[Federal Register Volume 62, Number 156 (Wednesday, August 13, 1997)]
[Notices]
[Pages 43365-43381]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10813]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice


Applications and Amendments to Facility Operating 
LicensesInvolving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 19, 1997, through August 1, 1997. The 
last biweekly notice was published on July 30, 1997, (62 FR 40843).

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 September 12, 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

[[Page 43366]]

why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. 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.

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

    Date of amendments request: July 8, 1997
    Description of amendments request: The proposed amendments remove 
the suppression chamber water volume band from Technical Specification 
(TS) 3.6.2.1.a.1 while retaining the equivalent water level band. The 
values for the suppression chamber water volume corresponding to the 
low and high suppression chamber water levels will be retained in the 
Bases section of the TS and will be revised by the proposed amendments 
to account for the displacement of water due to the planned 
installation of larger emergency core cooling system suction strainers. 
The revised relationship between the high and low suppression chamber 
water levels and suppression chamber water volume will also be 
described in the Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below: 1. The proposed amendments do 
not involve a significant increase in the probability or consequences 
of an accident previously evaluated.
    The proposed change revises the values of the minimum and 
maximum suppression chamber pool water volume limits. The water 
inventory of the suppression chamber pool is not a precursor of an 
accident and, therefore, cannot increase the probability of an 
accident previously evaluated. The pressure suppression chamber 
water pool mitigates the consequences of loss-of-coolant accidents 
(LOCAs) transients [sic], and other events by providing a heat sink 
for reactor primary system energy releases. The proposed minimum and 
maximum pool water volume values will be consistent with the current 
suppression chamber pool water level limits. No changes to setpoints 
will be made as a result of the proposed change. The impact of the 
proposed change to the minimum and maximum suppression chamber pool 
volume limits on the suppression chamber pool temperatures and 
pressures following a design basis LOCA, an Safety/Relief Valve 
(SRV) blowdown event, an Anticipated Transient Without Scram (ATWS) 
event, an Appendix R fire event, and a station blackout event has 
been evaluated and does not cause accident parameters to exceed 
acceptable values. In addition, the impact the proposed change has 
on the time to reach cold shutdown when using the alternate Residual 
Heat Removal (RHR) shutdown cooling function is negligible. The 
potential impact the proposed change to the suppression chamber pool 
water volume limits has on SRV line loads, SRV discharge line 
reflood height, wetwell pressurization, suppression chamber pool 
swell loads, vent thrust loads, and condensation oscillation and 
chugging loads was also reviewed. The change to the suppression 
chamber pool water volume limits has no significant adverse impact 
on any of these parameters. As delineated above, the capability of 
the suppression chamber water pool to perform its mitigative 
functions is not affected by the proposed change. Therefore, the 
proposed change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

[[Page 43367]]

    The proposed change revises the values of the minimum and 
maximum volume of the suppression chamber water pool. The proposed 
change will not alter any physical mechanism by which the 
suppression chamber water pool volume is maintained between the 
minimum and maximum values. The suppression chamber pool water level 
will continue to be maintained between -27 and -31 inches. The 
suppression chamber pool water level limits are retained in 
Technical Specification (TS) 3.6.2.1.a.1, since this is the 
information available to the operators regarding the suppression 
chamber pool water volume limits. These level limits are equivalent 
to the suppression chamber pool water volume limits; therefore, it 
is only the presentation of the equivalency that is being relocated 
to the Bases and the Updated Final Safety Analysis Report (UFSAR). 
As such, the relocated suppression chamber pool water volume limits 
are not required to be in the TS to provide adequate protection of 
the public health and safety. As a result of the proposed strainer 
changes, there are no physical changes to any other suppression 
chamber components or instrumentation. No new mode of operation is 
introduced as a result of the proposed change. Analyses have been 
performed which conclude that the proposed change will not affect 
the operability of the equipment designed to mitigate the 
consequences of an accident. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed change revises the values of the minimum and 
maximum suppression chamber water pool volumes. The pressure 
suppression chamber water pool mitigates the consequences of several 
postulated accidents and transients by providing a heat sink for the 
primary coolant system. These accidents and events are the 
postulated design basis LOCA, an SRV blowdown event, an ATWS event, 
an Appendix R fire, and station blackout events. The consequences of 
the change in the suppression pool water volume limits have been 
evaluated for these events, and there is no significant reduction in 
the margin of safety.
    The results of the analyses for the postulated accidents and 
events indicate the temperature of the suppression chamber pool 
water could increase slightly as a consequence of the decrease in 
the minimum suppression chamber pool water volume limit. However, 
the suppression chamber pool water and containment temperatures 
remain within acceptable values. The impact of the calculated 
increase in containment temperature on the available Net Positive 
Suction Head (NPSH) for the Residual Heat Removal (RHR) and Core 
Spray pumps has been evaluated for the postulated design basis LOCA 
and indicate[s] adequate NPSH is maintained throughout the event.
    The potential impact of the proposed change to the suppression 
chamber pool water volume limits on the SRV line loads, SRV 
discharge line reflood height, wetwell pressurization, suppression 
chamber pool swell loads, vent thrust loads, and condensation 
oscillation and chugging loads was evaluated with the conclusion 
that there are no adverse impacts on these parameters.
    In addition, a small suppression chamber pool water temperature 
increase could result due to the reduction in minimum suppression 
pool volume limit in the event reactor shutdown is conducted through 
a path utilizing the suppression chamber pool. Such a shutdown path 
is an alternative to the normal RHR shutdown cooling function, and 
the small potential increase in temperature results in a negligible 
increase in the time required to reach cold shutdown conditions. 
Cold shutdown conditions can still be reached well within the 
Technical Specification requirements.
    The proposed increase in the suppression pool water volume limit 
does not adversely impact containment parameters as a result of 
postulated accidents and events. The potential increase in 
temperature of the pressure suppression chamber pool water does not 
significantly decrease the ability to maintain containment 
parameters within acceptable limits. The potential increase in time 
to reach cold shutdown conditions utilizing the suppression pool as 
an alternative to the normal RHR shutdown cooling function is 
negligible. Therefore, the proposed change to revise the minimum and 
maximum suppression water pool volumes does not involve a 
significant reduction in a margin of safety.
    The suppression chamber pool water level limits are retained in 
TS 3.6.2.1.a.1, since this is the information available to the 
operators regarding the suppression chamber pool water volume 
limits. These level limits are equivalent to the suppression chamber 
pool water volume limits and the equivalency is being relocated to 
the Bases and the UFSAR. As such, the relocated suppression chamber 
pool water volume limits are not required to be in the TS to provide 
adequate protection of the public health and 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 North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Gordon E. Edison, Acting

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of amendment request: July 21, 1997
    Description of amendment request: Technical Specification Change 
Request Concerning Emergency Feedwater Surveillance Testing. This 
request is to make several changes to the ANO-2 Technical 
Specifications including an extension of the emergency feedwater (EFW) 
pump surveillance testing frequency, a reduction in the minimum steam 
generator pressure required to perform the surveillance testing on the 
turbine-driven EFW pump, and a modification to the EFW pump testing 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    The proposed changes included in this amendment request are 
being made to the emergency feedwater (EFW) system technical 
specification (TS) surveillances. These changes include surveillance 
interval modifications, allowances to perform the turbine driven EFW 
pump surveillance at a lower steam generator (S/G) pressure, 
removing the requirements to perform specific EFW surveillance 
requirements (SRs) during plant shutdowns, bases changes, and 
various administrative changes. These changes are consistent with 
the applicable SRs located in NUREG-1432 and have therefore, been 
previously approved by the NRC.
    These changes do not alter the functional characteristics of any 
plant component and do not allow any new modes of operation of any 
component. The accident mitigation features of the plant are not 
affected by the proposed amendment request. No modifications have 
been made to the EFW system due to this amendment request. Although 
the minimum steam generator pressure has been reduced for the 
turbine driven EFW pump testing, calculations show that significant 
margin exists between the proposed value and that needed to 
adequately perform the test. The capability of the EFW pumps to 
perform their required safety function is not impacted by this 
change. The addition of the electric driven EFW flow path 
verification will help [to] assure proper alignment of both trains 
of EFW following extended outages.
    The accident mitigation features of the plant are not affected 
by the proposed amendment. No modification has been made to the pump 
or turbine driver. The capability of the turbine driven EFW pump to 
perform its required function is not impacted by this change. The 
EFW pumps will be tested in accordance with the more restrictive of 
the

[[Page 43368]]

data points required by the safety analysis or the inservice testing 
program. Therefore, this change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated.
    No new possibility for an accident is introduced by modifying 
the proposed specifications for the surveillance testing of the EFW 
pumps. The EFW surveillance requirements will continue to 
demonstrated the pump's ability to perform its safety function. The 
modifications to the proposed EFW surveillance requirements are 
consistent with the current revision of NRC approved NUREG -1432, 
``Standard Technical Specifications Combustion Engineering Plants'' 
(ITS). Therefore, this change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does Not Involve a Significant Reduction in Margin of Safety.
    The safety function of the EFW system is not altered as a result 
of this change. The capability of the EFW pumps to perform their 
required function is not impacted by this change. The capability of 
the EFW pumps is not impacted by this change. The EFW pumps will be 
tested and proven operable in accordance with the more restrictive 
of the data points required by the safety analysis of the inservice 
testing program. The addition of the electric driven EFW flow path 
verification will help assure [to] proper alignment of both trains 
of EFW following extended outages. 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: James Clifford, Acting

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: June 26, 1997
    Description of amendment request: The proposed amendment would 
revise the Operating License No. DPR-72, License Condition 2.C.(5) and 
delete the requirement for installation and testing of flow indicators 
in the emergency core cooling system to provide indication of 40 
gallons per minute flow for boron dilution from the license. Approval 
of this amendment will allow removal of the appropriate flow 
indicators, DH-45-Fl and DH-46-Fl, from the Crystal River 3 (CR3) Final 
Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    This license amendment removes the requirement for flow 
indication on the DH drop line and auxiliary pressurizer spray line 
for boron precipitation mitigation during a LOCA [Loss of Coolant 
Accident]. The original need for these indicators was to provide 
flow indication to the operator to aid in decision making relative 
to an alternate active method for boron precipitation prevention. 
Alternate active methods have been replaced by the passive flow path 
through the gaps which exist between the reactor vessel and the 
reactor vessel internals. Since auxiliary pressurizer spray flow is 
no longer used, and no other active means is required to be employed 
by the operator in the event drop line flow is not indicated, the 
original usefulness of and need for this indication no longer 
exists. Removal of this requirement from the license condition does 
not involve a change in the Improved Technical Specifications. The 
operators do not use the flow indication for decision making in 
post-accident conditions. Since these instruments are no longer used 
for boron precipitation mitigation during a LOCA, abandonment or 
removal of flow indicator DH-45-Fl and DH-46-Fl does not increase 
the probability of an accident because no previously evaluated 
accidents at CR-3 are initiated by DH-45-Fl or DH-46-Fl. Those CR-3 
accidents that are analyzed are contained in the Final Safety 
Analysis Report (FSAR) and include events such as Loss-of-Coolant 
Accidents, Main Steam Line Breaks, Station Blackout, Anticipated 
Transients Without Scram, etc. Since DH-45-Fl and DH-46-Fl are 
attached to the outside of the DH drop line and auxiliary 
pressurizer spray line, their removal will not change the design, 
material, or construction standards applicable to the DH System 
piping. The removal of the indicator will not affect overall system 
performance of the ECCS. All of these previously evaluated accidents 
described in the CR-3 FSAR have dose consequences which remain well 
within the requirements of 10 CFR Part 100 (25 rem whole body, 300 
rem thyroid) and GDC [General Design Criterion] 19 (5 rem whole 
body, or its equivalent to any part of the body). Removal of DH-45-
Fl and DH-46-Fl will not alter any assumptions made in evaluating 
the radiological consequences of any accident described in the FSAR 
nor will it affect any fission product barriers since the ECCS and 
containment systems will still perform to meet design requirements. 
Therefore, removal of DH-45-Fl and DH-46-Fl will not alter the 
consequences of an accident previously evaluated.
    Criterion 2
    The change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed license amendment removes the requirement for 
indicators which were originally installed to aid the operator in 
decision making relative to an alternate flow path for boron 
precipitation mitigation during a LOCA. These indicators no longer 
serve this purpose, since alternate active flow paths are no longer 
considered. Evaluations which consider boron precipitation no longer 
rely on three active methods of mitigation, but rather one active 
and one passive. Operator action is not required to effect the 
backup method in the event that the primary method fails due to a 
single active failure. The flow indicators are external to the DH 
System piping. They do not penetrate any piping so their removal 
cannot create the possibility of a new or different kind of 
accident. The accident mitigation strategies remain the same 
regardless of whether or not the flow indicators are present. 
Therefore, the flow indicators serve no purpose in the analyses. The 
proposed amendment does not affect any of the parameters or 
conditions that could contribute to the initiation of any accidents.
    Criterion 3
    The change does not involve a significant reduction in the 
margin of safety.
    Boron precipitation within the reactor vessel during post-LOCA 
conditions, if it were to occur, would challenge the margin of 
safety that is provided by assuring compliance with Criterion 5 of 
10 CFR 50.46. The license amendment does not change the methodology 
of mitigating the consequences of boron precipitation following a 
LOCA as described in the current licensing basis. The primary method 
of flow through the DH drop line and the use of gap flow as the 
``backup'' method for prevention of boron precipitation have been 
analyzed, shown to meet all the criteria of 10 CFR 50.46, and 
accepted by the NRC. The passive method requires no specific 
operator action for initiation, in the event that the primary method 
fails due to a single active failure. Therefore, the indication 
serves no safety function and does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428

[[Page 43369]]

    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042
    NRC Project Director: Frederick J. Hebdon

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: July 18, 1997
    Description of amendment request: The proposed amendment would 
revise the Crystal River 3 (CR-3) technical specifications (TS) to 
incorporate a new TS 3.4.11 for a Low Temperature Overpressure 
Protection (LTOP) System. The proposed changes would be consistent with 
the recommendations in the NRC Generic Letter 88-11, ``NRC Position on 
Radiation Embrittlement of Reactor Vessel Materials and Its Impact on 
Plant Operations.'' TS 3.5.3 and associated TS Bases would also be 
revised to reflect the proposed change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    This change does not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    There are currently no LTOP requirements in the CR-3 Improved 
Technical Specifications. CR-3 currently implements LTOP features 
through administrative controls and a lowered PORV [power-operated 
relief valve] setpoint. The proposed change will establish new LTOP 
technical specification requirements necessary to preclude an LTOP 
event from occurring. The proposed LTOP requirements are based on 
safety analyses that apply ASME [American Society of Mechanical 
Engineers] Code Case N-514. These requirements will decrease the 
probability of a low temperature overpressure event by providing 
protection for all pressure and temperature combinations for which a 
low temperature overpressure event may be postulated.
    The consequences of a low temperature overpressure accident are 
not affected by this change. There is no change to the 10 CFR [Code 
of Federal Regulations] Part 100 dose calculation for a low 
temperature overpressure accident.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated
    This change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The new LTOP Technical Specification does not require 
modification to the plant nor does it create a new mode of plant 
operation. The LTOP system adds no new accident initiators.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The proposed change does not involve a significant reduction in 
the margin of safety and will provide added safety benefit gained 
through the requirements to preclude a low temperature 
overpressurization event to the RCS [reactor coolant system].
    The margin of safety prior to having an LTOP system was limited 
due to the informal, administrative method of minimizing the impact 
of a low temperature overpressure accident. By formalizing these 
requirements into a technical specification, at the least, margin of 
safety is retained and perhaps improved due to the elevated 
significance of required actions.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042
    NRC Project Director: Frederick J. Hebdon

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: July 29, 1997
    Description of amendment request: The proposed amendment would 
revise the Crystal River Nuclear Generating Unit 3 (CR3) technical 
specifications (TS) to add subcooling margin and decay heat removal 
(low pressure injection) flow and correct certain nomenclature in the 
post-accident monitoring (PAM) instrumentation TS. In addition, the 
licensee proposes to add emergency diesel generator (EDG) kilowatt (kW) 
indication to the PAM instrumentation. Specifically, the following TS 
would be revised:
    A. Table 3.3.17-1, Function 8: The descriptor is changed from 
``Containment Pressure (Narrow Range)'' to ``Containment Pressure 
(Expected Post-Accident Range).''
    B. Table 3.3.17-1, Function 18: The required channels for Core Exit 
Temperature (Backup) is changed from ``2 sets of 5'' to ``3 per core 
quadrant.''
    C. Table 3.3.17-1: A new Function 20 is added and designated as 
``Low Pressure Injection Flow'', with 2 required channels, and 
Condition E.
    D. Table 3.3.17-1: A new Function 21 is added and designated as 
``Degrees of Subcooling'', with 2 required channels, and Condition E.
    E. Table 3.3.17-1: A new Function 22 is added and designated as 
``Emergency Diesel Generator kW Indication'', with 2 required channels, 
and Condition E. A note clarifying the number of required channels is 
added: ``(c): one indicator per EDG''.
    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 items A, B, C, D and E 
corresponds to the specific TS changes described above.
    1. The proposed changes will not significantly increase the 
probability or consequences of an accident previously evaluated 
because:
    A/B. The changes in containment pressure and core exit 
thermocouple nomenclature do not reflect any physical changes to the 
facility. This would have no impact on accident probability or 
consequences.
    C/D/E. The addition of low pressure injection flow, degrees of
    subcooling, and EDG kW indication to the Post-Accident 
Monitoring Instrumentation LCO [Limiting Condition for Operation] is 
being done to comply with a commitment made during the technical 
specification improvement program to include in the technical 
specifications that instrumentation which monitors variables 
classified as Type A in accordance with Regulatory Guide 1.97. These 
three variables have been reclassified as Type A. The associated 
instruments are used in post-accident conditions to prompt the 
operators to take certain mitigative actions. Therefore, the 
probability of an accident occurring is unaffected. As part of the 
re-classification of these variables to Type A and inclusion in 
technical specifications, the associated monitoring instrumentation 
will be under more strict surveillance and control, which provides 
additional assurance that the prescribed manual operator actions 
will be implemented when necessary. This, in turn, assures the 
previously evaluated accident consequences remain valid.
    2. The proposed changes will not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because:
    A/B. The changes in containment pressure and core exit
    thermocouple nomenclature do not reflect any physical changes to 
the facility. The changes provide clarification for the instruments 
which are required to comply with the LCO. This would not create 
possibility of a new or different kind of accident.
    C/D/E.The addition of low pressure injection flow, degrees of 
subcooling, and EDG kW indication to the Post-Accident Monitoring 
Instrumentation LCO is being

[[Page 43370]]

done to comply with a commitment made during the technical 
specification improvement program to include in the technical 
specifications that instrumentation which monitors variables 
classified as Type A in accordance with Regulatory Guide 1.97. These 
three variables have recently been reclassified as Type A. The 
associated instruments are used after an accident occurs to prompt 
the operators to take certain mitigative actions. Since the 
instrumentation is used only post-accident, these changes do not 
create the possibility of a new or different kind of accident.
    3. The proposed change will not involve a significant reduction 
to the margin of safety because:
    A/B. The changes in containment pressure and core exit 
thermocouple nomenclature have no affect on the margin of safety. 
The changes provide clarification of the technical specifications. 
This reduces the potential for confusion regarding this 
instrumentation.
    C/D/E. The addition of low pressure injection flow, degrees of
    subcooling, and EDG kW indication to the post-accident 
monitoring instrumentation table in technical specifications results 
in added controls on the OPERABILITY of this post-accident 
monitoring instrumentation and provides greater assurance that it 
will be available should an accident occur.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: July 18, 1997
    Description of amendment request: The proposed amendment adds a new 
Technical Specification and associated Bases to address the operability 
of the steam generator atmospheric relief bypass valves (SGARBVs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed revision does not involve an SHC because the 
revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The operability of the SGARBVs provides a method to recover from 
a SGTR [steam generator tube rupture] event during which the 
operator is required to perform a limited cooldown to establish 
adequate subcooling as a necessary step to limit the primary to 
secondary break flow into the ruptured steam generator. For other 
design events, the SGARBVs provide a safety grade method for cooling 
the unit to residual heat removal entry conditions should the 
preferred heat sink via the steam bypass system or the steam 
generator atmospheric relief valves be unavailable. This proposed 
revision to the Technical Specifications will add a new Technical 
Specification 3/4.7.1.6 and its associated Bases Section 3/4.7.1.6 
which were developed bases on the information contained in the 
Westinghouse Improved Standard Technical Specifications, NUREG 1431, 
Rev. 1. The proposed specification and bases provide further 
assurance that the SGARBVs will be available to function as 
described in the accident analysis.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This proposed revision to the Technical Specifications to add a 
new specification and bases for the SGARBVs does not cause a change 
in the operation of any system or component during normal or 
accident conditions.
    Therefore, the proposed revision does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed new Technical Specification 3/4.7.1.6 and its 
associated Bases Section 3/4.7.1.6 were developed based on the 
information contained in the Westinghouse Improved Standard 
Technical Specifications, NUREG 1431, Rev. 1. The SGARBV's are not 
currently in the Technical Specifications of Millstone Unit No. 3 
and are being added to ensure accident mitigation functional 
capability. The NUREG 1431, Rev. 1 surveillance frequency is 18 
months. The NUREG 1431, Rev. 1 surveillance frequency bases reads 
``operating experience has shown that these components usually pass 
the surveillance when performed at the 18 month frequency''. The 
proposed frequency acceptability has been evaluated by reviewing 
SGARBV AWO's [automated work order's] for the period from Jan. 1990 
to April 1997 to confirm the absence of excessive work orders which 
indicate valve functional failures and none were identified. 
Additionally, each SGARBV line consists of one SGARBV and an 
associated block valve. These proposed changes are consistent with 
the design and operation of the SGARBVs. There is no negative affect 
on the dose consequences from any design basis event or core damage 
frequency.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Deputy Director: Phillip F. McKee

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: November 27, 1996
    Description of amendment requests: The proposed amendment[s] would 
incorporate new steam generator tube sleeve designs and installation 
and examination techniques into the Prairie Island 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[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The supporting technical evaluation and safety evaluation for 
the Combustion Engineering leak tight sleeves demonstrate that the 
sleeve configuration will provide steam generator tube structural 
and leakage integrity under normal operating and accident 
conditions. The sleeve configurations have been designed and 
analyzed in accordance with the requirements of the ASME [American 
Society of Mechanical Engineers] Code. Mechanical testing has shown 
that the sleeve and sleeve joints provide margin above acceptance

[[Page 43371]]

limits. Ultrasonic examination is used to verify the leak tightness 
of the above the [sic] tubesheet sleeve welds. Testing has 
demonstrated the leak tightness of the hard roll joint as well as 
the structural integrity of the hard roll joint. Tube rupture can 
not occur at the hard roll joint due to the reinforcing effect of 
the tubesheet. Tests have demonstrated that tube collapse will not 
occur due to postulated LOCA [loss-of-coolant accident] loadings.
    The existing Technical Specification leakage rate requirements 
and accident analysis assumptions remain unchanged in the event that 
significant leakage did occur from the sleeve joints or that a 
sleeve assembly ruptured. Any leakage through the sleeve assembly is 
fully bounded by the existing steam generator tube rupture analysis 
included in the Prairie Island Plant USAR [updated safety analysis 
report]. The proposed sleeving repair does not adversely impact any 
other previously evaluated design basis accident.
    The sleeve minimum acceptable wall thickness used for developing 
the depth based plugging limit for the sleeve is determined using 
the guidance of draft Regulatory Guide 1.121 [<gr-thn-eq>Bases for 
Plugging Degraded PWR [Pressurized-Water Reactor] Steam Generator 
Tubes<gr-thn-eq>] and the pressure stress equation of Section III of 
the ASME Code. Evaluation of the minimum acceptable wall thickness 
for normal, upset, and postulated accident condition loading per the 
ASME Code finds that the limiting condition is established from 
normal operating conditions which then bounds the upset and accident 
condition values. Allowance for non-destructive examination and 
growth of existing sleeve wall degradation must be made when 
determining the sleeve plugging limit. The proposed plugging limit 
is 40% through wall degradation. The sleeve assembly will be 
examined by state of the art non-destructive examination techniques 
on a periodic basis to provide early indication of sleeve 
degradation. The corrosion resistance of the Alloy 690 sleeve has 
been verified by field experience at Prairie Island. The oldest 
Alloy 690 sleeves were installed May 1987. No indication of 
corrosion of the sleeve or the parent tube in the weld joint has 
been identified by state-of-the-art eddy current examination. These 
oldest sleeve welds did not receive post weld heat treatment. In 
addition, 5 sleeves were removed for destructive examination in 
February, 1996. No corrosion was found in any of these sleeves 
including those dating from October 1992. The pulled sleeves had 
received post weld heat treatment. Post weld heat treatment can be 
optionally applied to the free span sleeve weld joints to reduce the 
susceptibility of the weld joint and parent tube to stress corrosion 
cracking. Since the sleeve design meets the requirements of the ASME 
code and mechanical tests have demonstrated margins above acceptance 
criteria, the installation of the Combustion Engineering leak tight 
sleeves will not increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    Installation of sleeves does not introduce any significant 
changes to the plant design basis. The use of a sleeve to span a 
degraded region of steam generator tubing restores the structural 
and leakage integrity of the tubing to meet the original design 
bases. Stress and fatigue analysis of the sleeve assembly shows that 
the requirements for ASME Code are met. Mechanical testing has 
demonstrated that margin exists above the design criteria. Any 
hypothetical accident as a result of any degradation in the sleeved 
tube would be bounded by the existing tube rupture accident 
analysis.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The use of the sleeves to repair degraded steam generator tubing 
has been demonstrated to maintain the integrity of the tube bundle 
commensurate with the requirements of the ASME Code and draft 
Regulatory Guide 1.121 and to maintain the primary to secondary 
pressure boundary under normal and postulated accident conditions. 
The safety factors used in the verification of the strength of the 
sleeve assembly are consistent with the safety factors in the ASME 
Boiler and Pressure Vessel Code used in steam generator design. The 
operational and faulted condition stresses and cumulative fatigue 
usage are bounded by the ASME Code requirements. The sleeve assembly 
has been verified by testing to prevent both tube pullout and 
significant leakage during normal and postulated accident 
conditions. A test program was conducted to ensure the rolled joint 
design for the lower joint in the tubesheet sleeve was leak tight 
and capable of withstanding the designs loads. The primary coolant 
pressure boundary of the sleeve assembly will be periodically 
inspected by non-destructive examination to identify sleeve 
degradation due to operation. Installation of sleeves will decrease 
the number of tubes which must be taken out of service. There is a 
small amount of primary coolant flow reduction due to sleeves for 
which an equivalent plugging sleeve to plug ratio is assigned and is 
used to assess the final equivalent plugging percentage used as an 
input to other safety analyses. Because the sleeve maintains the 
design basis requirements for the steam generator tubing, 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

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: May 15, 1997
    Description of amendment requests: The proposed amendments would 
change the Technical Specifications (TS) to revise certain limitations 
on reactor coolant system leakage and steam generator tube 
surveillance. The proposed changes would implement a voltage-based 
repair criteria per the requirements of NRC Generic Letter 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
Affected by Outside Diameter Stress Corrosion Cracking.'' In addition, 
a typographical error in TS Section 4.12.c. is being corrected.
    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[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The supporting technical evaluation and safety evaluation for 
the voltage based repair criteria demonstrate that steam generator 
tube structural and leakage integrity under normal operating and 
accident conditions will be maintained. Tube burst criteria are 
inherently satisfied during normal operating conditions due to the 
proximity of the tube support plate (TSP). Test data referenced in 
Generic Letter 95-05 indicates that tube burst cannot occur within 
the TSP, even for tubes which have 100% throughwall electric 
discharge machining notches, 0.75 inch long, provided that the TSP 
is adjacent to the notched area. Since tube-to-TSP proximity 
precludes tube burst during normal operating conditions, use of the 
criteria must retain tube integrity characteristics which maintain a 
margin of safety of 1.43 times the bounding faulted condition, main 
steamline break (MSLB) pressure differential. The Regulatory Guide 
(RG) 1.121 [<gr-thn-eq>Bases for Plugging Degraded PWR [Pressurized-
Water Reactor] Steam Generator Tubes<gr-thn-eq>] criterion requiring 
maintenance of a safety factor of 1.43 times the MSLB pressure 
differential on tube burst

[[Page 43372]]

is satisfied by 7/8'' diameter tubing with bobbin coil indications 
with signal amplitudes less than the current 8.7 volts structural 
limit, regardless of the indicated depth measurement.
    The upper voltage repair limit (V<INF>URL</INF>) will be 
determined prior to each outage using the most recently NRC approved 
database to determine the tube structural limit (V<INF>SL</INF>). 
The structural limit is reduced by allowances for nondestructive 
examination (NDE) uncertainty (V<INF>NDE</INF>) and growth 
(V<INF>GR</INF>) to establish V<INF>URL</INF>. Using the Generic 
Letter (GL) 95-05 NDE and growth allowances for an example, the NDE 
uncertainty component of 20% and a voltage growth allowance of 30% 
per full power year can be utilized to establish a V<INF>URL</INF> 
of 5.2 volts.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated MSLB 
outside of containment but upstream of the main steam isolation 
valve (MSIV) represents the most limiting radiological conditions to 
the plugging criteria. In support of [the] implementation of the 
revised plugging limit, analyses will be performed to determine 
whether the distribution of cracking indications at the tube support 
plate intersections during future cycles are projected to be such 
that primary-to secondary leakage would result in postulated off 
site and control room doses exceeding the limits established for 
application of the voltage-based repair criteria at Prairie Island. 
A separate calculation has determined the maximum allowable MSLB 
leakage limit in a faulted loop. This limit was calculated using the 
technical specification reactor coolant system (RCS) Iodine-131 
activity level of 1.0 microcuries per gram dose equivalent Iodine-
131 and the recommended Iodine-131 transient spiking values 
consistent with NUREG-0800 [<gr-thn-eq>Standard Review 
Plan<gr-thn-eq>]. The projected MSLB leak rate calculation 
methodology prescribed in Section 2.b of Generic Letter 95-05 will 
be used to calculate the end-of-cycle (EOC) leakage. Projected EOC 
voltage distribution will be developed using the most recent EOC 
eddy current results and considering an appropriate voltage 
measurement uncertainty and indication growth allowance. The log-
logistic probability of leakage correlation will be used to 
establish the MSLB leak rate used for comparison with the faulted 
loop allowable limit. Therefore, as implementation of the voltage-
based repair criteria does not adversely affect steam generator tube 
integrity and implementation will be shown to result in acceptable 
dose consequences, the proposed amendment[s] [do] not result in any 
increase in the probability or consequences of an accident 
previously evaluated in the Updated Safety Analysis Report (USAR).
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    Implementation of the proposed steam generator tube voltage-
based repair criteria does not introduce any significant changes to 
the plant design basis. Use of the voltage-based repair criteria 
does not provide a mechanism which could result in an accident 
outside of the region of the tube support plate elevations since 
tubes with outside diameter stress corrosion cracking (ODSCC) not 
occurring inside the thickness of the tube support plates will be 
plugged or repaired. Neither a single or multiple tube rupture event 
would be expected during all plant conditions in a steam generator 
in which the voltage based repair limit has been applied.
    Northern States Power will implement a maximum primary-to-
secondary leak rate limit of 150 gpd [gallons per day] per steam 
generator to help preclude the potential for excessive leakage 
during all plant conditions. The Regulatory Guide 1.121 criterion 
for establishing operational leak rate limits that require plant 
shutdown are based upon leak-before-break considerations to detect a 
free span crack before potential tube rupture during faulted plant 
conditions. The 150 gpd limit provides for leakage detection and 
plant shutdown in the event of the occurrence of an unexpected 
single crack resulting in leakage that is associated with the 
longest permissible crack length.
    The operational leakage limit will be reduced to 150 gpd limit 
consistent with Generic Letter 95-05. This limit is expected to 
provide for plant shutdown prior to reaching critical lengths for 
MSLB conditions using the lower 95% leak rate data. Additionally, 
this leak-before-break evaluation assumes that the entire crevice 
area is uncovered during blowdown. Partial uncover will provide 
benefit to the burst capacity of the intersection and only a small 
percentage of the TSPs are deflected greater than the TSP thickness 
during a postulated MSLB.
    As steam generator tube integrity upon implementation of the 
voltage-based repair criteria continues to be maintained through 
inservice inspection and primary-to secondary leakage monitoring, 
the possibility of a new or different kind of accident from any 
accident previously evaluated is not created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The use of the voltage-based repair criteria at Prairie Island 
maintains steam generator tube integrity commensurate with the 
criteria of the ASME [American Society of Mechanical Engineers] Code 
and Regulatory Guide 1.121. Regulatory Guide 1.121 describes a 
method acceptable to the Commission for meeting GDCs [General Design 
Criteria] 14, 15, 30, 31, and 32 by reducing the probability or the 
consequences of steam generator tube rupture. This is accomplished 
by determining the limiting conditions of degradation of steam 
generator tubing, as established by inservice inspection, for which 
tubes with unacceptable cracking should be repaired or removed from 
service. Upon implementation of the proposed criteria, even under 
the worst case conditions, the occurrence of ODSCC at the tube 
support plate elevations is not expected to lead to the steam 
generator tube rupture event during normal or faulted plant 
conditions. The EOC distribution of crack indications at the tube 
support plate elevations will be confirmed to result in acceptable 
primary-to-secondary leakage during all plant conditions in order to 
assure that radiological consequences meet the requirements of 
Generic Letter 95-05.
    Previous evaluations have indicated a potential for tube 
deformation and collapse during a postulated loss-of-coolant 
accident (LOCA) plus safe-shutdown-earthquake (SSE) event. The tube 
collapse potential arises from TSP deformation at the support plate 
wedges. Evaluation of the Westinghouse umbrella seismic spectra 
provided in Westinghouse letter NSP-92-152 for Model 51 steam 
generators shows that Prairie Island is bounded by those spectra and 
that no tubes will undergo deformation due to the combined effects 
of LOCA plus SSE. Therefore, no tubes need to be excluded from 
application of the voltage based criteria due to deformation 
resulting from combined LOCA plus SSE loadings. Addressing 
Regulatory Guide 1.83 [<gr-thn-eq>Inservice Inspection of 
Pressurized Water Reactor Steam Generator Tubes<gr-thn-eq>] 
considerations, implementation of the voltage-based repair criteria 
is supplemented by enhanced eddy current inspection guidelines to 
provide consistency in voltage normalization, by an extensive bobbin 
coil inspection which will include 100% of the hot leg TSP 
intersections and cold leg intersections down to the lowest cold leg 
TSP with known ODSCC, by the determination of the TSPs having ODSCC 
using at least 20% random sampling of tubes inspected over their 
full length, and by rotating pancake coil inspection (or equivalent) 
requirements for the larger indications left in service to 
characterize the principal degradation as ODSCC.
    As noted previously, implementation of the tube support plate 
intersection voltage-based repair criteria will decrease the number 
of tubes which must be repaired. The installation of steam generator 
tube plugs or sleeves reduces the RCS flow margin. Thus, 
implementation of the voltage-based repair criteria will maintain 
the margin of flow that would otherwise be reduced in the event of 
increased tube plugging.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the USAR or any 
Bases of the plant Technical Specifications.
    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. In addition, the proposed correction to a typographical 
error has no effect on the three standards of 10

[[Page 43373]]

CFR 50.92(c). 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

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

    Date of application for amendments: June 4, 1997
    Description of amendment request: The proposed Technical 
Specifications (TSs) amendment revises TS Surveillance Requirement 
3.8.2.1 to no longer require that automatic emergency diesel generator 
(EDG) auto-start and trip bypass features must be functional when the 
emergency core cooling system (ECCS) is not required to be operable.
    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 to the facility does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change will eliminate an inconsistency between 
Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying 
that the EDG auto-start and EDG trip bypass on ECCS initiation 
capability is not required during periods in which ECCS is not 
required to be OPERABLE. No physical changes to the facility will be 
made per this change. The systems, structures, and components 
affected by this change are considered to be accident mitigators and 
not accident initiators. The affected systems, structures, and 
components will continue to operate within the current design 
parameters. The ability of the EDGs to auto-start on a loss of 
offsite power or degraded voltage will remain unchanged. No new 
failure modes or conditions adverse to safety will be created as a 
result of this change. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change to the facility does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change will eliminate an inconsistency between 
Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying 
that the EDG auto-start and EDG trip bypass on ECCS initiation 
capability is not required during periods in which ECCS is not 
required to be OPERABLE. No physical changes to the facility will be 
made per this change. The systems, structures and components 
affected are considered to be accident mitigators not accident 
initiators. The affected systems, structures and components will 
continue to operate within the current design parameters. No new 
failure modes or conditions adverse to safety will be created as a 
result of this change. The plant conditions which do not require any 
ECCS to be OPERABLE, (i.e., the plant in MODE 5, the spent fuel 
storage pool gates are removed, water level is greater than or equal 
to 458 inches above reactor pressure vessel instrument zero, and 
there are no OPDRVs [operations with the potential of draining the 
reactor vessel] in progress) ensure sufficient coolant inventory to 
allow operator action to prevent uncovering the fuel. The ability of 
the EDGs to auto-start on a loss of offsite power or degraded 
voltage will remain unchanged. Therefore, the proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    3. The proposed change to the facility does not involve a 
significant reduction in a margin of safety.
    The proposed change will eliminate an inconsistency between 
Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying 
that the EDG auto-start and EDG trip bypass on ECCS initiation 
capability is not required during periods in which ECCS is not 
required to be OPERABLE. The ECCS and EDGs capability to perform the 
required safety functions as described/required in the bases of the 
current plant Technical Specifications will be maintained. 
Therefore, the proposed change to the facility does not result in a 
significant reduction in any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101
    NRC Project Director: John F. Stolz

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit 
No. 3, YorkCounty, Pennsylvania

    Date of application for amendment: June 30, 1997
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) 2.1.1.2 safety limit minimum 
critical power ratios (SLMCPRs) to be consistent with the use of GE 13 
fuel in the Unit 3 core for operating cycle 12.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The derivation of the cycle-specific SLMCPRs for incorporation 
into the TS, and its use to determine cycle-specific thermal limits, 
have been performed using the methodology discussed in ``General 
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, 
and U.S. Supplement, NEDE-24011-P-A-13-US, August, 1996, and the 
``Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-
A (GESTAR II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 
was submitted by GENE to the U.S. Nuclear Regulatory Commission 
(USNRC) on December 13, 1996. This change in SLMCPRs cannot increase 
the probability or severity of an accident.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling and fuel damage in the event of a 
postulated accident. The fuel licensing acceptance criteria for the 
SLMCPR calculation apply to PBAPS, Unit 3, Cycle 12 in the same 
manner as they have applied previously. The probability of fuel 
damage is not increased. Therefore, the proposed TS changes do not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR is a TS numerical value, designed to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core during the limiting postulated accident. It cannot create the 
possibility of any new type of accident. The new SLMCPRs are 
calculated using methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, and U.S. 
Supplement, NEDE-24011-P-A-13-US, August, 1996, and the ``Proposed

[[Page 43374]]

Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR 
II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was 
submitted by GENE to the U.S. Nuclear Regulatory Commission (USNRC) 
on December 13, 1996.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident, from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS Bases will remain the 
same. The new SLMCPRs are calculated using methodology discussed in 
``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-13, and U.S. Supplement, NEDE-24011-P-A-13-US, August, 
1996, and the ``Proposed Amendment 25 to GE Licensing Topical Report 
NEDE-24011-P-A (GESTAR II) on Cycle Specific Safety Limit MCPR.'' 
Amendment 25 was submitted by GENE to the U.S. Nuclear Regulatory 
Commission (USNRC) on December 13, 1996. The fuel licensing 
acceptance criteria for the calculation of the SLMCPR apply to PBAPS 
[Peach Bottom Atomic Power Station], Unit 3 Cycle 12 in the same 
manner as they have applied previously. The SLMCPRs ensure that 
greater than 99.9% of all fuel rods in the core will avoid 
transition boiling if the limit is not violated, thereby preserving 
the fuel cladding integrity. Therefore, the proposed TS changes do 
not involve a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: April 14, 1997
    Description of amendment request: The proposed amendment revises 
Appendix A, Section 6 of the Technical Specifications. The changes will 
enable Safety Review Committee (SRC) to review plant staff performance 
by deleting the plant staff performance requirement from Section 
6.5.2.9.b and incorporating a plant staff review requirement in Section 
6.5.2.8. The amendment also replaces the position title of Vice 
President (VP) Regulatory Affairs and Special Projects (RASP) with 
Director of Regulatory Affairs and Special Projects.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response:
    This amendment application does not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed. The proposed changes allow the SRC to perform a 
review, rather than an audit, of plant staff performance. This 
change does not diminish the SRC's effectiveness. A review of the 
1995 QA [quality assurance] audit of plant staff performance shows 
that no findings related to plant staff performance were issued. 
This indicates that the other review mechanisms currently in place 
are sufficient to ensure that plant staff performance is monitored.
    The position title change of VP-RASP to Director-RASP is an 
administrative change as all previously performed functions are 
being maintained. Therefore, the proposed changes do not affect the 
probability or consequences of any previously analyzed accident.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    This amendment application does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed changes affect an SRC audit requirement and 
a management position title. These changes do not affect plant 
equipment or the way the plant operates. Therefore, they cannot 
create a new or different kind of accident.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    This amendment application does not involve a significant 
reduction in a margin of safety. The requested Technical 
Specification revisions require the SRC to review rather than audit 
facility staff performance and will not diminish the effectiveness 
of the SRC. A review of the 1995 audit confirms that performance of 
the annual audit is redundant as no findings or recommendations 
concerning plant staff performance were made. The QA/ORG quarterly 
trend reports and SRC review of facility staff performance are 
adequate to ensure that plant staff performance is properly 
monitored.
    The position title change (VP-RASP to Director-RASP) is an 
administrative change as all previously performed functions are 
being maintained. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposed to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019
    NRC Project Director: Alexander W. Dromerick, Acting Project 
Director

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

    Date of amendment request: March 31, 1997, as supplemented by 
letter dated July 16, 1997. The July 16, 1997, supplement supersedes 
the March 31, 1997 application.
    Description of amendment request: The proposed amendment would 
provide changes to Technical Specification (TS) 2.1.2, ``THERMAL POWER, 
High Pressure and High Flow,'' ACTION a.1.c for TS 3.4.1.1, 
``Recirculation Loops,'' and the Bases for TS 2.1, ``Safety Limits.'' 
These changes are being made to implement an appropriately conservative 
Safety Limit Minimum Critical Power Ratio, to include Cycle 8 specific 
analyses, for all Hope Creek core and fuel designs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The derivation of the revised SLMCPRs for Hope Creek for 
incorporation into the Technical Specifications, and its use to 
determine cyclespecific thermal limits, have been performed using 
NRC approved methods. Additionally, interim implementing procedures 
which incorporate cyclespecific parameters have been used which 
result in a more restrictive value for SLMCPR. These calculations do 
not change the method of operating the plant and have no effect on 
the probability of an accident initiating event or transient.
    There are no significant increases in the consequences of an 
accident previously evaluated. The basis of the MCPR Safety Limit is 
to ensure that no mechanistic fuel damage is calculated to occur if 
the limit is not violated. The new SLMCPRs preserve the

[[Page 43375]]

existing margin to transition boiling and the probability of fuel 
damage is not increased. Therefore, the proposed change does not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes contained in this submittal result from an 
analysis of the Cycle 7 and Cycle 8 core reloads using the same fuel 
types as previous cycles. These changes do not involve any new 
method for operating the facility and do not involve any facility 
modifications. No new initiating events or transients result from 
these changes. Therefore, the proposed Technical Specification 
changes do not create the possibility of a new or different kind of 
accident, from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety as defined in the Technical Specification 
bases will remain the same. The new SLMCPRs are calculated using NRC 
approved methods which are in accordance with the current fuel 
design and licensing criteria. Additionally, interim implementing 
procedures, which incorporate cyclespecific parameters, have been 
used. The MCPR Safety Limit remains high enough to ensure that 
greater than 99.9% of all fuel rods in the core will avoid 
transition boiling if the limit is not violated, thereby preserving 
the fuel cladding integrity. Therefore, the proposed Technical 
Specification changes do not involve a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070
    Attorney for licensee: J. J. Keenan, Esquire, Nuclear Business Unit 
- N21, P.O. Box 236, Hancocks Bridge, NJ 08038
    NRC Project Director: John F. Stolz

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

    Date of amendment request: April 1, 1997, as supplemented by letter 
dated May 30, 1997
    Description of amendment request: The proposed amendment would 
provide changes to Technical Specifications (TSs) 4.6.1.1, ``Primary 
Containment Integrity,'' 3/4.6.1.2, ``Primary Containment Leakage,'' 3/
4.6.1.3, ``Primary Containment Air Locks,'' 4.6.1.5.1, ``Primary 
Containment Structural Integrity,'' and 4.6.1.8.2, ``Drywell and 
Suppression Chamber Purge System.'' The amendment would also change the 
Bases for 3/4.6.1.2, ``Primary Containment Leakage,'' 3/4.6.1.3, 
``Primary Containment Air Locks,'' 3.4.6.1.5, ``Primary Containment 
Structural Integrity,'' Section 6, ``Administrative Controls,'' and 
License Condition 2.D of Facility Operating License NPF-57. A new TS, 
6.8.4.e, ``Primary Containment Leakage Rate Testing Program,'' would be 
added. These changes modify the TSs and the Facility Operating License 
to adopt the performance based containment leak rate testing 
requirements (Option B) of 10 CFR Part 50, Appendix J.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Containment leak rate testing is not an initiator of any 
accident. The proposed changes do not make any physical changes to 
the containment and do not affect reactor operations or the accident 
analyses. Therefore, the proposed changes do not involve a 
significant increase in the probability of any previously evaluated 
accident.
    Since the allowable leakage rate is not being changed and since 
the analysis documented in NUREG-1493, ``Performance-Based 
Containment Leak-Test Program'' concludes that the impact on public 
health and safety due to extended intervals is negligible, the 
proposed changes will not involve a significant increase in the 
consequences of any previously evaluated accident.
    Therefore, adoption of a performance-based leakage testing 
requirements will provide an equivalent level of safety and 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.
    No physical changes are being made to the plant, nor are there 
any changes being made to the operation of the plant as a result of 
the proposed changes. In addition, no new failure modes of plant 
equipment previously evaluated are being introduced.
    Therefore, the proposed amendment 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 proposed changes are based on NRC-accepted provisions and 
maintain adequate levels of reliability of containment integrity. 
The performance-based approach to leakage rate testing recognizes 
that historically good results of containment testing provide 
appropriate assurance of future containment integrity. This supports 
the conclusion that the impact on the health and safety of the 
public as a result of extended test intervals is negligible. Since 
the analysis documented in NUREG-1493 confirms that the performance 
based schedule continues to maintain a minimal impact on public 
risk, it can be concluded that the margin of safety is not 
significantly affected by the proposed changes.
    Therefore, the proposed amendment 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit - N21, P. O. Box 236, Hancocks Bridge, New Jersey 08038
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 3, 1997
    Description of amendment request: The proposed amendment would 
change Technical Specification Table 3.6.3-1, ``Primary Containment 
Isolation Valves'' to add valves to the list, therein.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The accidents previously evaluated in the UFSAR [Updated Final 
Safety Analysis Report] that could be possibly affected by this 
proposal are those involving loss of coolant scenarios such as a 
piping or instrument line break. The proposed relief valves, 
associated piping and the affected portions of containment 
penetration piping are not initiators of those accidents evaluated 
in the UFSAR. The proposed relief valves limit the post-accident 
maximum expected pressures of the affected piping segments within 
ASME [American Society of Mechanical Engineers] code allowables and 
system design pressures. The modification does not cause any system 
or component to be operated outside of their design rating

[[Page 43376]]

allowed by applicable codes. The proposed relief valves will be 
safety-related and Seismic Category I components (except for the 
relief valve discharge piping, which will be non-safety related and 
seismically analyzed, and will meet the design, material and 
construction standards applicable to the affected piping 
segments[)].
    The proposed modifications do not jeopardize the capability of 
the containment isolation valves in the affected penetrations to 
close on the receipt of a containment isolation signal or to 
mitigate the consequences of design basis accidents evaluated in the 
UFSAR. Although the modifications will result in system pressures to 
be above their currently established design values, the new peak 
operating pressures of the affected piping segments will be limited 
to within the requirements of the ASME code. The modification will 
not alter any assumptions previously made or change, degrade, or 
prevent actions described in or assumed in evaluating the 
radiological consequences of the postulated design basis accidents. 
Containment structure temperature and pressure limits will not be 
exceeded with this modification and the offsite dose consequences 
will not be affected.
    Therefore these changes will not significantly increase the 
probability of an accident previously evaluated, nor involve a 
significant increase in the 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.
    Accidents or malfunctions of equipment important to safety 
previously evaluated in the UFSAR relating to the proposed 
modification involve the single active failure of a containment 
isolation valve to close upon receipt of a containment isolation 
signal or its failure to limit the containment bypass leakage 
following its closure. The proposed modification: 1) does not impact 
the automatic closure times of the containment isolation valves; 2) 
does not impact their capability to maintain leak tightness during a 
postulated design basis accident; and 3) does not adversely impact 
the manner in which any system is operated. The proposed 
modification does not compromise the UFSAR accident analysis 
assumptions and/or limits. The licensing basis safety analysis 
limits for all systems important to safety continue to be met. 
Furthermore, there is no change in plant testing proposed in this 
change request which could initiate an event. Therefore, these 
changes will 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 modifications and Technical Specification changes 
do not change the design limits, acceptance criteria or accident 
analysis assumptions pertaining to the containment isolation valves, 
their associated piping or any other safety-related systems, 
structures or components. The proposed modification does not impact 
the automatic closure times of the containment isolation valves, nor 
does it impact their capability to maintain leak tightness during a 
postulated design basis accident. For the systems affected by these 
penetration modifications, there is no change in system function or 
structural integrity introduced with these proposed changes. 
Therefore, the changes contained in this request do not result in a 
significant reduction in a margin of safety for the containment 
isolation capability of Hope Creek.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070
    Attorney for licensee: J. J. Keenan, Esquire, Nuclear Business Unit 
- N21, P.O. Box 236, Hancocks Bridge, NJ 08038
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 7, 1997
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.8.4.2, ``Motor Operated Valves 
- Thermal Overload Protection (BYPASSED),'' to relocate the list of 
applicable valves (TS Table 3.8.4.2-1) to the Hope Creek (HC) 
Generating Station Updated Final Safety Analysis Report (UFSAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS revisions involve: 1) no hardware changes; 2) no 
changes to the operation of any systems or components in normal or 
accident operating conditions; and 3) no changes to existing 
structures, systems or components. The relocation of Technical 
Specification Table 3.8.4.2-1 to the UFSAR and existing surveillance 
procedures will continue to ensure that safety-related motor-
operated valves (MOVs) are capable of performing their intended 
safety functions. Therefore these changes will not significantly 
increase the probability of an accident previously evaluated. To the 
extent practicable, these proposed changes were developed consistent 
with the changes approved by the NRC when developing NUREG-1433, 
``Standard Technical Specifications, General Electric Plants, BWR/
4'', with the intent of having this relocated information controlled 
in other plant documents subject to 10CFR50.59 provisions. Since the 
plant systems associated with these proposed changes will still be 
capable of: 1) meeting all applicable design basis requirements; and 
2) retain the capability to mitigate the consequences of accidents 
described in the HC UFSAR, the proposed changes were determined to 
be justified. Therefore, these changes will not involve a 
significant increase in the 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.
    Relocation of Technical Specification Table 3.8.4.2-1 to the 
UFSAR will not adversely impact the operation of any safety related 
component or equipment. Since the proposed changes involve: 1) no 
hardware changes; 2) no changes to the operation of any systems or 
components; and 3) no changes to existing structures, systems or 
components, there can be no impact on the occurrence of any 
accident. To the extent practicable, these proposed changes were 
developed consistent with the changes approved by the NRC when 
developing NUREG-1433, ``Standard Technical Specifications, General 
Electric Plants, BWR/4'', with the intent of having this relocated 
information controlled in other plant documents subject to 
10CFR50.59 provisions. Furthermore, there is no change in plant 
testing proposed in this change request which could initiate an 
event. Therefore, these changes will 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.
    Relocation of Technical Specification Table 3.8.4.2-1 to the 
UFSAR is consistent, to the extent practicable, with the changes 
approved by the NRC when developing NUREG-1433, ``Standard Technical 
Specifications, General Electric Plants, BWR/4''. The MOV thermal 
overload protection table will reside in the UFSAR and will ensure 
that the associated MOVs will be capable of performing their 
intended safety functions. Any changes to this UFSAR table will be 
subject to the provisions of 10CFR50.59 and a separate safety 
evaluation would be developed to support any proposed changes that 
would subsequently be made. Therefore, the changes contained in this 
request do not result in a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070
    Attorney for licensee: J. J. Keenan, Esquire, Nuclear Business Unit 
- N21,

[[Page 43377]]

P.O. Box 236, Hancocks Bridge, NJ 08038
    NRC Project Director: John F. Stolz

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

    Date of amendment request: June 2, 1997 (TS 387)
    Description of amendment request: The proposed amendment allows 
continued plant operation with a single reactor recirculation loop in 
service. The Nuclear Regulatory Commission has previously determined 
single loop operation is generically acceptable as set forth in Generic 
Letter 86-09, ``Technical Resolution of Generic Issue B-59-(N-1) Loop 
Operation in BWRs [boiling water reactors] and PWRs [pressurized-water 
reactors].'' Single loop operation is also recognized as a standard 
mode of operation in the BWR/4 Improved Standard TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    An analysis of the limiting operational transients has been 
performed by GE [General Electric] for BFN as documented in NEDO-
24236 to demonstrate adequate margin to the Safety Limit Minimum 
Critical Power Ratio (SLMCPR). In addition, SLO [single loop 
operation] has been specified as a operating option for the 
transient and accident evaluations performed as part of the cycle-
specific core reload analyses for Units 2 and 3 which ensure that 
operating limit Minimum Critical Power Ratios (OLMCPRs) for the 
current fuel types are established that maintain required margin to 
the fuel cladding safety limit. A cycle-specific analysis with SLO 
will be performed for Unit 1 prior to restart and experience 
indicates similar results are expected as those for Units 2 and 3.
    A review of the values used in the statistical analysis used in 
the basis of the fuel cladding safety limit determined that, due to 
increased uncertainties in total core flow readings and Traversing 
In-Core Probe (TIP) readings during SLO, an increase in the SLMCPR 
of .02 is bounding when in SLO. Therefore, while operating in 
single-loop mode, an additional .02 is added to the OLMCPR which 
maintains the same margin to the fuel cladding safety limit as that 
established for two-loop operation. This is a conservative approach 
because the two-loop transients have been shown to be more severe 
than the equivalent single-loop events and, therefore, the OLMCPRs 
established for two-loop operation would always be bounding. Thus, 
the margin of safety for fuel clad integrity is assured and the 
probability or consequences associated with reactor transients is 
not increased for SLO.
    SLO results in backflow through the jet pumps in the inactive 
recirculation loop which perturbs the relationship between the core 
flow and recirculation drive flow on which the flow biased Average 
Power Range Monitor (APRM) and Rod Block Monitor (RBM) setpoint 
equations are based. To compensate, the proposed TS [Technical 
Specification] changes modify the setpoint equations to correct for 
one-loop operation. With this adjustment, the setpoint equations 
preserve the original relationship between the setpoints and the 
effective recirculation drive flow such that the consequences of a 
RWE [rod withdrawal event] in SLO are bounded by the cycle-specific 
RWE analyses. Therefore, these changes do not increase the 
probability or consequences of the RWE transient previously 
evaluated.
    Average Planar Linear Heat Generation Rate (APLHGR) limits are 
established to ensure the acceptance criteria for fuel and Emergency 
Core Cooling Systems established in 10 CFR 50.46 are met. A SLO Loss 
of Coolant Accident (LOCA) analysis was performed using the SAFER/
GESTR computer code as documented in NEDC-32484P, Revision 1, 
``Browns Ferry Nuclear Plant, Units 1, 2, and 3, SAFER/GESTR-LOCA, 
Loss-of-Coolant Accident Analysis.''
    The LOCA [loss of cooling accident] results for SLO using SAFER/
GESTR showed that, with the application of an APLHGR multiplier as 
proposed in the TS change, the LOCA peak clad temperature for SLO 
will always be lower than that for limiting design basis pipe break 
for two-loop operation. An APLHGR multiplier of 0.9 is applicable 
for all current fuel types being used. This multiplier is documented 
in each cycle-specific reload analysis and included in the COLR 
[core operating limits report]. NEDC-32484P Revision 1 also 
concludes that the design basis accident (large breaks) are more 
affected than small break sequences and, therefore, the large break 
results are bounding for SLO.
    The Recirculation Pump Seizure event in SLO was evaluated in 
NEDO-24236 and shown to be a non-limiting event. This conclusion is 
also supported by GE analyses on other BWRs.
    In summary, based on the above discussion, the proposed changes 
for SLO do not increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Although the proposed change allows extended operation in a 
configuration that was previously allowed for a limited period, 
analysis has shown (as described in item A above), that operation 
with one recirculation pump out-of-service is within existing 
analyses based on the proposed TS requirements. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to operate in single-loop recirculation mode 
has been analyzed in accordance with established transient and 
accident methodologies, and margins of safety for the design basis 
accidents and transients analyzed in Chapter 14 of the BFN UFSAR 
[updated final safety analysis report] have not been significantly 
reduced. The basis for this conclusion is outlined in item A above. 
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, 405 E. 
South Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: June 9, 1997
    Description of amendment request: The amendment proposes to update 
the Technical Specifications, Section 6.0, to add a reference to NRC-
approved methodologies which will be used to validate or generate the 
operating limits in the Vermont Yankee Core Operating Limits Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change will not involve any significant increase 
in the probability or consequences of an accident. The change 
updates the Technical Specifications to include [an] NRC approved 
method reference to allow calculation of thermal hydraulic stability 
limits. It does not affect plant operation and will not weaken or 
degrade the facility.
    2. The proposed change will not create the possibility of a new 
or different kind of accident since the change is administrative. No 
physical alterations of the plant, setpoint changes, or operating 
conditions are proposed.
    3. The proposed change will not involve a significant reduction 
in a margin of safety. The change involves an update to the 
Administrative Controls in Section 6.0 of the Technical 
Specifications by adding a reference to NRC approved methods. This 
administrative change does not alter plant safety margins.

[[Page 43378]]

    The NRC staff has reviewed the licensee's analysis and, based 
onthis 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Attorney for licensee: R. K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624
    NRC Project Director: Ronald B. Eaton, Acting

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

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

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: June 9, 1997
    Description of amendment request: The proposed amendments authorize 
a revision to the realistic dose values for the process gas system 
rupture in Section 15.0 of the Byron/Braidwood (B/B) Updated Final 
Safety Analysis Report (UFSAR). During preparation of a UFSAR change 
package, ComEd discovered that the Final Safety Analysis Report (FSAR) 
had not been updated to correct an error from the previous revision of 
the dose calculation. Since the correct dose value is greater than that 
previously reported, the consequences of the accident had increased, 
and an unreviewed safety question resulted.
    Date of publication of individual notice in Federal Register: July 
10, 1997 (62 FR 37079).
    Expiration date of individual notice: August 11, 1997 (as corrected 
(62 FR 39282)).
    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

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: June 27, 1997, as supplemented by letter 
dated July 2, 1997 The supplemental letter provided clarifying 
information and did not change the initial proposed no significant 
hazards consideration determination.
    Brief description of amendment request: These amendments clarify, 
in the technical specifications (TSs) for each unit, the methodology 
used to satisfy surveillance requirements for the laboratory analysis 
of activated carbon (charcoal) samples from the standby gas treatment 
system (SGTS) and the control room emergency outside air supply system 
(CREOASS). The specific changes are made to Sections 4.6.5.3.b.2 and 
4.6.5.3.c for the SGTS and to Sections 4.7.b.2 and 4.7.2.c for the 
CREOASS, to include a reference to American Society for Testing 
Materials (ASTM), ``Radioiodine Testing of Nuclear-Grade Gas Phase 
Adsorbents,'' ASTM D3803-79. Date of publication of individual notice 
in Federal Register: July 8, 1997 (62 FR 36580)
    Expiration date of individual notice: August 7, 1997
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: April 14, 1997
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.3.8, ``Feedwater/Main Turbine Trip System 
Actuation Instrumentation'' by changing the minimum channels required 
from three to four. This change reflects a modification that is being 
installed to add an auxiliary contact to the trip system logic. In 
addition, the amendments revise the TS action statement for inoperable 
channels to be consistent with the Improved Standard Technical 
Specifications and to account for the additional channel.
    Date of issuance: July 29, 1997
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 119 and 104
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33120). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 29, 1997. No significant 
hazards consideration comments received: No.

[[Page 43379]]

    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan

    Date of application for amendment: March 27, 1997, as supplemented 
July 7, 1997
    Brief description of amendment: The amendment revises the Palisades 
Plant license and technical specifications to reflect the licensee's 
name change from ``Consumers Power Company'' to ``Consumers Energy 
Company.''
    Date of issuance: July 21, 1997
    Effective date: July 21, 1997
    Amendment No.: 176
    Facility Operating License No. DPR-20: Amendment revised the 
license and the technical specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19828) The July 7, 1997, letter provided supplementary information 
within the scope of the original application and did not change the NRC 
staff's initial proposed no significant hazards considerations 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 21, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

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

    Date of application for amendments: May 27, 1997
    Brief description of amendments: The amendments delete Section 
4.7.13.3.a.2 of each unit's Technical Specifications, regarding the 
minimum volume and boron concentration of borated water available to 
the Standby Makeup Pump of the Standby Shutdown System.
    Date of issuance: July 21, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 160 and 152
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33121) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 21, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: February 17, 1997, as revised 
May 1, 1997.
    Brief description of amendment: Changes to Technical Specification 
(TS) to implement 10 CFR 50, Appendix J Option B relating to 
containment leakage tests.
    Date of issuance: July 24, 1997
    Effective date: July 24, 1997
    Amendment No.: 156
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1997 (62 
FR 9214), as superseded June 4, 1997 (62 FR 30632) The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated July 24, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629

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

    Date of application for amendment: April 28, 1997
    Brief description of amendment: Technical Specification (TS) 3.7.6 
requires that flood protection be provided for the service water pump 
cubicles and components when the water level exceeds a specific value. 
The amendment (1) adds the closing of the service water pump cubicle 
sump drain valves to the TS, (2) revises the wording of the action 
statement to be consistent with the limiting condition for operation, 
and (3) revises the associated Bases section.
    Date of issuance: July 28, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 144
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30636) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 28, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of application for amendment: January 23, 1997, as 
supplemented January 28, March 4, June 19, July 2, July 16 (2 letters), 
July 21, and July 25, 1997
    Brief description of amendment: The amendment documents the staff's 
review and approval of the apparent unreviewed safety questions (USQs) 
associated with (1) the updated analysis of the design-basis accident 
(DBA) containment temperature and pressure response, and (2) the 
reliance on containment pressure to compensate for the potential 
deficiency in net positive suction head (NPSH) for the emergency core 
cooling system (ECCS) pumps during a DBA with the worst case scenario 
assumptions. The amendment also authorizes the licensee to change the 
Technical Specification bases and the Updated Safety Analysis Report, 
to reflect the reliance of containment pressure to compensate for the 
potential deficiency in NPSH for the ECCS pumps following a DBA.
    Date of issuance: July 25, 1997
    Effective date: July 25, 1997. Implementation shall be as specified 
in Appendix C to the license.
    Amendment No.: 98
    Facility Operating License No. DPR-22: Amendment revised the 
license and the licensee's updated safety analysis report.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6576) The June 19, 1997, submittal, expanded the scope of the 
initial submittal dated January 23, 1997, and therefore, another notice 
was issued in Federal Register on June 24, 1997 (62 FR 34086). The July 
2, July 16 (2 letters), July 21, and July 25, 1997, submittals provided 
additional clarifying information within the scope of the application 
and did not change the NRC staff's proposed no significant hazards 
considerations determination that was based on the June 19, 1997, 
submittal. Therefore, renoticing was not warranted. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated July 25, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Minneapolis Public Library,

[[Page 43380]]

Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

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 application for amendments: June 27, 1997, as supplemented 
by letter dated July 2, 1997 The supplemental letter provided 
clarifying information and did not change the initial proposed no 
significant hazards consideration determination.
    Brief description of amendments: These amendments clarify, in the 
technical specifications (TSs) for each unit, the methodology used to 
satisfy surveillance requirements for the laboratory analysis of 
activated carbon (charcoal) samples from the standby gas treatment 
system (SGTS) and the control room emergency outside air supply system 
(CREOASS). The specific changes are made to Sections 4.6.5.3.b.2 and 
4.6.5.3.c for the SGTS and to Sections 4.7.b.2 and 4.7.2.c for the 
CREOASS, to include a reference to American Society for Testing 
Materials (ASTM), ``Radioiodine Testing of Nuclear-Grade Gas Phase 
Adsorbents,'' ASTM D3803-79.
    Date of issuance: July 30, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 167 and 141
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications. Public comments requested as to 
proposed no significant hazards consideration: Yes (62 FR 36580). That 
notice provided an opportunity to submit comments on the Commission's 
proposed no significant hazards consideration determination by July 22, 
1997. No comments have been received. The notice also provided an 
opportunity to request a hearing by August 7, 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 
amendment. On July 9, 1997, the NRC staff issued a Notice of 
Enforcement Discretion in order to delay enforcement of the current, 
subject, TS requirements until the NRC could take formal action on the 
July 2, 1997, application. The Commission's related evaluation of the 
amendments, finding of exigent circumstances, consultation with the 
State of Pennsylvania, and final no significant hazards consideration 
determination are contained in a Safety Evaluation dated July 30, 1997.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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

    Date of application for amendment: February 11, 1997.
    Brief description of amendment: This amendment changes the Hope 
Creek Technical Specification (TS) Sections 3/4.8.1, ``A.C. Sources,'' 
6.8, ``Procedures and Programs,'' and the Bases for Section 3/4.8, 
``Electrical Power Systems,'' to include: 1) the relocation of existing 
surveillance requirements related to diesel fuel oil chemistry; 2) the 
introduction of a new program under TS 6.8.4.e, ``Diesel Fuel Oil 
Testing Program<gr-thn-eq>; 3) revisions to the TS Bases for Section 3/
4.8 to incorporate information associated with the TS changes; and 4) 
editorial changes to implement required corrections.
    Date of issuance: July 24, 1997
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 100
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14469) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 24, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: March 3, 1997, as supplemented 
by letter dated May 5, 1997
    Brief description of amendment: This amendment changes Hope Creek 
TSs as follows: (1) TS 3/4.3.1, ``Reactor Protection System 
Instrumentation,'' TS 3/4.3.2, ``Isolation Actuation Instrumentation,'' 
and TS 3/4.3.3, ``Emergency Core Cooling System Actuation 
Instrumentation,'' to include additional information concerning 
response time testing; (2) TS 4.0.5 to reference inservice inspection 
and test requirements; (3) TS 3/4.6.1, ``Primary Containment,'' and 
associated Bases to reflect a design modification; (4) TS 3/4.7.7, 
``Main Turbine Bypass System,'' to specify a new operability 
requirement; and (5) the Bases for TS 3/4.8, ``Electrical Power 
Systems.''
    Date of issuance: July 24, 1997
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 101
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33131) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 24, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendments: February 11, 1997, as 
supplemented on May 1, June 12, and July 23, 1997
    Brief description of amendments: The amendments add a new Technical 
Specification, 3/4.7.10, ``Chilled Water System - Auxiliary Building 
Subsystem,'' and an associated Bases section to address the support 
function this system provides to other necessary safety systems.
    Date of issuance: July 29, 1997
    Effective date: Unit 1 to be implemented prior to entering Mode 6 
from the current unit outage; Unit 2 as of its date of issuance, to be 
implemented within 10 days of issuance.
    Amendment Nos.: 199 and 182
    Facility Operating License Nos. DPR-70 and DPR-75.: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11497) The licensee's supplemental letters provided additional 
information that did not affect the staff's proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
July 29, 1997. No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of application for amendment: October 23, 1996, as 
supplemented

[[Page 43381]]

December 11, 1996, January 31, February 10 and 24, March 11, April 4 
and 11, May 28, June 26, and July 15, 1997.
    Brief description of amendment: The amendment changes the Watts Bar 
Nuclear Plant, Unit 1, Technical Specifications (TS) to increase the 
spent fuel storage capacity from 484 fuel assemblies to 1610 fuel 
assemblies and to increase the initial enrichment of the fuel to be 
stored in the spent fuel storage racks from 3.5 weight percent (wt%) to 
5.0 wt%. This modification also changes the center-to-center spacing of 
stored fuel assemblies and reflects the use of burnup credit rack 
modules to be installed peripherally along the pool walls.
    The amendment, as proposed by the licensee, would also involve the 
installation of spent fuel racks in the spent fuel cask pit for 225 
storage spaces thus increasing the total WBN spent fuel storage 
capacity to 1835 spent fuel assemblies. The licensee proposed to 
provide an impact shield that would be placed over the fuel in the cask 
pit when heavy loads are moved near or across the cask pit area. The 
staff is continuing its review of this aspect of the licensee's 
proposal. Accordingly, this amendment authorizes the reracking and 
usage of the main spent fuel pool, as proposed for a total of 1610 
spent fuel spaces. However, it does not authorize the installation of 
storage racks or storage of spent fuel in the spent fuel cask pit. The 
staff's review of that aspect of the licensee's application will be 
addressed by further correspondence.
    Date of issuance: July 28, 1997
    Effective date: July 28, 1997
    Amendment No.: 6
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: April 2, 1997 (62 FR 
15733) The April 4, and 11, May 28, June 26 and July 15, 1997 letters 
provided clarifying informaion that did not change the initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in an environmental assessment dated April 7, 1997, and a Safety 
Evaluation dated July 28, 1997. No significant hazards consideration 
comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402

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

    Date of application for amendments: November 9, 1987, as 
supplemented March 31, 1988, June 8, 1992, and February 4, 1997
    Brief description of amendments: These amendments reformat the 
operability and surveillance requirements for the intermediate range 
channels.
    Date of issuance: July 30, 1997
    Effective date: July 30, 1997
    Amendment Nos.: 206 and 187
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33136) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 30, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: February 17, 1997
    Brief description of amendment: The amendment revises the technical 
specifications to move Table 3.6-1, ``Containment Isolation Valves'' to 
Wolf Creek Generating Station procedures. In addition, the technical 
specifications have been modified to remove all references to Table 
3.6-1. This change is in accordance with the guidance provided in 
Generic Letter 91-08, ``Removal of Component Lists from Technical 
Specifications,'' dated May 6, 1991.
    Date of issuance: July 23, 1997
    Effective date: July 23, 1997, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 108
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications and the Operating License.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19838) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 23, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Dated at Rockville, Maryland, this 6th day of August, 1997.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation.
[Doc. 97-21244 Filed 8-12-97; 8:45 am]
BILLING CODE 7590-01-F