consequences of an accident previously evaluated. The proposed 
amendment deletes limiting condition for operation (LCOs) from the 
TMI-1 Technical Specifications that are no longer required to be 
addressed in Technical Specifications per 10 CFR 50.36(c)(2)(ii). 
The proposed amendment also deletes a Surveillance requirement from 
the TMI-1 Technical Specifications. This surveillance requirement 
has no corresponding LCO and is formatted in the typical LCO format. 
These items are addressed in licensee controlled documents. This 
proposed amendment incorporates relaxation of selected timeclocks 
and surveillances frequencies consistent with NUREG 1430 and adds a 
timeclock to a unique LCO. The proposed changes do not modify the 
operation, limits or controls of systems, structures or components 
relied upon to prevent or mitigate the consequences [of] accidents 
previously evaluated. Also, the reliability of systems and 
components relied upon to prevent or mitigate the consequences of 
accidents previously evaluated is not degraded by the proposed 
changes. Therefore, this change does not involve a significant 
increase in the probability of occurrence or the consequences of an 
accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated because no 
new failure modes are created by the proposed changes.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment does not change any operating limits 
for reactor operation.
    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 staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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 amendment requests: October 25, 1996
    Description of amendment requests: The proposed amendments would 
incorporate the requirements of 10 CFR Part 50, Appendix J, Option B 
for containment leakage tests. In addition, the amendments would add a 
new section to Technical Specifications, which establishes the 
requirements of the containment leakage rate testing program, 
consistent with the Improved Standard Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes provide a mechanism within the TS for 
implementing a performance-based leakage rate test program which was 
promulgated by the revision to 10 CFR Part 50 to incorporate Option 
B to Appendix J. The proposed changes do not involve any physical or 
operational changes to structures, systems or components. The 
current safety analyses and safety design basis for the accident 
mitigation functions of the containment, the airlocks, and the 
containment isolation valves are maintained. Since the allowable 
containment leakage is still maintained within the analyzed limit 
assumed in the accident analysis, there is no adverse effect on 
either onsite or offsite dose consequences. Therefore, these changes 
will not increase the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not involve any physical or operational 
changes to structures, systems or components. No new failure 
mechanisms beyond those already considered in the current plant 
safety analyses are introduced. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Extending containment leakage rate test intervals from those 
currently provided in the Technical Specifications to those provided 
for in 10 CFR (Part) 50 Appendix J, Option B may slightly increase 
the risk due to an increased likelihood of containment leakage 
corresponding to the increased testing intervals. However, this is 
somewhat compensated by the corresponding risk reduction benefits 
received from the reduction in component cycling, stress, and wear 
associated with the increased intervals. When considering the total 
integrated risk, which includes all analyzed accident sequences, the 
possible additional risk associated with increasing test intervals 
is negligible.
    The NRC letter to NEI (Nuclear Energy Institute) dated November 
2, 1995, recognizes that changes similar to the proposed changes at 
PINGP (Prairie Island Nuclear Generating Plant) are required to 
implement Option B of 10 CFR (Part) 50, Appendix J. In NUREG-1493, 
``Performance-Based Containment Leak-Test Program'', dated September 
1995, which forms the basis for the Appendix J revision, the NRC 
concludes that adoption of performance-based testing will not 
significantly reduce the margin of safety. The containment leak rate 
data and component performance history at PINGP are consistent with 
the conclusions reached in NUREG-1493 and NEI 94-01. Thus, the 
proposed license amendments do not involve a significant reduction 
in a margin of safety and will continue to support the regulatory 
goal of ensuring an essentially leak-tight containment boundary.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the USAR or the Technical Specification 
Bases.
    Based on the evaluation described above, and pursuant to 10 CFR 
Part 50, Section 50.91, Northern States Power Company has determined 
that operation of the Prairie Island Nuclear Generating Plant in 
accordance with the proposed license amendment request does not 
involve any significant hazards considerations as defined by NRC 
regulations in 10 CFR Part 50, Section 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: November 18, 1996
    Description of amendment request: The amendments would amend the 
Technical Specifications for Susquehanna Steam Electric Station (SSES), 
Units 1 and 2 by increasing the maximum isolation times for the reactor 
core isolation cooling inboard warm-up line isolation valves (HV129F088 
and HV249F088) from 3 seconds to 12 seconds, the high pressure core

[[Page 2192]]

injection inboard warm-up line isolation valves (HV-155F100 and HV-
255F100) from 3 seconds to 6 seconds and the reactor recirculation 
process sample line (RRPSL) isolation valves (HV143F019 and HV243F019) 
from 2 seconds to 9 seconds.
    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.
    Chapters 6,9, and 15 of the FSAR [final safety analysis report], 
current operating cycles Reload Summary Reports for Units 1 and 2, 
Design Basis Document DBD046 (Seismic and Hydrodynamic Loads), and 
NUREG-0776 (Safety Evaluation Report for SSES), were reviewed to 
determine if the proposed action has an effect on the spectrum of 
analyzed anticipated operational transients or postulated design 
basis accidents.
    The proposed modifications involve replacing the pilot solenoid 
valves on the Reactor Recirculation Loop ``B'' Process Sample Line 
Isolation Valve (HV1/243F019) and the inboard RCIC [reactor core 
isolation cooling] and HPCI [high pressure core injection] Steam 
Warm-Up Line Isolation Valves (HV-1/249F088 and HV-1/255F100). They 
do not alter any system operation or control logic other than to 
increase the time it takes for the associated containment isolation 
valve to close. As discussed above, the effects of the increased 
isolation times for RCIC and HPCI impacted lines are bounded by the 
larger parallel lines with isolation times much greater than the new 
isolation times for the smaller lines. In the case of the Reactor 
Recirculation Loop ``B'' Process Sample Line, the worst case 
scenario for a line of that size is addressed in FSAR Section 15.6.2 
and the results have been found acceptable. In fact, the line 
breakage event analyzed in the FSAR section postulates a break 
outside containment that is not isolable and that does not require 
operator action for up to 10 minutes.
    The modifications enhance isolation valve performance by 
ensuring proper operation in the event of a degraded air system.
    Failures within the Process Sampling, RCIC or HPCI systems or 
their components are not postulated as causes of accident scenarios 
nor is increasing the stroke time of the subject containment 
isolation valves [HV-1/243F019]. These systems provide safety 
features utilized to mitigate the consequences of the accidents. 
However, the failure mode of the replacement solenoid valve is 
similar in each case to that of the solenoid valve being replaced in 
that it closes upon loss of power or loss of air supply. The current 
ability of the plant design to meet the single failure criterion is 
unchanged by this modification.
    Based on the above discussion, the proposed action does not 
involve a significant increase in the probability or consequences of 
an accident as previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Chapters 6, 9, and 15 of the FSAR were reviewed to determine if 
the proposed action [valve replacement with increased isolation 
times for associated HPCI, RCIC, RRPSL valves] has the potential of 
creating a postulated initiating event which is different than the 
analyzed anticipated operational transients or postulated design 
basis accident addressed. The review did not identify a postulated 
initiating event which would create the possibility for an accident 
of a different type due to replacing the pilot solenoid valves of 
the affected Reactor Recirculation LOOP ``B'' Process Sample Line or 
RCIC or HPCI Steam Warm-Up Line isolation valves.
    Also, the Reactor Recirculation Process Sample Line, as part of 
the Process Sampling System described in FSAR section 9.3.2.3, does 
not perform any safety functions. It is simply an alternate means 
for in line reactor water chemistry monitoring upon the loss of the 
RWCU system, and its loss does not create any possibility for 
unevaluated accidents or malfunctions.
    Thus, replacing the pilot solenoid valves on the affected 
Reactor Recirculation Process Sample Line, RCIC Steam Warm-Up Line, 
and HPCI Steam Warm-Up Line isolation valves as well as relocating 
the Process Sample Line solenoid valve for EQ [equipment 
qualification] purposes 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 proposed action involves replacing existing pilot solenoid 
valves on containment isolation valves for the Process Sampling, 
RCIC, and HPCI Systems, as listed above, with direct acting solenoid 
valves to ensure proper valve operation in the event of a degraded 
air or gas system as well as relocating the Process Sampling pilot 
solenoid valve for EQ purposes.
    a. Reactor Recirculation Loop ``B'' Process Sample Line
    The limiting condition for the operation of the Reactor 
Recirculation Loop ``B'' Process Sample Line Inboard Isolation Valve 
(HV-1/243F019) is governed by Technical Specification Section 3/
4.6.3 and its Bases which presently requires this valve to close 
within 2 seconds as defined in Technical Specification Table 3.6.3-
1. The proposed modifications involve replacing the pilot solenoid 
valve of the normally open isolation valve (HV-1/243F019) with a 
direct acting pilot solenoid valve as well as relocating the pilot 
solenoid valve to assure an EQ life which supports a 24 month 
operating cycle. The combined effects of a lower flow coefficient 
and relocating the solenoid valve will require an increase in the 
Technical Specification Table 3.6.3-1 isolation time from 2 seconds 
to 9 seconds.
    This increase in isolation time does not reduce the margin of 
safety as defined in the Technical Specification Section Basis, 
because breakage of lines of this size is addressed in the 
Susquehanna SES [steam electric station] FSAR Section 15.6.2 and the 
results found acceptable. In fact, the line breakage event analyzed 
postulates a break outside containment that is not isolable and that 
does not require operator action for up to 10 minutes. Also, it is 
noted that the outboard isolation valve, HV-1/243F020, also closes 
on the same containment isolation signal, and its Technical 
Specification isolation time limit remains 2 seconds.
    The failure mode of the affected Reactor Recirculation Loop 
``B'' Process Sample Line Inboard isolation valve is to close on 
loss of power or air supply, therefore, the proposed modifications 
do not affect the operability of the isolation valve or reduce the 
margin of safety.
    b. RCIC
    The limiting condition for operation of the RCIC system is 
governed by Technical Specification Section 3/4.7.3 and its Bases 
which requires RCIC to be operable as the primary non-ECCS source of 
emergency core cooling. The proposed modifications involve replacing 
the pilot solenoid valve of the normally closed Steam Warm-Up Line 
Isolation Valve (HV-1/249F088). This valve can be manually opened in 
the absence of an isolation signal to permit steam from the reactor 
to pressurize and warm the steam supply line downstream of the HV-1/
249F007 valve.
    Installation of the direct acting solenoid valve will require an 
increase in the Technical Specification Section 3/4.6.3 isolation 
time for the RCIC Steam Warm-Up Line Isolation Valve (HV-1/249F088) 
from 3 seconds to 12 seconds but does not reduce the margin of 
safety as defined in the Technical Specification Section Basis. The 
increase in closure time for the HV-1/249F088 isolation valve does 
not compromise the overall line isolation due to the fact that the 
impact of these 1'' warm up line valves is enveloped by the impact 
of the much larger 4'' RCIC inboard and outboard isolation valves 
(HV-1/249F007 and HV-1/249F008), which remain open an additional 8 
seconds before isolating. The 4'' valves are the limiting components 
for providing containment isolation for this line.
    The failure mode of the affected RCIC Steam Warm-Up Line 
Isolation Valve is to close, if open, on loss of power or air 
supply, therefore, the proposed modifications do not affect the 
operability of the isolation valve or reduce the margin of safety.
    c. HPCI
    The limiting condition for operation of the HPCI system is 
governed by Technical Specification Section 3/4.5.1 and its Bases 
which requires HPCI to be operable for proper Emergency Core Cooling 
System operation. Operability includes the HPCI pump and a flow path 
capable of taking suction from the suppression pool and delivering 
the water to the reactor vessel. The proposed modifications involve 
replacing the pilot solenoid valve of the normally closed Steam 
Warm-Up Line Isolation Valve (HV-1/255F100). This valve can be 
manually opened in the absence of an isolation signal, to permit 
steam from the reactor to pressurize

[[Page 2193]]

and warm the steam supply line downstream of the HV-1/255F002 valve.
    Installation of the direct acting solenoid valve will require an 
increase in the Technical Specification Section 3/4.6.3 isolation 
time for the HPCI Steam Warm-Up Line Isolation Valve (HV-1/255F100) 
from 3 seconds to 6 seconds but does not reduce the margin of safety 
as defined in the Technical Specification Section Basis. The 
increase in closure time for the HV-1/255F100 isolation valve does 
not compromise the overall line isolation due to the fact that the 
impact of these 1'' warm up line valves is enveloped by the impact 
of the much larger 10'' HPCI inboard and outboard isolation valves 
(HV-1/255F002 and HV-1/255F003) which remain open an additional 44 
seconds before isolating. The 10'' valves are the limiting 
components for providing containment isolation for this line.
    The failure mode of the affected HPCI Steam Warm-Up Line 
Isolation Valve is to close, if open, on loss of power or air 
supply, therefore, the proposed modifications do not affect the 
operability of the isolation valve or reduce the margin of safety.
    Thus, based on a review of the Technical Specification, their 
Bases, the FSAR and NUREG 0776 (Safety Evaluation Report for SSES), 
the replacement of the pilot solenoid valves does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: December 18, 1996
    Description of amendment request: The amendment would change the 
Susquehanna Steam Electric Station Unit 2 Technical Specifications to 
reflect the use of a 24-month operating cycle and the use of the 
ATRIUM-10 fuel design. The amendment includes changes to two 
definitions in Section 1, inclusion of new minimum critical power ratio 
safety limits in Sections 2.1.2 and 3.4.1.1.2, changes in Section 5.3.1 
to reflect the new fuel design, and the listing of Siemens Power 
Corporation topical reports in Section 6.9.3.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The applicable sections of the FSAR [Final Safety Analysis 
Report] are Chapters 5,6.3,9, and 15 of the FSAR. Chapter 5 
discusses the results of the ASME overpressure analyses for the 
reactor pressure boundary. Chapter 6.3 discusses the LOCA [loss-of-
coolant accident]. Chapter 9 discusses fuel storage and handling. 
Chapter 15 describes the transient and accident analyses, a majority 
of which have been generically dispositioned to be non-limiting. A 
discussion of the impact of the Technical Specification changes is 
provided below.
    The change to Definitions 1.2 and 1.3 makes the definitions 
applicable to ATRIUM-10. There are no effects on safety functions 
from this change.
    A cycle specific MCPR [minimum critical power ratio] Safety 
Limit analysis was performed for PP&L [Pennsylvania Power & Light 
Company] by SPC [Siemens Power Corporation]. This analysis used NRC 
approved methods described in Technical Specification Reference 13 
(ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2.). The SAFETY 
LIMIT MCPR calculation statistically combines uncertainties on 
feedwater flow, feedwater temperature, core flow, core pressure, 
core power distribution, and the uncertainty in the Critical Power 
Correlation. The SPC analysis used cycle specific power 
distributions and calculated MCPR values such that at least 99.9% of 
the fuel rods are expected to avoid boiling transition during normal 
operation or anticipated operational occurrences. The resulting two-
loop and single-loop values (Technical Specification sections 2.1.2 
and 3.4.1.1.2) are included in the proposed change. Thus, the 
cladding integrity and its ability to contain fission products is 
not adversely affected.
    The change to the Design Features (Section 5.3) increases the 
allowable enrichment. Analyses have demonstrated that the ATRIUM-10 
fuel will remain subcritical (k-effective<0.95) in both the spent 
fuel pool and the new fuel vault. Thus, the change to allowable 
enrichment has no impact on safety functions. The description of a 
fuel assembly (Section 5.3) is also revised to reflect the ATRIUM-10 
central water channel, and reference to an active fuel length of 150 
inches was deleted. This change reflects the physical 
characteristics of the ATRIUM-10 fuel and has no impact on the 
probability or consequences of an event.
    Included in the revised Technical Specifications via reference 
(Section 6.9.3.2) are additional NRC approved methodology reports. 
The NRC approved topical reports contain methodology which is used 
to assure safe operation of Unit 2 with ATRIUM-10 fuel. These 
methodologies assure that the core meets appropriate margins of 
safety for all expected plant operational conditions ranging from 
refueling and cold shutdown of the reactor through power operation. 
Thus, the results obtained from the analyses will provide assurance 
that the reactor will perform its design safety function during 
normal operation and design basis events.
    The BASES changes for Section 2.1.1 (THERMAL POWER,Low Pressure 
or Low Flow) reflect that the Safety Limit is valid for both 9x9-2 
and ATRIUM-10.
    Therefore, the proposed action does 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 changes to the Unit 2 Technical Specifications (Definitions, 
MCPR safety limits, Design Features, and inclusion of methodology 
references) to allow use of ATRIUM-10 fuel do not require any 
physical plant modifications, physically affect any plant 
components, or entail significant changes in plant operation. Thus, 
the proposed change does not create the possibility of a previously 
unevaluated operator error or a new single failure. The referenced 
methodology added to Section 6.9.3.2 contains NRC approved 
acceptance criteria. The consequences of transients and accidents 
will remain within the criteria approved by the NRC. Therefore, the 
proposed change does not create the possibility or 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 applicable Technical Specification Sections include 1.0, 
2.0, 3/4.4, 5.3, and 6.9.3.2.
    The changes to the Unit 2 Technical Specifications discussed in 
Item 1 above (Definitions, MCPR Safety Limits, Design Features, and 
inclusion of methodology references) to allow use of ATRIUM-10 fuel 
do not require any physical plant modifications, physically affect 
any plant components, or entail significant changes in plant 
operation. Therefore, the proposed change will not jeopardize or 
degrade the function or operation of any plant system or component 
governed by Technical Specifications. The NRC approved methods 
detailed in the references added to Section 6.9.3.2 maintain an 
equivalent margin of safety as currently defined in the bases of the 
applicable Technical Specification sections.
    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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library,

[[Page 2194]]

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: December 11, 1996 (TS 386)
    Description of amendment request: The proposed amendment changes 
the as-found tolerance for the main steam system safety/relief valves 
(S/RV) from plus or minus 1% to plus or minus 3%. The licensee states 
that the proposed change is consistent with methodology submitted by 
the Boiling Water Reactor Owners Group (BWROG) and approved by the NRC.
    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.
    TVA [the Tennessee Valley Authority, the licensee] is proposing 
a change to the ``as-found'' tolerances for the S/RV set points. 
This proposed TS [technical specification] amendment does not alter 
the frequency of verifying the S/RV lift set points, or the number 
of S/RVs required to be operable. The amendment does not involve 
physical changes or modifications to the S/RVs, or change the 
operating mode or safety function of the S/RVs. The safety lift set 
points will still be required to be set within a tolerance of plus 
or minus 1% following testing.
    S/RV actuation is not a precursor to any design basis accident 
analyzed in the BFN [Browns Ferry Nuclear Plant] UFSAR [Updated 
Final Safety Analysis Report]. Therefore, this change does not 
increase the probability of any previously evaluated accident.
    Generic considerations related to the set point tolerances were 
addressed in NEDC-31753P [BWROG In-Service Pressure Relief Valve 
Technical Specification Licensing Topical Report] and previously 
reviewed by NRC. In accordance with the NRC SER [Safety Evaluation 
Report, see letter from A. C. Thadani, NRC to C. L. Tully, BWROG, 
dated March 8, 1993] on utilizing the NEDC results, certain plant 
specific evaluations were performed to support the proposed change. 
Specifically, the current Unit 2 reload licensing report includes 
the transient analyses for the anticipated operational occurrences 
and the limiting overpressurization transient utilizing the plus or 
minus
    3% S/RV set point tolerance and were performed in accordance 
with NRC approved methods. The alternate operating modes were also 
included in the reload licensing report. These analyses concluded 
there is adequate margin to design core thermal limits and pressure 
limits for the reactor vessel. The corresponding Unit 3 core reload 
licensing report for the next operating cycle (starts in March 1997) 
is in progress and will also use the plus or minus 3% S/RV set point 
tolerance. Prior to the return of Unit 1 to service, the same reload 
analysis will be performed. Similar results to those for Unit 2 are 
expected.
    The operation of high pressure injection systems have been 
determined not to be adversely affected by the proposed change. LOCA 
[loss of coolant accident] response, containment hydrodynamic loads, 
pump and valve performance, and instrumentation performance were 
likewise satisfactorily evaluated. Therefore, this proposed change 
does not significantly increase the consequences of any previously 
evaluated accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a modification to plant 
equipment. No new failure modes are introduced. Plant systems will 
continue to function and no new system interactions are introduced 
by this proposed change. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change has been analyzed in accordance with NRC 
approved methodology and the margins of safety for the design basis 
accidents and transients analyzed in Chapter 14 of the BFN UFSAR 
have not been significantly reduced. 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 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

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: December 11, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 1.0, ``Definitions,'' TS 
Table 1.2, ``Frequency Notation,'' TS Section 3/4.3, 
``Instrumentation,'' and TS Section 3/4.5, ``Emergency Core Cooling 
Systems.'' Surveillance requirements would be modified to account for 
the increase in the fuel cycle, consistent with Generic Letter 91-04, 
``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-month Fuel Cycle,'' dated April 2, 1991. 
Administrative changes consistent with the fuel cycle change are also 
proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in 
accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed revisions to increase the surveillance test 
intervals from 18 to 24 months for the subject Technical 
Specifications (TS) 3.4.3.1.1, Reactor Protection System 
Instrumentation; TS 3/4.3.2.1, Safety Features Actuation System 
Instrumentation; TS 3/4.3.2.2, Steam and Feedwater Rupture Control 
System Instrumentation; TS 3/4.3.3.1, Radiation Monitoring 
Instrumentation; TS 3/4.3.3.5.2, Remote Shutdown Instrumentation; TS 
3/4.3.3.6, Post-Accident Monitoring Instrumentation, TS 3/4.5.1, 
Emergency Core Cooling Systems (ECCS), Core Flooding Tanks; and TS 
3/4.5.2, Emergency Core Cooling Systems, ECCS Subsystems - Tavg 
greater than or equal to 280 deg.F. Initiating conditions and 
assumptions remain as previously analyzed for accidents in the DBNPS 
Updated Safety Analysis Report.
    These revisions do not involve any physical changes to systems 
or components, nor do they alter the typical manner in which the 
systems or components are operated.
    Review results of historical 18 month surveillance data and 
maintenance records support an increase in the surveillance test 
intervals from 18 to 24 months (and up to 30 months on a non-routine 
basis) because little, if any, potential for an increase in a 
failure rate of a system or component was identified during these 
reviews.
    These proposed revisions are consistent with NRC guidance on 
evaluating and proposing such revisions as provided in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2, 
1991.
    The proposed revision to Technical Specification Table 1.2, 
Frequency Notation, and the related proposed revision from an

[[Page 2195]]

``R'' frequency notation to an ``E'' frequency notation for 
Technical Specification Surveillance Requirements that are remaining 
on an 18 month frequency, are administrative in nature, do not 
change current actual Technical Specification requirements, and do 
not affect previously evaluated accidents.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the source term, containment 
isolation or radiological releases are not being changed by these 
proposed revisions. Existing system and component redundancy is not 
being changed by these proposed changes. Existing system and 
component operation is not being changed by these proposed changes. 
The assumptions used in evaluating the radiological consequences in 
the DBNPS Updated Safety Analysis Report are not invalidated.
    The proposed revision to Technical Specification Table 1.2, 
Frequency Notation, and the related proposed revision from an ``R'' 
frequency notation to an ``E'' frequency notation for Technical 
Specification Surveillance Requirements that are remaining on an 18 
month frequency, are administrative in nature, do not change current 
actual Technical Specification requirements, and do not affect 
previously evaluated accidents.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because these 
revisions do not involve any physical changes to systems or 
components, nor do they alter the typical manner in which the 
systems or components are operated.
    Review results of historical 18 month surveillance data and 
maintenance records support an increase in the surveillance test 
intervals from 18 to 24 months (and up to 30 months on a non-routine 
basis) because little, if any, potential for an increase in a 
failure rate of a system or component was identified during these 
reviews. No changes are being proposed to the type of testing being 
performed, only to the length of the surveillance test interval.
    The proposed revision to Technical Specification Table 1.2, 
Frequency Notation, and the related proposed revision from an ``R'' 
frequency notation to an ``E'' frequency notation for Technical 
Specification Surveillance Requirements that are remaining on an 18 
month frequency, are administrative in nature, do not change current 
actual Technical Specification requirements, and do not affect the 
manner in which systems and components are being operated or tested.
    3. Not involve a significant reduction in a margin of safety 
because the review results of the historical 18 month surveillance 
data and maintenance records identified little, if any, potential 
for an increase in a failure rate of a system or component due to 
increasing the surveillance test interval to 24 months. Existing 
system and component redundancy is not being changed by these 
proposed changes.
    The proposed revision to Technical Specification Table 1.2, 
Frequency Notation, and the related proposed revision from an ``R'' 
frequency notation to an ``E'' frequency notation for Technical 
Specification Surveillance Requirements that are remaining on an 18 
month frequency, are administrative in nature, do not change current 
actual Technical Specification requirements, and do not reduce the 
margin of safety.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences, 
therefore there are no significant reductions 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: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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.

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

    Date of application for amendment: May 1, 1996, as 
supplementedNovember 26, 1996.
    Brief description of amendment: The proposed amendment will modify 
Table 3.1.1, ``Reactor Protection System (SCRAM) Instrumentation 
Requirement,'' Table 3.2.C.1, ``Instrumentation That Initiates Rod 
Blacks,'' and Technical Specification 3/4.4, ``Standby Liquid 
Control.''
    Date of issuance: December 27, 1996
    Effective date: December 27, 1996
    Amendment No.: 169
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (61 FR 
28606) The November 26, 1996, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration.The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 27, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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

    Date of application for amendments: July 26, 1996, as supplemented 
on September 3, 1996, September 18, 1996, two submittals dated October 
14, 1996, October 22, 1996, two submittals dated November 8, 1996, and 
December 17, 1996.
    Brief description of amendments: The amendments allow Commonwealth 
Edison Company to control the reactor coolant system pressure and 
temperature limits for heatup, cooldown, low temperature operation and 
hydrostatic testing. They also revise the reactor vessel material 
surveillance program specimen withdrawal schedule

[[Page 2196]]

such that the Unit 2 removal of capsule X is delayed until 19 Effective 
Full Power Years.
    Date of issuance: December 20, 1996
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 177 and 164
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 25, 1996 (61 
FR 50341). The September 3, 1996, September 18, 1996, two submittals 
dated October 14, 1996, October 22, 1996, two November 8, 1996, and 
December 17, 1996, submittals provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
December 20, 1996.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
    Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe 
County, Michigan
    Date of application for amendment: March 25, 1996 (NRC-96-0003)
    Brief description of amendment: The amendment revises the testing 
requirements used to determine the operability of the charcoal in the 
engineered safety feature systems.
    Date of issuance: December 23, 1996
    Effective date: December 23, 1996, with full implementation within 
45 days
    Amendment No.: 110
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40014) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 23, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

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

    Date of application for amendments: December 11, 1996, as 
supplemented December 17, 19, and 26, 1996
    Brief description of amendments: The amendments approve changes to 
the Updated Final Analysis Report (UFSAR), and require that the changes 
be submitted with the next update of the UFSAR pursuant to 10 CFR 
50.71(e). The associated Safety Evaluation delineates the staff's 
review and findings regarding the one-time emergency power engineered 
safeguards functional test.
    Date of issuance: January 2, 1997
    Effective date: January 2, 1997
    Amendment Nos.: 220, 220, 217
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Updated Final Safety Analysis Report. Public 
comments requested as to proposed no significant hazards consideration: 
Yes. (61 FR 66699 December 18, 1996) The notice provided an opportunity 
to submit comments on the Commission's proposed no significant hazards 
consideration determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by January 2, 
1997, as corrected to read January 17, 1997, but indicated that if the 
Commission makes a final no significant hazards consideration 
determination, any such hearing would take place after issuance of the 
amendments.
    The December 17, 19, and 26, 1996, letters provided additional 
information that did not change the scope of the December 11, 1996, 
application and initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and a final no significant hazards consideration 
determination are contained in a Safety Evaluation dated January 2, 
1997.
    Local Public Document Room location:  Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska

    Date of amendment request: May 31, 1996
    Brief description of amendment: The amendment revises the technical 
specifications to increase the amount of trisodium phosphate (TSP) 
dodecahydrate located in the containment sump storage baskets.
    Date of issuance: December 30, 1996
    Effective date: December 30, 1996
    Amendment No.: 179
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40025) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 30, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

Philadelphia Electric Company, Docket No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: August 1, 1996
    Brief description of amendment: This amendment revised the 
Technical Specifications Section 3/4.4.6 (i.e., Figure 3.4.6.1-1) to 
reflect the addition of two hydrotest curves, effective for 6.5 and 8.5 
Effective Full Power Years (EFPY), to the existing Pressure-Temperature 
Operating Limit (PTOL) curves for LGS Unit 2.
    Date of issuance: December 30, 1996
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment No.: 80
    Facility Operating License No. NPF-85. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57490) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 30, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464

Notice Of Issuance Of Amendment To Facility Operating License And 
FinalNo Significant Hazards Consideration Determination

    During the period since publication of the last biweekly notice, 
individual notices of issuance of amendments have been issued for the 
facilities as listed below. These notices were previously published as 
separate individual notices. They are repeated here because this 
biweekly notice lists all amendments that have been issued for which 
the Commission has made a final determination that an amendment 
involves no significant hazards consideration.
    In this case, a prior Notice of Consideration of Issuance of 
Amendment, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing was issued, a hearing was requested, and 
the amendment was issued before any hearing because the Commission made 
a final determination that the

[[Page 2197]]

amendment involves no significant hazards consideration.
    Details are contained in the individual notice as cited.

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

    Date of amendment request: December 6, 1996
    Brief description of amendment request: The amendment would revise 
Technical Specification (TS) Section 2.1 and its associated TS Basis to 
reflect the change in the Minimum Critical Power Ratio Safety Limit due 
to the use of GE13 fuel product line and the cycle-specific analysis 
performed by General Electric Company (GE), for Limerick Generating 
Station, Unit 2, Cycle 5.
    Date of publication of individual notice in Federal Register: 
December 23, 1996 (61 FR 67582)
    Expiration date of individual notice: January 22, 1997
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Dated at Rockville, Maryland, this 8th day of January 1997.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 97-848 Filed 1-14-97; 8:45 am]
BILLING CODE 7590-01-F