[Federal Register Volume 59, Number 177 (Wednesday, September 14, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10914]


[[Page Unknown]]

[Federal Register: September 14, 1994]


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NUCLEAR REGULATORY COMMISSION

 

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving 
No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 22, 1994 through September 1, 1994. 
The last biweekly notice was published on August 31, 1994 (59 FR 
45015).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 
20555. The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By October 14, 1994, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC 20555 and at the local 
public document room for the particular facility involved. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
room for the particular facility involved.

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

    Date of amendments request: August 2, 1994
    Description of amendments request: The proposed amendment would 
revise Technical Specifications (TSs) 3.9.1 and 3.1.2.7 and the Bases 
to Specification 3.1.2.7. Specifically, TS 3.9.1, ``Refueling 
Operations, Boron Concentration,'' would be revised to require action 
to restore boron concentration to within its limits in place of the 
current requirement to initiate and continue boration at a rate greater 
than or equal to 40 gpm of 2300 ppm boric acid solution or its 
equivalent until the boron concentration is within its limit. TS 
3.1.2.7, ``Borated Water Sources - Shutdown,'' gives the operability 
requirement for borated water sources including the Refueling Water 
Tank (RWT), in Modes 5 and 6. The minimum boron concentration is given 
as 2300 ppm. While this minimum value is correct for Mode 5, a larger 
boron concentration may be necessary in Mode 6. The RWT is the 
preferred borated water source for restoring the required boron 
concentration as required by TS 3.9.1. Therefore, the RWT boron 
concentration in Mode 6 should be at least be that required by TS 
3.9.1. The proposed change to TS 3.1.2.7 would clarify the boron 
concentration requirements. In Mode 5, 2300 ppm will continue to be 
required. In Mode 6, the boron concentration limit for the RWT will be 
the boron concentration limits given in TS 3.9.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    During refueling operations, the reactivity condition of the 
core is maintained consistent with the initial conditions assumed 
for the boron dilution event in the accident analysis (Updated Final 
Safety Analysis Report Section 14.3) and is sufficient to ensure the 
core remains subcritical during core alterations. Technical 
Specification 3.9.1 requires that the boron concentration be 
maintained to ensure a keff [is less than or equal to] 0.95. 
Should the boron concentration drop below the Technical 
Specifications limit, the Action requires boration at a specified 
flow rate and boron concentration until the boron concentration is 
restored to within its limit. Refueling boron concentrations higher 
than the concentration specified by the Action in [Technical] 
Specification 3.9.1 are allowed by the Technical Specifications and 
clarification of the Action for that circumstance is needed. The 
proposed change eliminates the specified flow rate and boron 
concentration in the Action and substitutes a directive to 
immediately initiate action to restore the boron concentration to 
within its limits. The accident analysis does not assume a specific 
boration rate, but only assumes that the operator acts to terminate 
the dilution.
    Therefore, the consequences of the event are unchanged. In 
addition, the proposed change revises the boron concentration limit 
on the Refueling Water Tank in Mode 6 to make the boron 
concentration limit on the tank the same as the boron concentration 
limit on the reactor coolant system. This will ensure that the RWT 
will contain water of a sufficient boron concentration to respond to 
a boron dilution event.
    The proposed change does not change the boron concentration or 
shutdown margin required by [Technical] Specification 3.9.1 and 
continues to meet the initial conditions of the boron dilution 
event. Therefore, the probability of a boron dilution event is not 
increased. Furthermore, the revised action ensures that the 
appropriate actions for a boron dilution event will be taken and 
that a borated water source of sufficient concentration is available 
to respond to that event. Therefore, the consequences of a boron 
dilution event are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change does not represent a significant change in 
the configuration or operation of the plant. The proposed actions 
will results in the same operator actions as the current Technical 
Specifications. The minimum boron concentration of the Refueling 
Water Tank in Mode 6 may be increased above the current value, but 
the concentrations will be within the analyzed maximum concentration 
for that tank,
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The margin of safety provided by [Technical] Specification 3.9.1 
is to ensure that the core remains subcritical during a boron 
dilution event and during core alterations. The proposed change does 
not alter the required shutdown margin or significantly change the 
actions to be taken if that shutdown margin is lost. The proposed 
change ensures that all assumed borated water sources will have 
sufficient boron concentration to respond to boron dilution event.
    Therefore, the proposed change does not involve a significant 
reducation in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Michael J. Case

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

    Date of amendments request: August 2, 1994
    Description of amendments request:  The proposed change would 
revise Technical Specifications (TSs) regarding surveillances 
associated with the Emergency Diesel Generators (EDGs). Specifically, 
TS 4.8.1.1.2.d.3.c would be revised to add high crankcase pressure to 
the EDG trips which are verified to be automatically bypassed on a 
Safety Injection Actuation Signal (SIAS). In addition, a footnote would 
be added stating that verification of the high crankcase pressure trip 
bypass will not be required on a particular EDG until the modification 
has been completed for that 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:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The Calvert Cliffs Emergency Diesel Generators (EDGs) are used 
to provide electrical power for the operation of Engineered Safety 
Features (ESF) and safe shutdown equipment for events involving a 
loss of offsite power. The EDGs are also called upon to 
automatically start if an accident condition (SIAS) is present. In 
the event of an automatic start from a SIAS, the EDGs do not assume 
any load until the preferred, offsite power source is actually lost. 
On an undervoltage condition on a vital bus, the corresponding EDGs 
automatically start and load.
    Emergency diesel generator trips are provided to initiate engine 
shutdown during abnormal diesel-run conditions, thereby protecting 
the EDGs from any resulting damage. Under emergency conditions, EDG 
reliability is a key accident-mitigating factor; therefore, upon 
receipt of a SIAS, the EDG control logic blocks two of the normal 
shutdown signals so that the only signals remaining are those 
required to prevent rapid destruction of the diesel engine. High 
crankcase pressure is typically not an indication of impending rapid 
diesel engine failure; therefore, this trip will be added to those 
shutdown signals bypassed on a SIAS. The proposed Technical 
Specification change adds the high crankcase pressure trip as one of 
the EDG trips verified to be bypassed by a SIAS. A high crankcase 
pressure condition on one EDG will not impact either of the two 
unaffected EDGs, or any other equipment required to mitigate 
accident consequences, and satisfies the single failure criteria. 
The manufacturer concurs with the proposed change to bypass this 
trip on a SIAS. In blocking this trip on a SIAS, the ultimate effect 
is an increase in the reliability of the effected EDG, and 
therefore, no increase in the consequences of a previously evaluated 
accident.
    Additionally, the EDGs are not initiators to any previously 
evaluated accident. Therefore, blocking the high crankcase pressure 
trip on a SIAS will not increase the probability of an accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase to the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The function of the EDGs is to provide power to ESF and safe 
shutdown equipment for events involving a loss of offsite power. The 
proposed change does not represent a significant change in the 
configuration or operation of the plant; therefore, the EDGs 
continue to function in an accident mitigation role. The EDGs are 
not accident precursors, either in the current configuration, or 
following the modification to block the high crankcase pressure 
trip.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The margin of safety credited with the EDG function associated 
with this change is the reliability of the EDGs following an event 
involving a loss of offsite power. By blocking high crankcase 
pressure trips on a SIAS, this change increases the likelihood that 
an EDG will be able to supply power when it is needed most, during a 
SIAS, because the probability of an unnecessary EDG shutdown is 
decreased. In effect, the margin of safety associated with this 
function, EDG reliability, is increased.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Michael J. Case

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

    Date of amendments request: August 4, 1994
    Description of amendments request: The proposed amendment would 
eliminate Technical Specifications 3/4.3.3.3, 6.9.2.b, and 6.9.2.d and 
Bases 3/4.3.3.3 which gives requirements for seismic monitoring 
instrumentation. Specifically, the requirements for operation and 
testing of the seismic monitoring instrumentation would be relocated to 
the Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis 
Report (UFSAR) and plant procedures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change has been evaluated against the standards in 
10 CFR 50.92 and has been determined to not involve a significant 
hazards consideration, in that operation of the facility in 
accordance with the proposed amendments:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The seismic monitoring system is used to measure the seismic 
response of selected Class 1 structures, provide time-history 
records of seismic events, and would indicate if predetermined 
seismic acceleration values had been exceeded. The seismic 
monitoring system itself has no safety function. The system measures 
values which are used after the fact to assess the intensity of an 
earthquake.
    The proposed change will relocate requirements regarding the 
operability and testing of the seismic monitors from the Technical 
Specifications to the UFSAR and plant procedures. This will allow 
changes to the requirements to be made without Commission approval 
as long as the changes meet the criteria of 10 CFR 50.59. Associated 
Technical Specification Special Report requirements and Bases will 
be deleted. Changes to the seismic monitoring system requirements 
which do not meet the criteria of 10 CFR 50.59 must be approved by 
the Commission by license amendment.
    The seismic monitoring system is not an initiator and does not 
act to minimize the consequences of any accident previously 
evaluated. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated?
    The proposed relocation of seismic monitor requirements from the 
Technical Specifications to the UFSAR and plant procedures does not 
represent a change in the configuration or operation of the plant. 
The seismic monitoring system will continue to be controlled under 
10 CFR 50.59. Associated Technical Specification Special Report 
requirements and Bases will be deleted. The proposed change will not 
add any new hardware and will not introduce any new accident 
initiators. Therefore, the proposed change does not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Does operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    The seismic monitoring system is used to measure the response of 
selected Class 1 structures to seismic events. The plant is designed 
to withstand the loads imposed by the maximum hypothetical accident 
and the design seismic disturbance without loss of functions 
required for reactor shutdown and emergency core cooling. As a 
consequence, the seismic monitoring system makes no contribution to 
the margin of safety, and neither do the associated special reports.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Michael J. Case

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 amendment request: July 18, 1994
    Description of amendment request: The purpose of the proposed 
amendment is to separate the Technical Specification (TS) into two 
separate volumes, one volume explicitly for Unit 1 and one volume 
explicitly for Unit 2. At present, each unit has a single volume of TS 
which contains the specifications covering both units. In anticipation 
of the steam generator (SG) replacement project scheduled to begin in 
the fall of 1994, the licensee is requesting that the TS reflect unit 
specific data. Since the SG project outlines a schedule for single 
units, the present documentation reflecting both units in one volume 
will make it difficult to facilitate TS changes to a single unit. The 
proposed TS will modify the current situation as follows:1) The pages 
will now contain the same information as found before with the 
exception of references to different units. The Unit 1 volume will only 
contain parameter and setpoint values applicable to Unit 1; the Unit 2 
volume will only contain information applicable to Unit 2.2) The limits 
established by the TS (the definitions, the limiting conditions for 
operation, the surveillance requirements, the Bases, etc.) will be 
unchanged by this amendment, with the exception of (3) below. The 
effect of the amendment will be that the Unit 1 TS will be found only 
in the volume dedicated solely to Unit 1 and likewise for Unit 2. 3) TS 
Sections 3.0.5 and 4.0.6 will be deleted and minor editorial changes, 
such as the correction of misspellings and the deletion of obsolete 
footnotes, will be made. TS 3.0.5 and 4.0.6 define the applicability of 
the current joint TS volume to each unit individually. Since each 
unit's TS will be located in a separate volume, no statements are 
necessary to indicate differences in parameters between units and TS 
3.0.5 and 4.0.6 may be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed amendments would not involve a significant increase 
in the probability or consequences of a previously evaluated 
accident. The separation of the existing technical specification 
manual into unit-specific volumes is a strictly administrative 
process which will not affect the probability or consequence of any 
accident.
    They will not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The changes do 
not have any impact upon the design or operation of plant equipment; 
therefore, they cannot serve to initiate a new type of accident.
    The proposed amendments would not involve a reduction in a 
margin of safety. The changes would not impact the design or 
operation of any plant systems or components.
    Based upon the preceding analysis, Duke Power Company concludes 
that the proposed amendments do not involve a significant hazards 
consideration as defined by 10 CFR 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 request involves no significant hazards consideration.
    Local Public Document Room: York County Library, 138 East Black 
Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: July 18, 1994
    Description of amendment request: The purpose of the proposed 
amendment is to separate the Technical Specifications (TS) into two 
separate volumes, one volume explicitly for Unit 1 and one volume 
explicitly for Unit 2. At present, each unit has a single volume of TS 
which contains the specifications covering both units. In anticipation 
of the steam generator (SG) replacement project scheduled to begin in 
the fall of 1994, the licensee is requesting that the TS reflect unit 
specific data. Since the SG project schedules SG replacement for each 
unit at different times, the present common TS would make it difficult 
to facilitate TS changes to a single unit. The proposed amendment will 
modify the current TS as follows:1) The pages will now contain the same 
information as found before with the exception of references to 
different units. The Unit 1 volume will only contain parameter and 
setpoint values applicable to Unit 1; the Unit 2 volume will only 
contain information applicable to Unit 2.2) The limits established by 
the TS (the definitions, the limiting conditions for operation, the 
surveillance requirements, the Bases, etc.) will be unchanged by this 
amendment, with the exception of (3) below. The effect of the amendment 
will be that the Unit 1 TS will be found only in the volume dedicated 
solely to Unit 1 and likewise for Unit 2.3) TS Sections 3.0.5 and 4.0.6 
will be deleted and minor editorial changes, such as the correction of 
misspellings and the deletion of obsolete footnotes, will be made. TS 
3.0.5 and 4.0.6 define the applicability of the current joint TS volume 
to each unit individually. Since each unit's TS will be located in a 
separate volume, no statements are necessary to indicate differences in 
parameters between units and TS 3.0.5 and 4.0.6 may be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed amendments would not involve a significant increase 
in the probability or consequences of a previously evaluated 
accident. The separation of the existing technical specification 
manual into unit-specific volumes is a strictly administrative 
process which will not affect the probability or consequence of any 
accident.
    They will not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The changes do 
not have any impact upon the design or operation of plant equipment; 
therefore, they cannot serve to initiate a new type of accident.
    The proposed amendments would not involve a reduction in a 
margin of safety. The changes would not impact the design or 
operation of any plant systems or components.
    Based upon the preceding analysis, Duke Power Company concludes 
that the proposed amendments do not involve a significant hazards 
consideration as defined by 10 CFR 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 request involves no significant hazards consideration.
    Local Public Document Room: Atkins Library, University of North 
Carolina, Charlotte (UNCC Station), North Carolina 28223

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

    Date of amendment request: June 17, 1994, as supplemented by letter 
dated August 17, 1994.
    Description of amendment request: The amendment requests the 
removal of license conditions for Transamerica Delaval (TDI) Emergency 
Diesel Generators (EDGs) associated with NUREG-1216.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or the 
consequences of an accident previously evaluated:
    The proposed amendment would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Elimination of the required teardowns and inspections has 
no effect on the probability of an accident occurring, because the 
diesel generators are not accident initiating equipment. Also, 
deleting the teardowns and inspections would decrease the 
consequences of an accident because the availability of the engines 
would increase as a result of the less frequent teardowns. 
Additionally, the high average reliability of the TDI engines would 
not be negatively affected due to this change. NRC research has 
shown there is a period of decreased reliability immediately 
following intrusive teardowns, (break in period), followed by a long 
period of high reliability.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated:
    The proposed amendment would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed amendment will not cause any physical change 
to the plant or the design or operation of the diesel units.
    3. Involve a significant decrease in the margin of safety.
    The proposed amendment would not involve a significant reduction 
in a margin of safety. The proposed amendment will increase the 
reliability and availability of the EDGs and therefore will not 
result in a decrease in a margin of safety at Grand Gulf.
    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: Judge George W. Armstrong Library, Post 
Office Box 1406, S. Commerce at Washington, Natchez, Mississippi 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: August 9, 1994
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) by relocating the functions 
under review and audit to the Waterford 3 quality assurance program 
manual. The proposed change also incorporates the TS line-item-
improvement of Generic Letter 93-07, ``Modification Of The Technical 
Specification Administrative Control Requirements For Emergency And 
Security Plans,'' dated December 28, 1993. The changes are proposed to 
reduce regulatory burden by relocating TS requirements that are 
duplicated by other regulatory 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:
    The proposed change will have no affect on design bases 
accidents nor will the change directly affect any material condition 
of the plant that could directly contribute to causing or mitigating 
the effects of an accident. Relocating Review and Audit functions 
from the TS is consistent with the NRC Final Policy Statement on 
Technical Specifications Improvements and will have no negative 
impact on plant operation or safety. Therefore, the proposed change 
will not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    The proposed change will not alter the operation of the plant or 
the manner in which the plant is operated. The change will not 
involve a design change or introduce any new failure modes. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change is administrative in nature. The Waterford 3 
safety margins are defined and maintained by the Technical 
Specifications in Sections 2-5 which are unaffected. Therefore, the 
proposed change will not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room: University of New Orleans Library, 
Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of amendment request: August 16, 1994
    Description of amendment request: The proposed changes revise VEGP 
Technical Specification 3/4.7.1.1 and its bases regarding the setpoint 
tolerance for the Main Steam Safety Valves (MSSVs).
    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 setpoint tolerance change for the MSSVs from plus or minus 1% to 
+2%, -3% is intended to accommodate setpoint drift that may occur 
with these valves during plant operation. However, this change will 
not adversely affect the pressure boundary integrity or safety 
function of the valves. The increase in MSSV setpoint tolerance was 
also reviewed with respect to the accident analyses presented in the 
VEGP Final Safety Analysis Report (FSAR). The evaluation 
demonstrated that the acceptance criteria of the accident analyses 
continued to be met. Additionally, the radiological consequences 
associated with the accident analysis are unaffected by the proposed 
changes. Accordingly, since the performance and capability of the 
MSSVs will be maintained as a result of the proposed changes with no 
increase in radiological consequences, there will be no significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed changes do not involve any change to the 
configuration or method of operation of any plant equipment, and no 
new failure modes have been defined for any plant system or 
component. The design basis requirement for the MSSVs will continue 
to be met and the structural integrity of the valves will not be 
challenged. Also, the setpoint tolerance change will not adversely 
affect the capability of the MSSVs to perform their pressure relief 
function to ensure the secondary side steam design pressure is not 
exceeded. Additionally, the as-left lift setpoints following testing 
of the MSSVs will continue to be within plus or minus 1% of their 
lift settings, further ensuring their safety function capability. 
Therefore, since the function of the MSSVs is unaffected by the 
proposed changes, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. All applicable acceptance criteria associated 
with increasing the MSSV setpoint tolerance will continue to be met. 
This includes the structural integrity of the valves and the effect 
of the setpoint change on the accident analyses presented in the 
VEGP FSAR. Therefore, since the MSSVs remain in compliance with the 
appropriate codes and standards and all applicable acceptance 
criteria continue to be met, there will not be a significant 
reduction in a margin of safety.
    Based on the preceding analysis, Georgia Power Company has 
determined that the proposed changes to the VEGP Technical 
Specifications will not significantly increase the probability or 
consequences of an accident previously evaluated, create the 
possibility of a new or different kind of accident than any 
previously evaluated, or involve a significant reduction in a margin 
of safety. Therefore, the proposed changes meet the requirements of 
10 CFR 50.92(c) and do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards
    Local Public Document Room: Burke County Public Library, 412 Fourth 
Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: August 19, 1994
    Description of amendment request: The amendment updates and 
clarifies the surveillance requirements for control rod exercising and 
standby liquid control pump operability testing including the bases to 
be consistent with Generic Letter 93-05.
    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. Although the surveillance requirements are lessened by these 
proposed changes, the changes are consistent with those found 
acceptable by the NRC in GL 93-05. The proposed changes have been 
determined to be compatible with our plant operating experience. 
Based on these considerations, it is concluded that the changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed changes do not involve physical changes to the 
plant or changes in plant operating configuration. The changes only 
involve frequency of testing required to be performed. The changes 
are consistent with those found acceptable by the NRC in GL 93-05. 
Thus, it is concluded that the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Although the surveillance requirements are lessened by these 
proposed changes, the changes are consistent with those found 
acceptable by the NRC in GL 93-05. The proposed changes have been 
determined to be compatible with our plant operating experience. 
Based on these considerations, it is concluded that the changes do 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room: Ocean County Library, Reference 
Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: August 15, 1994
    Description of amendment request: The proposed amendment would 
increase the allowable main steam isolation valve (MSIV) leakage and 
delete the Technical Specifications requirements applicable to the MSIV 
leakage control system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Description of Amendment Request:
    Proposed Change 1
    This proposed change increases the allowable leak rate specified 
in Technical Specification (TS) 4.7.A.2.c.3 from 11.5 standard cubic 
feet per hour (scfh) for any one main steam isolation valve (MSIV) 
when tested at 24 psig to 100 scfh for any one MSIV with a total 
maximum pathway leakage rate of 200 scfh through all four main steam 
lines when tested at 24 psig. If an MSIV exceeds 100 scfh, it will 
be restored to less than or equal to 11.5 scfh.
    Basis for proposed no significant hazards consideration 
determination:
    1. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
proposed amendment does not involve a change to structures, 
components, or systems which would affect the probability of an 
accident previously evaluated in the DAEC Updated Final Safety 
Analysis Report (UFSAR). It results in acceptable radiological 
consequences for the design basis loss of coolant accident (LOCA) 
which was previously evaluated in the UFSAR.
    Plant specific radiological analyses have been performed to 
assess the effects of the proposed increase in the allowable MSIV 
leak rate in terms of control room, technical support center (TSC), 
and offsite doses following a postulated design basis LOCA. These 
analyses utilize the hold-up volumes of the main steam piping and 
condenser as an alternate method for treating MSIV leakage. The 
radiological analyses use standard conservative assumptions for the 
release of source terms consistent with Regulatory Guide 1.3, 
``Assumptions Used for Evaluating the Potential Radiological 
Consequences of a Loss of Coolant Accident for Boiling Water 
Reactors,'' Revision 2, dated June 1974.
    Dose contributions from the proposed MSIV leakage rate limit of 
100 scfh per MSIV (with a maximum pathway leakage rate not to exceed 
200 scfh through all four main steam lines) were calculated. The 
analysis demonstrated that the dose contributions from the proposed 
MSIV leakage rate resulted in an acceptable increase to the LOCA 
doses previously evaluated against the regulatory limits for the 
offsite, control room, and TSC doses as contained in 10 CFR 100 and 
10 CFR 50, Appendix A (General Design Criterion 19). The revised 
LOCA doses are the LOCA doses previously evaluated in the UFSAR plus 
the MSIV leakage doses calculated assuming use of the alternate 
treatment method. Table 1 of Attachment 2 shows the previously 
calculated doses and the newly calculated doses.
    It is important to note that the resulting doses are dominated 
by the organic iodine fractions which occur because of the 
conservative source term assumptions used in this analysis. For a 
total leakage rate of 200 scfh through all four main steam lines, 
more than 90 percent of the offsite, control room, and TSC iodine 
doses are due to the organic iodine from the Regulatory Guide 1.3 
source term and organic iodine converted from the elemental iodine 
deposited in main steam piping systems. If the actual iodine 
composition from the fuel release (cesium iodine) is used in the 
calculations, essentially all of this organic iodine dose would be 
eliminated.
    The TSC doses due to MSIV leakage are especially conservative. 
It is not expected that there will be any radioactive releases to 
the TSC due to MSIV leakage during the initial stages of a LOCA 
since it would take considerable time for the MSIV leakage to travel 
through the main steam lines and main steam line drain system to the 
condenser, into the turbine building, and finally to the atmosphere 
and TSC. It was conservatively estimated that the 30-day integrated 
dose to personnel in the TSC would increase by only 0.02 rem. The 
dose calculations were performed using control room occupancy 
factors specified in NUREG-0800, Standard Review Plan (SRP) Section 
6.4.
    Therefore, we conclude that the proposed change will not 
significantly increase the probability or consequences of any 
previously analyzed accidents.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any previously evaluated. The 
BWROG evaluated MSIV leakage performance and concluded that MSIV 
leakage rates up to 100 scfh will not inhibit the capability and 
isolation performance of the valves to isolate the primary 
containment. There is no new modification to the MSIVs which could 
impact their operability. The LOCA has been analyzed using the main 
steam piping and condenser as a treatment method to process MSIV 
leakage at the proposed maximum rate of 200 scfh through all four 
main steam lines. Therefore, the proposed change will not create any 
new or different kind of accident from any accident previously 
analyzed in the UFSAR.
    3. Operation of the DAEC in accordance with the proposed change 
will not involve a significant reduction in the margin of safety. 
The allowable leak rate limit specified for the MSIVs is used to 
quantify a maximum amount of bypass leakage assumed in the LOCA 
radiological analysis. Results of the analysis are evaluated against 
the dose requirements contained in 10 CFR 100 for the offsite doses 
and 10 CFR 50, Appendix A (General Design Criterion 19) for the 
control room and TSC doses.
    The margins of safety are not significantly affected because the 
dose levels remain well below the limits of 10 CFR 100 and General 
Design Criterion 19. Therefore, the proposed change does not involve 
a significant reduction in the margin of safety at the DAEC.
    Description of Amendment Request:
    Proposed Change 2
    This proposed change to delete TS 3.7.E and 4.7.E and Bases 
section 3.7.E and 4.7.E involves eliminating the MSIV leakage 
control system (LCS) requirements from the TS.
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. As currently described in the UFSAR, the LCS is manually 
initiated after a design basis LOCA occurs. Since the LCS is 
operated only after an accident has occurred, this proposed 
amendment has no effect on the probability of an accident. The 
proposed change results in acceptable radiological consequences of 
the design basis LOCA previously evaluated in the UFSAR.
    The DAEC has an inherent MSIV leakage treatment capability. IES 
Utilities Inc. proposes to use the main steam line drains and 
condenser as an alternative to the LCS. Figure 1.1 of Attachment 2 
shows the primary and alternate drain paths. The proposed primary 
drain path at DAEC employs an MSL drain downstream of the MSIVs. 
There are two motor-operated valves (MOVs) in series in this line 
between the MSL and the main condenser. Both valves must be open to 
establish the required drain path. Both MOVs will be provided with 
essential power to assure that they can be opened following the DBA 
LOCA to establish a large enough drain path to support the 
radiological analysis.
    An alternate drain path will be available to convey MSIV leakage 
to the isolated condenser if either MOV fails to open. The alternate 
drain path consists of the bypass lines around the MOVs in the 
primary drain path. This alternate path contains a ``fail open'' 
valve and a restricting orifice. Consequently, if either primary MOV 
failed to open as required, the second drain path would be available 
to convey MSIV leakage to the main condenser. Radiological dose 
calculations have been performed for this alternate path as well as 
for the primary path. The results were acceptable. IES Utilities 
Inc. will update DAEC procedures as necessary to address the 
applicable alternate leakage treatment methods.
    IES Utilities Inc. contracted with EQE Engineering Consultants 
(EQE) to confirm the seismic capability of the DAEC's main steam 
piping and condenser to serve as an alternate leakage treatment 
system. Seismic verification walkdowns were performed to assure that 
the MSLs, the steam drain lines, the condenser, and interconnecting 
piping and equipment that were not seismically analyzed fall within 
the bounds of the design characteristics of the seismic experience 
database as discussed in Section 6.7 of the BWROG report.
    The DAEC main steam lines, main steam drain lines, condenser, 
and applicable interconnecting piping and equipment, are well 
represented by the earthquake experience data demonstrating good 
seismic performance, are confirmed to exhibit excellent resistance 
to damage from a design basis earthquake and have been shown to have 
substantial margin for seismic capability. The outliers that were 
identified are discussed in Attachment 7. They have been either 
evaluated to demonstrate their acceptability as they currently 
exist, or plant modifications will be implemented to resolve the 
concerns. By taking the measures discussed in Attachment 7 to ensure 
resolution for all of the identified outliers, IES Utilities Inc. is 
assured that the damage reported for the database components should 
not occur to the DAEC main steam piping and condenser or to the 
associated support systems.
    Therefore, the proposed method for MSIV leakage treatment is 
seismically adequate to withstand the DAEC design basis earthquake 
and maintain pressure retaining integrity and serve as an acceptable 
alternative to the currently installed LCS. The capability of the 
alternate MSIV leakage treatment system to withstand the effects of 
the safe shutdown earthquake and continue to perform its intended 
function (treatment of MSIV leakage) satisfies the intent of the 
seismic requirement of Appendix A to 10 CFR 100.
    Plant specific radiological analyses have been performed to 
assess the effects of MSIV leakage in terms of control room and 
offsite doses following a postulated design basis LOCA. While not 
previously considered a requirement for the design of the LCS, dose 
calculations were also performed for the TSC. These analyses utilize 
the hold-up volumes of the main steam piping and condenser as an 
alternate treatment method for the MSIV leakage. The analysis 
demonstrates that the proposed change results in an acceptable 
increase in the radiological consequences of a LOCA previously 
evaluated in the UFSAR. The LOCA previously evaluated in the UFSAR 
is still the bounding accident; the proposed change will not involve 
a significant increase in the consequences of an accident previously 
analyzed.
    The LCS lines will be disconnected, capped and welded, ensuring 
that the integrity of the primary containment is maintained. IES 
Utilities Inc. will incorporate the alternate leakage treatment 
system into the inservice inspection (ISI) and inservice testing 
(IST) programs, consistent with program requirements.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated. The 
purpose of the LCS is to reduce the untreated MSIV leakage when 
isolation of the primary coolant system and containment are 
required. Radiological dose contributions due to MSIV leakage are 
bounded by a LOCA. The LOCA has been analyzed using the main steam 
piping and condenser as a treatment method to process MSIV leakage 
at the proposed maximum rate of 100 scfh per MSIV and 200 scfh total 
maximum pathway leakage, and determined to be within the regulatory 
requirements. The LCS lines connected to the main steam lines will 
be permanently closed to assure the primary containment integrity, 
isolation, and leak testing capability are not compromised.
    3. The proposed change to delete TS 3.7.E and 4.7.E and Bases 
section 3.7.E and 4.7.E does not involve a significant reduction in 
the margin of safety. The intended function of the LCS for treatment 
of MSIV leakage will be performed by using the more effective 
alternate path via the main steam drain lines and condenser. This 
treatment method is effective for treatment of MSIV leakage over an 
expanded leakage range. Except for the requirement to assure that 
certain valves are opened to establish a proper flow path from the 
MSIVs to the condenser and that certain valves are closed to 
establish the seismic boundary, the proposed method is passive and 
does not require any logic controls or interlocks. On the other 
hand, the LCS consists of complicated logic controls and sensitive 
equipment which must be maintained at significant cost and radiation 
exposure. The radiological effects on the margin of safety are 
discussed above for Change 1. The safety significance of the LCS in 
terms of public risk was addressed in NUREG/CR-4330 which contains 
the evaluation for eliminating the LCS and disabling the systems 
currently installed at BWRs. The conclusion was that the increased 
public risk is less than 1 percent. Therefore, the proposed change 
does not involve a significant reduction in the margin of safety at 
the DAEC.
    The various attachments referred to in the above analysis may be 
found in the licensees request for amendment dated August 15, 1994. 
This document is available in the NRC's Public Document Room located at 
the Gelman Building, 2120 L. Street, NW., Washington, DC 20555 and at 
the local public document room address below.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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: Cedar Rapids Public Library, 500 First 
Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036
    NRC Project Director: John N. Hannon

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

    Date of amendment request: June 2, 1994, as supplemented August 25, 
1994
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) to remove expired one-time 
extensions of surveillances, remove an obsolete definition of charging 
pump operability, and incorporate 11 line item improvements in 
accordance with the guidance provided in Generic Letter (GL) 93-05. 
Other editorial changes would be made to renumber some pages and delete 
the blank pages from the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated (10 
CFR 50.92(c)(1)). The expired one-time extensions were in effect to 
September 30, 1993. Since these extensions have expired and the 
appropriate surveillances were performed, the proposed changes do not 
effect the configuration, operation, or performance of any system, or 
component.
    The proposals to delete Definition 1.45, ``THE CHARGING PUMP 
OPERABILITY,'' and modify the Index to reflect this change are 
administrative changes. Definition 1.45 was applicable only for cycle 4 
operation. Northeast Nuclear Energy Company (NNECO) has completed the 
necessary modifications and no longer rely on a temporary heating 
source. Therefore, the elimination of Definition 1.45 does not involve 
a significant increase in the probability or consequences of an 
accident previously analyzed.
    The proposed changes to incorporate the recommendations of GL 93-05 
do not affect the configuration, operation or performance of the 
subject systems. Increasing the surveillance test intervals as proposed 
will reduce the number of surveillance tests and minimize the potential 
for inadvertent actuation of an engineered safety feature. The increase 
in the surveillance test intervals will enhance the operational 
effectiveness of plant personnel, by reducing the amount of time that 
the plant staff has available to perform other tasks, such as 
additional preventive maintenance. Additionally, increasing the 
surveillance test interval will reduce unnecessary wear to equipment. 
NNECO's proposals to delete pages that were intentionally left blank, 
to renumber remaining pages and renumber Sections, and modify the Index 
to reflect these changes are purely administrative and editorial 
changes. Proposals to correct typographical errors on TS pages are also 
administrative changes. These changes would not affect the 
configuration, operation, or performance of any system, structure, or 
component.
    The proposed changes do not affect the manner by which the facility 
is operated and do not change any facility design feature or equipment. 
The proposed changes involve administrative or programmatic 
requirements or merely involve editorial changes, corrections, or 
clarifications. Since there is no change to the facility or operating 
procedures, there is no affect upon the probability or consequences of 
any accident previously analyzed.
    B. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because they do not affect the manner by which the 
facility is operated and do not change any facility design feature or 
equipment which affects the operational characteristics of the 
facility. The proposed changes involve administrative or programmatic 
requirements or merely involve editorial changes, corrections, or 
clarifications.
    C. The changes do not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)) because the proposed changes do not 
affect the manner by which the facility is operated or involve 
equipment or features which affect the operational characteristics of 
the facility.
    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: Learning Resource Center, Three Rivers 
Community-Technical College, Thames Valley Campus, 574 New London 
Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 22, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to incorporate a different setpoint 
and transient methodology for determining the maximum allowable power 
range neutron flux setpoint. The changes would allow Millstone Unit 3 
to operate with a reduced number of main steam-line safety valves at a 
reduced power level, as determined by the high neutron flux setpoint.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ...The proposed changes do not involve an SHC [significant 
hazards consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Technical Specification Tables 3.7-1 and 3.7-2 are being revised 
to reflect a reduction in the maximum allowable power range neutron 
flux high setpoint with inoperable steam generator safety valves. 
The new setpoints reflect a change in the methodology for 
calculating the setpoints.
    Westinghouse has determined that under certain conditions with 
typical safety analysis assumptions, the current setpoints in Tables 
3.7-1 and 3.7-2 may not provide adequate steam generator 
overpressure protection for a Loss of Load/Turbine Trip transient at 
reduced power levels. At reduced power levels, a reactor trip may 
not be actuated early in the transient. An overtemperature delta T 
trip may not be generated since the core thermal margins are 
increased at lower power levels. The PORVs [power-operated relief 
valves] and pressurizer spray may control RCS [Reactor Coolant 
System] pressure such that a high pressurizer pressure trip isn't 
generated. The reactor would eventually trip on low steam generator 
water level, but this may not occur before steam pressure exceeds 
110% of the design value if one or more MSSVs [main steam-line 
safety valves] are inoperable.
    To address this issue, Westinghouse has developed a new method 
for determination of the required power range neutron flux high 
setpoint. The new setpoint is based upon the heat removal capability 
of the operable MSSVs, rather than the previous method based only on 
flow capacity. The new equation is shown in the proposed changes to 
the Technical Specification basis. This new method has been 
developed by Westinghouse generically and a Millstone Unit No. 3 
specific calculation has been performed. The new setpoints are being 
incorporated in this proposed Technical Specification change.
    The new method includes several conservative assumptions. The 
equation is developed assuming that the maximum number of inoperable 
MSSVs applies to each loop. For example, for four loop operation, 
the maximum allowable power range neutron flux high setpoint of 65% 
is based upon four inoperable MSSVs, one per steam generator. Thus, 
in the event that only one MSSV is inoperable, the application of 
the new setpoint is very conservative. In addition, the setpoint is 
based upon the assumption that the largest capacity MSSV is 
inoperable. For the case where one of the lower capacity MSSVs is 
inoperable, the setpoint will be conservative.
    The method of calculating the setpoint provides assurance that 
the heat removal capability of the operable MSSVs is sufficient for 
reactor power up to the power range neutron flux high setpoint 
taking into account instrument and channel uncertainties. 
Consequently, steam generator pressure will remain below 110% of 
design in the event of the limiting overpressurization transient, 
the Loss of Load/Turbine Trip.
    Reducing the power range neutron flux high setpoint and 
consequently the allowable reduced power level has no impact on the 
consequences of any other accident. In addition, since the proposed 
changes only involve a reduction in the allowable power range 
neutron flux high setpoint, and operation at a lower power level, 
they cannot affect the probability of any design basis accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Since the proposed changes just reduce the existing limit on the 
power range neutron flux high setpoint with inoperable MSSVs, the 
change cannot create the possibility for a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    The reduced setpoint provides additional assurance that the 
steam generator pressure will remain below 110% of design for the 
limiting overpressurization transient, the Loss of Load/Turbine 
Trip. Thus, the proposed changes do not reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room: Learning Resource Center, Three Rivers 
Community-Technical College, Thames Valley Campus, 574 New London 
Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, 
Connecticut, 06141-0270.
    NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket No. 50-387 Susquehanna 
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

    Date of amendment request: July 27, 1994
    Description of amendment request: By letter dated June 15, 1992, 
Pennsylvania Power and Light Company (PP&L) submitted ``Licensing 
Topical Report NE-092-001, Revision 0, Power Uprate With Increased Core 
Flow,'' for Susquehanna Steam Electric Station, Units 1 and 2. The 
report was submitted to support future amendments to the Units 1 and 2 
licenses to permit a 4.5-percent increase in reactor thermal power and 
an 8-percent increase in core flow for each unit. The initial submittal 
was revised and supplemented by letters of July 24, September 17, and 
December 18, 1992, and January 8, January 25, April 2, August 5, August 
12, and September 29, 1993. The Commission's safety evaluation on these 
submittals was issued November 30, 1993 (Letter, Thomas E. Murley, NRC, 
to Robert G. Byram, PP&L). The Commission concluded that the revised 
(Revision 2) licensing topical report adequately supports PP&L's 
proposed power uprate. The Commission also concluded that SES, Units 1 
and 2, can operate safety with the proposed 8-percent increase in core 
flow, the proposed 4.5-percent increase in reactor thermal power, the 
corresponding 5-percent increase in main turbine inlet steam flow, and 
the corresponding increases in flows, temperatures, pressures, and 
capacities required in supporting systems and components at these 
uprated conditions.This amendment will change several Technical 
Specifications sections (listed below in the no significant hazards 
consideration) for Susquehanna Steam Electric Station, Unit 1, to 
increase the licensed power level from the current 3293 MWt to a new 
limit of 3441 Mwt.
    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 following three questions are addressed for each of the 
proposed Technical Specification Changes:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Section 1.0, Definitions, Definition 1.33, Rated Thermal Power
    This change redefines Rated Thermal Power as 3441 megawatts 
thermal.
    1. No. Neither the probability (frequency of occurrence) nor 
consequences of any accident previously evaluated is significantly 
affected by the increased power level because the design and 
regulatory criteria established for plant equipment remain imposed 
for the uprated power level. The PP&L assessment to increase the 
rated thermal power level at Susquehanna SES Unit 1, followed the 
guidelines of NEDC-31879P (Generic Guidelines for General 
Electric Boiling Water Reactor Power Uprate,'' G.E. Nuclear Energy, 
June 1991). NEDC-31879P provides generic licensing criteria, 
methodology, and a defined scope of analytical and equipment review 
to be performed to demonstrate the ability to operate safely at the 
uprated power level which have been approved by the NRC. NE-092-001 
(Licensing Topical Report for Power Uprate With Increased 
Core Flow,'' Pennsylvania Power & Light Company, December 1992) 
provides the description of the power uprate licensing analysis 
methodology and the results of the evaluations performed to support 
the proposed uprated power operation consistent with the methodology 
presented in NEDC-31879P. NE-092-001 provides a description of the 
power uprate licensing analysis methodology which will be used to 
determine cycle specific thermal limits for Unit 1, Cycle 9 and 
future cycles and concludes that an uprated power level of 3441 
megawatts thermal can be achieved without significant effect on 
equipment or safety analyses.
    2. No. The methodology and results described above do not 
indicate that a possibility for a new or different kind of accident 
from any previously evaluated has been created by uprated operation.
    3. No. Based on the response to Question 1 above, the 
methodology and results do not indicate a significant reduction in a 
margin of safety.
    Section 2.1, Safety Limits
    The reference to ``rated core flow'' in Technical Specification 
2.1.1 and 2.1.2 has been replaced with a reference to actual core 
flow. The references to ``rated core flow'' have been deleted to 
avoid confusion since allowable core flow is being increased by 8%. 
10 Mlbm/hr is being used in these specifications to be consistent 
with other similar Technical Specification changes (Technical 
Specifications 3.2.2, 4.4.1.1.1.2, 4.4.1.1.2.5, 3.4.1.3 and Figure 
3.4.1.1.1-1).
    1. No. The probability and consequences of accidents previously 
evaluated are not affected by this change. The basis for Technical 
Specification 2.1.1 is that boiling transition will not occur in 
bundles if core power is less than 25% of rated thermal power, 
regardless of pressure or core flow. Consequently, the specification 
of less than 10% rated core flow is not crucial to the basis and, 
thus, the use of 10 Mlbm/hr. is acceptable and has no effect on the 
probability or consequences of a previously evaluated accident.
    For Technical Specification 2.1.2, the XN-3 critical power 
correlation is valid for pressure greater than or equal to 580 psig 
and bundle flow greater than or equal to 0.25 Mlbm/hr-ft2. As 
stated in the basis for Technical Specification 2.1.1, if vessel 
downcomer water level is above TAF [top of active fuel], and core 
power greater than 25%, bundle flows for potentially limiting 
bundles will be greater than 0.25 Mlbm/hr-ft2 due to natural 
circulation. In addition, Technical Specification 3.4.1.1.1 requires 
at least one (1) recirculation loop in operation to run in Condition 
2, which would produce a core flow in excess of 30 Mlbm/hr. 
Therefore, core flows below about 30 Mlbm/hr-ft2 are prohibited 
when the reactor is at power. Thus, the change from ``10%'' to ``10 
million lbm/hr'' is acceptable and has no effect on the probability 
or consequences of a previously evaluated accident.
    2. No. The basis for Technical Specification 2.1.1 is that 
boiling transition will not occur in bundles if core power is less 
than 25% of rated thermal power, regardless of pressure or core 
flow. The proposed change is not crucial to this basis. The XN-3 
critical power correlation is valid for pressures greater than or 
equal to 580 psig and bundle flow greater than or equal to 0.25 
Mlbm/hr-ft2. The specification is based upon vessel downcomer 
water level being above TAF and core power greater than 25% which 
yields a bundle flow for potentially limiting bundles greater than 
0.25 Mlbm/hr-ft2 due to natural circulation. Based on Technical 
Specification 3.4.1.1.1, core flows below about 30 Mlbm/hr-ft2 
are prohibited when the reactor is at power. Therefore, the change 
to a limit of 10 Mlbm/hr is acceptable and does not create the 
possibility for a new or different kind of accident from any 
accident previously evaluated.
    3. No. As explained above, the margin of safety has not been 
reduced.
    Table 2.2.1-1 (Items 2.a, 2.b, and 2.c) and Specifications 
3.2.2, 3.4.1.1.2.a.2, 3.4.1.1.2.a.3, 3.4.1.1.2.a.5.b and 3.3.6-2 
(Item 2.a.1, 2.c, and 2.d), APRM Flow Biased Setpoints and Allowable 
Values
    Although the equation for determining these setpoints does not 
change as a result of the power uprate, because the setpoints in 
these technical specifications are referenced to rated thermal 
power, the current limits do change in that the top portion of the 
operating map (power vs. reactor flow) is raised by 4.5%.
    1. No. The safety analyses contained in NE-092-001 evaluated 
operation at both uprated power with 4.5% higher rod lines and 
increased core flow. In addition, General Electric Co. has analyzed 
and received generic approval for their BWR/4 product line operation 
in the Maximum Extended Operating Domain (MEOD). Operation at the 
4.5% higher rod lines is bounded by the MEOD analysis. Additional 
justification for this small increase in the power flow operating 
range is contained in Section C.2.3 of NEDC-31984P.
    Cycle specific reload analyses will evaluate operation at the 
increased power vs. flow conditions (100% uprated power vs. 87% core 
flow to 100% uprate power vs. 108% core flow). These analyses will 
ensure that the limits established in the Core Operating Limits 
Report are applicable to rated power operation from 87% to 108% core 
flow.
    Based on the above analyses, increasing the current limits do 
not represent a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. No. The analyses described above in response to Question 1 do 
not indicate that a possibility for a new or different kind of 
accident from any previously evaluated has been created by the 
proposed change.
    3. No. Based on the response to Question 1 above, the proposed 
change does not result in a reduction in the margin of safety.
    Table 2.2.1-1, Item 3, Reactor Steam Dome Pressure - High Scram
    The reactor steam dome pressure-high scram trip setpoint and 
allowable values are being changed to less than or equal to 1087 
psig and less than or equal to 1093 psig respectively.
    1. No. This scram function is designed to terminate a pressure 
increase transient not terminated by direct scram or high flux 
scram. The nominal trip setpoint is maintained above the reactor 
vessel maximum operating pressure and the specified analytical limit 
is used in the transient analyses. The analytical limit of 1105 psig 
is used in the uprated transient analyses. The results of the 
overpressure protection analysis indicate that the peak pressure 
will remain below the 1375 psig ASME limit which meets plant 
licensing requirements. In accordance with the methodology described 
in NE-092-001, transient analyses will be performed using the 
analytic limit and the results will be incorporated into the Core 
Operating Limits Report. Therefore, this proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. No. The purpose of this scram function is to terminate a 
pressure increase transient not terminated by direct scram or high 
flux scram. The nominal trip setpoint is maintained above the 
reactor vessel maximum operating pressure and the specified 
analytical limit is used in the transient analysis. 1105 psig is 
being used as the analytical limit in the uprated transient 
analysis. The results of the overpressure protection analysis 
indicate peak pressure will remain below the ASME limit of 1375 psig 
which satisfies plant licensing requirements. Based upon that 
result, it is concluded that the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. No. The results of the overpressure protection analysis 
indicate peak pressure will remain below the 1375 psig licensing 
limit, therefore, it is concluded that the proposed change does not 
result in a significant reduction in a margin of safety.
    Specification 4.1.5.c, Standby Liquid Control System
    This specification has been revised to require SLC [Standby 
Liquid Control] pumps to develop a discharge pressure of greater 
than or equal to 1224 psig.
    1. No. The ability of the SLC system to achieve and maintain 
safe shutdown is a function of the amount of fuel in the core and is 
not directly affected by core thermal power. The SLC pump test 
discharge pressure acceptance criteria are based on the lowest 
relief valve setpoint. The lowest setpoint is being increased by 30 
psi (to 1106) due to power uprate. Operating with increased core 
flow will result in additional friction losses through the core and 
a slightly larger core differential pressure (approximately 4 psi). 
Therefore, increasing the SLC pump test discharge pressure 
acceptance criteria ensures the capability of SLC injection. The 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. No. The ability of the SLC system to achieve and maintain 
safe shutdown is a function of the amount of fuel in the core and is 
not directly affected by core thermal power. Therefore, the proposed 
change does not result in a new or different kind of accident from 
any previously evaluated.
    3. No. The ability of the SLC system to achieve and maintain 
safe shutdown is a function of the amount of fuel in the core and is 
not directly affected by core thermal power. As stated in the 
response to question 1 above, the SLC pump discharge pressure 
acceptance criteria are based upon the lowest relief valve setpoint. 
The lowest setpoint is being increased by 30 psi. As the SLC pumps 
are positive displacement pumps, the uprate will not adversely 
affect the performance of the pumps to achieve proper injection. 
Based on above, the proposed change does not result in a significant 
reduction in a margin of safety.
    Specifications 3.2.2, 4.4.1.1.1.2, 4.4.1.1.2.5, 3.4.1.3 and 
Figure 3.4.1.1.1-1, Rated Core Flow References
    Technical Specification 3.2.2 contains the definition of ``W'' 
for the flow biased APRM scram equation. The word ``rated'' is being 
deleted from the definition of ``W'' since rated core flow is being 
increased. The definition of ``W'' is not altered. The change is 
being made for editorial purposes.
    Technical Specifications 4.4.1.1.1.1.2, 4.1.1.1.2.5, 3.4.1.3, 
and Figure 3.4.1.1.1-1 specify performance requirements and limits 
for the Reactor Recirculation System. These specifications are 
referenced to the current rated core flow. The references to ``rated 
core flow'' are being replaced with actual equivalent core flows. 
The specifications are equivalent and unchanged. This change is 
being made for editorial purposes to avoid confusion since rated 
core flow is being increased. These changes are also consistent with 
the changes made in Section 2.1.
    1. No. The proposed changes are editorial and do not effect the 
probability or consequences of an accident previously evaluated.
    2. No. The proposed changes are editorial and do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. No. The proposed changes are editorial and do not involve a 
significant reduction in a margin of safety.
    Specification Table 3.3.1-1, Note (j) and Action 6, Reactor 
Protection System Instrumentation, and Table 3.3.4.2-1, Note b, End-
of-Cycle Recirculation Pump Trip System Instrumentation
    The turbine first stage pressure scram bypass at 30% power in 
Technical Specification Table 3.3.1-1, Note (j) and Table 3.3.4.2-1, 
Note (b) is revised to indicate that the uprated equivalent 
allowable value of first stage turbine pressure is 136 psig. This 
value ensures that the analytical limit of 147.7 psig, which 
represented 30% rated thermal power, is not exceeded.
    As currently written Note (j), Note (b) and Table 3.3.1-1, 
ACTION 6 are unclear and could be misinterpreted. They apply only 
when RPS scram functions and End-of-Cycle Recirculation Pump Trip on 
turbine main stop valves closure or control valve fast closure are 
not automatically bypassed. ACTION 6 provides no guidance in the 
event the bypass fails to lift when thermal power is above 30%. In 
the worst case, the action statement could be interpreted literally 
to allow full power operation with the RPS function still bypassed. 
Such operation would violate the licensing basis analysis for the 
MCPR operating limit (for the Generator Load Rejection Without 
Bypass transient), which takes credit for operation of the 
anticipatory scram on control valve fast closure at greater than 30% 
of rated thermal power.
    1. No. The revisions to Table 3.3.1-1, ACTION 6, Table 3.3.1-1, 
Note (j), and Table 3.3.4-1 Note (b) clarify the current 
requirements; they do not change their intent.
    FSAR Chapter 15 transient analyses and reload licensing analyses 
take credit for operation of the anticipatory scram function on 
turbine stop valve closure and control valve fast closure for power 
levels greater than 30% of rated thermal power. The proposed 
revision to Table 3.3.1-1, ACTION 6 provides better assurance of the 
availability of the anticipatory scram function, since the current 
specifications could be interpreted literally to allow full power 
operation with the RPS function bypassed.
    The proposed revision to Table 3.3.1-1, Note (j) and Table 
3.3.4.2-1, Note (b) does not change the operation of the RPS and 
EOC-RPT bypasses on turbine stop valve closure and control valve 
fast closure below 30% power. The turbine first stage pressure 
switches will still be calibrated in the same manner, and, by 
procedure, the reactor operator will not exceed 30% power if the 
trip bypass annunciator does not clear.
    The setpoints for the RPS and EOC-RPT bypass functions were 
selected to allow sufficient operating margin to avoid scrams during 
low power turbine generator trips. As discussed in NEDC-31894P, 
Section F4.2(c) and in Section 5.1.2.8 of NEDC 31948P, this small 
absolute setpoint increase maintains the safety basis for the 
setpoint.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. No. The changes proposed are clarifications and do not change 
specification intent. The proposed change to Table 3.3.1-1, Action 6 
provides better assurance of the availability of the anticipatory 
scram function as the specification could currently be interpreted 
to allow full power operation with the RPS function bypassed. The 
proposed changes to Table 3.3.1-1, Note (j) and Table 3.3.4-1, Note 
(b) do not change the operation of the RPS and EOC-RPT bypasses on 
turbine stop valve closure and control valve fast closure below 30% 
power. Therefore, the possibility for a new or different kind of 
accident is not created.
    3. No. The proposed changes are clarification and do not change 
intent. Operation of the RPS and EOC-RPT bypasses on turbine stop 
valve closure and control valve fast closure below 30% power is not 
changed. Therefore, there is no reduction in the margin of safety.
    Specification Table 3.3.2-2, Item 3.d, Main Steam Line Flow 
Differential Pressure Setpoint
    The main steam line flow high differential pressure setpoint and 
allowable value are revised to read trip setpoint and allowable 
values of 113 psid and 121 psid respectively. Footnote ``**'' was 
added to Table 3.3.2-2 to indicate that these values will be 
confirmed during the power uprate start-up testing. If revisions to 
the setpoint and allowable value are required, they will be 
forwarded to the Commission for approval within 90 days of 
completion of the test program.
    1. No. The main steam line flow high differential pressure 
setpoint changes reflect the redefinition of rated main steam line 
flow that occurs with power uprate. The allowable value is 
maintained at the same percentage of rated steam flow as the 
differential pressure changes due to the increased uprate steam 
flow. The analytical limit of 140% of uprated steam flow is 
maintained for the uprated analyses. The relationship between the 
allowable value and the analytical limit was retained to ensure that 
a trip avoidance margin is maintained for the normal plant testing 
of MSIV's and turbine stop valves. The increase in the absolute 
value of the trip setpoint still provides a high assurance of 
isolation protection for a main steam line break accident which 
satisfies the original intent of the design. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. No. The increase in the absolute value of the trip setpoint 
still provides a high assurance of isolation protection for the main 
steam line break accident which satisfies the original intent of the 
design and, therefore does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. No. The increase in the absolute value of the trip setpoint 
still provides a high assurance of isolation protection for a main 
steam line break accident which satisfies the original intent of the 
design and, therefore, does not involve a significant reduction in a 
margin of safety.
    Specification Table 3.3.2-2, Item 4.f, Isolation Actuation 
Instrumentation Setpoints
    The RWCU system flow-high isolation trip setpoint and allowable 
value are being changed. System flow is being increased by 10% to 
maintain reactor coolant water chemistry at a level equal to pre 
uprate levels. The isolation setpoint change is being made to 
adequately maintain operating margin between normal process values 
and the isolation setpoints.
    1. No. The basis for the RWCU flow-high isolation is to ensure a 
RWCU System isolation in case of a pipe break. The high flow 
setpoint is set high enough to avoid spurious trips from normal 
operating transients but low enough to ensure an isolation during a 
pipe break. The proposed Technical Specification limits will result 
in a negligible reduction in the margin between the RWCU isolation 
setpoint and the 4350 gpm flow postulated during a RWCU line break 
and will avoid spurious isolations. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. No. As stated above, the proposed change will result in only 
a negligible reduction in the margin between the RWCU isolation 
setpoint while avoiding spurious isolation. Therefore, this change 
maintains the original design intent and does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. No. See 1. above.
    Specification Table 3.3.2-2, Items 5.a and 6.1, Isolation 
Actuation Instrumentation Setpoints
    The HPCI and RCIC Steam Line Flow-High Technical Specifications 
are being changed to account for changes in steam conditions and 
flows that result from operation at the uprated conditions. The 
setpoint and allowable value for HPCI Steam Line Flow-High isolation 
are less than or equal to 387 inches H2O setpoint and allowable 
value for the RCIC Steam Line Delta Pressure-High isolation are less 
than or equal to 188 inches H2O and less than or equal to 193 
inches H2O respectively.
    1. No. The bases for these setpoints are contained in the 
General Electric Design Specification Data Sheets for the HPCI and 
RCIC systems. The Design Specification Data Sheets specify that the 
setpoint and allowable value be set so that the isolation occurs at 
greater than 272% normal steam flow and less than 300% steam flow. 
General Electric has historically seen start-up transients as high 
as 272% of normal steam flow. Setting the isolation above this value 
prevents spurious isolations and ensures availability of the system 
and its safety function. Setting the isolation at less than or equal 
to 300% of normal flow insures that the isolation will occur if a 
steam line should rupture.
    The existing setpoints were calculated using information 
obtained during the recent surveillance tests. The revised setpoints 
and allowable values were calculated using the current system 
performance and adjusted for uprate conditions in accordance with 
additional guidance provided in General Electric Information Letter 
(SIL) No. 475, Revision 2, NEDC-31336, ``General Electric Setpoint 
Methodology,'' and GE Letter SPU-9378, ``HPCI and RCIC Steam Line 
Break Detection Setpoints''.
    Based on the above approach, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. No. The setpoint and allowable value are set so that 
isolation occurs at greater than 272% normal steam flow and less 
than 300% steam flow. Setting the isolation at less than or equal to 
300% of normal flow ensures that the isolation will occur if a steam 
line rupture should occur. Therefore, no new events are postulated 
as a result of this change.
     3. No. The proposed change does not involve a significant 
reduction in a margin of safety as the setpoint and allowable value 
are set to isolate at greater than 272% normal steam flow and less 
than 300% steam flow which are the setpoints contained in the 
General Electric Design Specification Data Sheets for the HPCI and 
RCIC systems.
    Specification Table 4.3.2.1-1, footnote ``**''
    The footnote is being changed to delete reference to reactor 
pressure.
    1. No. The original purpose of Footnote ``**'' to Technical 
Specification Table 4.3.2.1-1 was to describe the functioning of the 
permissive circuitry that allowed the MSIV low condenser pressure 
isolation to be bypassed. The original circuitry required the Mode 
Switch not be in Run, the Turbine Stop Valves closed, and reactor 
pressure to be above setpoint. In the start-up phase of the 
Susquehanna Units, General Electric deleted the reactor pressure 
setpoint input to the bypass circuitry. Therefore, this change is 
being made to make the footnote conform to the installed 
configuration. The revised footnote is the same as found in the BWR/
4 Standard Technical Specifications (NUREG 1433). This change is 
editorial in nature and, therefore, does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. No. Based on the response to Question 1 above, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. No. Based on the response to Question 1 above, the proposed 
change does not involve a significant reduction in a margin of 
safety.
    Specification Table 3.3.6-2, Item 1.a and Specification 
3.4.1.1.2.a.5.a, Rod Block Monitor Flow Biased Rod Blocks
    The Rod Block Monitor (RBM) flow biased rod blocks are being 
changed as follows:
    a. Technical Specification Table 3.3.6-2, Item 1.a is revised to 
read trip setpoint and allowable values of less than or equal to 
0.63 W + 41% and less than or equal to 0.63 W + 43%, respectively.
    b. Technical Specification 3.4.1.1.2.a.5.a is being revised to 
read trip setpoint and allowable values of less than or equal to 
0.63 W + 35% and less than or equal to 0.63 W + 37%, respectively.
    1. No. These Technical Specification changes do not represent a 
change from current limits. The change reflects the rescaling made 
necessary by the re-definition of rated thermal power.
    The RBM flow biased rod blocks are used in the Rod Withdrawal 
Error (RWE) analysis. In order to maintain Critical Power Ratio 
(CPR) margins similar to previous Susquehanna cycles, the flow 
biased rod blocks were changed in terms of megawatts thermal but the 
change was not appreciable. The rescaling of the RBM flow biased rod 
block to reflect the re-definition of Rated Thermal Power maintains 
the same level of protection as previously provided. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. No. These changes do not represent a change from current 
limits but are rather a rescaling made necessary by the re-
definition of rated thermal power.
    3. No. These changes do not represent a change from current 
limits but are rather a rescaling made necessary by the re-
definition of rated thermal power. The rescaling of the RBM flow 
biased rod block maintains the same level of protection as 
previously provided.
    Specification Table 3.3.6-2, Item 2.a, Control Rod Block 
Instrumentation Setpoints
    The APRM rod block upscale value has been changed to add a high 
flow clamp setpoint at 108% with a high flow clamped allowable value 
at 111%.
    1. No. The addition of the high flow clamp to the flow biased 
APRM rod block function maintains the normal margins between the rod 
block and the scram power levels in the increased core flow regions. 
When the reactor core flow is greater than 100 million lbm/hr, the 
APRM clamp provides an alarm to help the operator avoid scrams while 
operating in the ICF region. This action does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. No. The changes maintain the normal margins between the rod 
block and the scram power levels in ICF regions. The clamp provides 
an alarm to avoid scrams in the ICF region.
    3. No. The changes maintain the normal margins between the rod 
block and the scram power levels.
    Specification Table 3.3.6-2, Item 6.a, Reactor Coolant System 
Recirculation Flow Upscale Rod Block Setpoint and Allowable Value 
Change
    The reactor coolant system recirculation flow upscale rod block 
setpoint and allowable value are being increased to 114/125 
divisions of full scale and 117/125 divisions of full scale 
respectively.
    1. No. The Reactor Coolant System recirculation flow upscale rod 
block setpoint and allowable value are being increased to allow 
operation in the ICF region. The 114/125 divisions setpoint and 117/
125 divisions allowable value, specified by General Electric, are 
based on BWR operating history.
    The purpose of the Reactor Coolant System recirculation flow 
upscale rod block is to prevent rod movement when an abnormally high 
increase in reactor recirculation flow exists. An increase in 
reactor recirculation flow causes an increase in neutron flux that 
results in an increase in reactor power. However, this increase in 
neutron flux is monitored by the Neutron Monitoring System that can 
provide a rod block. No design basis accident or transient analysis 
takes credit for rod block signals initiated by the Reactor Coolant 
Recirculation System. Therefore, this change does not increase the 
probability or consequences of an accident previously evaluated.
    2. No. Rod block signal initiation by the Reactor Coolant 
Recirculation System is not taken credit for in the mitigation of a 
design basis accident or in any transient analysis.3. No. Rod block 
signal initiation by the Reactor Coolant Recirculation System is not 
taken credit for in any transient analysis or in the mitigation of a 
design basis accident.
    Specification 4.4.1.1.1.2 and 4.4.1.1.2.5 Reactor Coolant System
    The reactor recirculation pump motor generator set scoop tube 
electrical and mechanical overspeed stop setpoints are being 
increased to a core flow of 109.5 million lbm/hr. and 110.5 million 
lbm/hr., respectively.
    1. No. The reactor recirculation pump motor generator set scoop 
tube stops are being increased to allow operation at core flows in 
the ICF region of up to 108 million lbm/hr.
    The electrical stop is maintained above the maximum operating 
core flow and below the mechanical stop. The 109.5 million lbm/hr. 
electrical stop setpoint, specified by General Electric, is based on 
BWR operating history. The electrical stop is a system design 
feature and is not used in any safety analyses.
    The 110.5 million lbm/hr. mechanical stop setpoint is used in 
transient analysis to limit core flow during a recirculation pump 
controller failure. The 110.5 million lbm/hr. mechanical stop 
setpoint, specified by General Electric, is also based on BWR 
operating history. The cycle specific analyses, performed for power 
uprate, used the 110.5 million lbm/hr. mechanical stop setpoint.
    Based on the above, this change does not involve a significant 
increase of the probability or consequences of an accident 
previously evaluated.
    2. No. Increasing the reactor recirculation motor generator set 
scoop tube electrical and mechanical overspeed stop setpoints is 
being done to allow operation at core flows in the ICF region up to 
108 Mlbm/hr. The electrical stop setpoint is a design feature and is 
not used in any safety analysis. The mechanical stop setpoint is 
used in transient analysis to limit core flow during a recirculation 
pump controller failure. Changing of this setpoint was considered in 
appropriate transient analyses, and will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. No. See 1. above. This change does not significantly reduce 
the margin of safety.
    Specification Figure 3.4.1.1.1-1, Thermal Power Restrictions
    This figure has been redrawn to reflect the new definition of 
Rated Thermal Power to retain the same stability operating 
restrictions in terms of megawatts thermal as were previously 
described by this graph.
    1. No. The core thermal hydraulic stability curve and associated 
bases are maintained at the current rod lines and power levels. 
Those values are redefined to reflect the redefinition of rated 
thermal power. Since the current operating restrictions are 
maintained, power uprate has no detrimental effect on the level of 
protection provided by these Technical Specifications. This position 
is consistent with NEDC-31894P, Section 5.3.3 and with NEDC-31984P, 
Section 3.2.
    2. No. The core thermal hydraulic stability curve and associated 
bases are maintained at the current rod lines and power levels. 
Those values are changed to reflect the redefinition of rated 
thermal power. Since the current operating restrictions are 
maintained, power uprate has no detrimental effect on the level of 
protection provided and does not create the possibility for a new or 
different kind of accident.
    3. No. The core thermal hydraulic stability curve and associated 
bases are maintained at the current rod lines and power levels. 
Those values are redefined to reflect the redefinition of rated 
thermal power. Since the current operating restrictions are 
maintained, there is no detrimental effect on the level of 
protection provided, and therefore no significant decrease in any 
margin of safety.
    Specifications 3.4.1.1.2.5, 3.4.1.1.2.6, Reactor Coolant System, 
Recirculation Loops - Single Loop Operation
    Specification 3.4.1.1.2.5 is being renumbered to 3.4.1.1.2.6. A 
new specification 3.4.1.1.2.5 is being added to specify that a 0.70 
LHGR multiplier has been applied to Specification 3.2.4 when in 
single recirculation loop operation.
    1. No. Operation with one recirculation loop out of service is 
allowed, but it is not considered a normal mode of operation. Single 
loop operation (SLO) is a special operational condition when only 
one of the two recirculation loops is operable. In this operating 
condition, the reactor power will be limited to less than 80% of 
rated by the maximum achievable core flow, which is typically less 
than 60% of rated core flow. A postulated LOCA occurring in the 
active recirculation loop during SLO would cause a more rapid 
coastdown of the recirculation flow than would occur in two loop 
operation, where one active loop would remain intact. This rapid 
coastdown causes an earlier boiling transition and deeper 
penetration of boiling transition into the bundle, which tends to 
increase the calculated PCT. However, the PCT effects of early 
boiling transition are substantially offset by the mitigating effect 
of the lower power level achievable at the start of such an event. 
The SAFER/GESTR-LOCA analysis results for Susquehanna for SLO and 
two loop operation are well below 2200 deg.F and are documented in 
NEDC-32064P-1, Revision 1, ``Power Uprate with Increased Core Flow 
Safety Analysis for Susquehanna 1 and 2'', GE Nuclear Energy, July 
1993.
    The ECCS performance for Susquehanna under SLO was evaluated 
using SAFER/GESTR-LOCA. Calculations for the DBA were performed 
using both nominal and Appendix K inputs. The SLO SAFER/GESTR-LOCA 
analysis for the DBA assumes that there is essentially no period of 
recirculation pump coastdown. Thus, dryout is assumed to occur 
simultaneously at all axial locations of the hot bundle shortly 
after initiation of the event. Dryout is assumed to occur in one 
second for the nominal case and 0.1 second for the Appendix K case. 
These assumptions are very conservative and provide bounding results 
for the DBA under SLO.
    The two-loop Appendix K break spectrum documented in NEDC-
32064P-1 is representative of SLO because the two-loop spectrum was 
analyzed assuming a one second dryout time for all axial locations 
of the hot bundle. As shown by the two-loop break spectrum, the DBA 
is the limiting case for SLO. With breaks smaller than the DBA, 
there is a longer period of nucleate and/or film boiling prior to 
fuel uncovery to remove the fuel stored energy.
    An LHGR multiplier of 0.70 will be imposed when the plant is in 
SLO. As shown in Table 5-6 of NEDC-32064P-1, the SLO results are 
less limiting (i.e., lower PCT's) than the results for the two loop 
DBA LOCA.
    Thus, the licensing PCT is based appropriately on two loop 
operation rather than SLO.
    2. No. The licensing PCT is based upon two loop operation rather 
than SLO, thus the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. No. Based on the response to Question 1 above, the proposed 
change does not involve a significant reduction in a margin of 
safety.
    Specification 4.4.1.1.2.3, Reactor Coolant System
    Footnote **** to this Specification is being changed to 
reference the power uprate startup test program.
    1. No. This footnote provided a mechanism for changing the power 
limits specified if the results of the initial startup test program 
determined that it was necessary. The footnote is being modified to 
allow operation at uprated power with the present power limits. 
Should the power uprate startup test program determine a need to 
change the power limits they will be submitted to the Commission 
within 90 days as required by the revised footnote. This is 
consistent with the original BWR startup test program philosophy and 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. No. See 1. above; this change is administrative in nature and 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. No. See 1. above; this change is administrative in nature and 
does not involve a significant reduction in a margin of safety.
    Specification 3.4.2, Reactor Coolant system, Safety Relief 
Valves
    The safety relief valve specification is being changed to reduce 
the number of setpoint groups from 5 to 3. Two valves will be set at 
1175 psig plus or minus 1%, 6 will be set at 1195 psig plus or minus 
1%. Also, the number of Operable safety valves is being increased 
from 10 to 12.
    1. No. This change does not increase the probability of 
occurrence of an accident previously evaluated as, with one 
exception, the accidents described in FSAR Sections 5.2.2, 7.2.3, 
15.1, 15.2 and 15.3 do not document any cases where the SRV's are 
designated as the cause or initiator of an accident. The exception 
is inadvertent safety relief valve opening which results in a 
decrease in reactor coolant inventory and/or reactor coolant 
temperature. The revised setpoints and proposed groupings will not 
increase the probability of occurrence of this type of accident.
    The change does not increase the probability of occurrence of a 
malfunction of equipment important to safety as previously evaluated 
in the FSAR. The margin between peak allowable pressure and the 
maximum safety setpoint is unchanged. The reactor vessel and 
components were evaluated for the setpoint change to assure 
continued compliance with the structural requirements of the ASME 
Code. Analysis was performed on the effects of the setpoint change 
for the design conditions, the normal and upset conditions and the 
emergency and faulted conditions. The increasing RPV dome pressure 
does not affect the design condition and, therefore, stresses remain 
unchanged.
    The proposed change will also not adversely affect HPCI and RCIC 
system performance.
    There is no indication that changed setpoints contribute to an 
increase in probability of SRV malfunction. Reduction in the simmer 
margin will be compensated for by more stringent leak test 
requirements during valve refurbishment.
    2. No. This change does not involve any hardware changes or 
changes in system function. Relief and safety setpoints are only 
slightly increased and the maximum safety setpoint remains 
unchanged, thus the margin between peak allowable pressure and the 
setpoint remains unchanged.
    3. No. The technical specifications were reviewed for margins of 
safety applicable to the components and systems affected by the 
change. Analysis has been performed that demonstrates that reactor 
pressure will be limited to within ASME Section III allowable values 
for the worst case upset transient. The margin of safety is inherent 
in the ASME Section III allowable pressure values.
    Specification 3.4.3.2.d, Reactor Coolant System, Operational 
Leakage
    This specification is being revised to indicate that the 1 gpm 
leakage rate limit currently applicable applies at the uprated 
maximum allowable pressure of 1035 psig, plus or minus 10 psig.
    1. No. The steam dome pressure for leakage is being increased by 
35 psig to 1035 psig (reactor design pressure). This pressure is 
chosen on the basis of steam line pressure drop characteristics and 
excess steam flow capability of the turbine observed during plant 
operation up to the current rated power level. Increasing the 
leakage rate pressure to 1035 psig is consistent with the expected 
uprated operating pressure. Increasing the reactor steam dome 
pressure has been analyzed and found to be within allowable limits. 
Maintaining the leakage rate limit at 1 gpm does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. No. This change does not involve any hardware changes or 
change in safety function. The reactor steam dome pressure has been 
analyzed and found to be within allowable limits.
    3. No. Maintaining leakage the rate limit at 1 gpm is 
conservative and does not involve a reduction in the margin of 
safety.
    Specifications 3.4.6.2 and 4.4.6.2, Reactor Coolant System, 
Reactor Steam Dome
    The reactor steam dome pressure limits have been changed to 1050 
psig.
    1. No. Operating pressure for uprated power is increased by a 
minimum amount necessary to assure that satisfactory reactor 
pressure control is maintained. The operating pressure was chosen on 
the basis of steam line pressure drop characteristics and excess 
steam flow capability of the turbine observed during plant operation 
up to the current rated power level. Satisfactory reactor pressure 
control requires an adequate flow margin between the uprated 
operating condition and the steam flow capability of the turbine 
control valves at their maximum stroke. An operating dome pressure 
of 1032 psig is expected and is being assumed in the transient 
analyses. The 1050 psig limit was chosen to maintain an adequate 
level of operating flexibility while maintaining an adequate 
distance from the high pressure scram for trip avoidance. This limit 
is the initial pressure value used in the overpressure protection 
safety analysis for power uprate, for which all licensing criteria 
have been met. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. No. Based on the response to Question 1. above, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. No. As described in 1. above, the 1050 psig limit was chosen 
to maintain an adequate level of operating flexibility while 
maintaining an adequate distance from the high pressure scram. This 
limit is the initial pressure value used in the over pressure 
protection safety analysis for power uprate, for which all licensing 
criteria have been met. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    Specification 4.5.1.b.3, Emergency Core Cooling Systems
    This specification has been revised to permit a test line 
pressure for the flow surveillance of greater than or equal to 1140 
psig at nominal reactor operating conditions.
    1. No. Currently, the HPCI pump test acceptance criteria 
discharge pressure is greater than or equal to 1266 psig. This is 
based, in part, on the lowest SRV setpoint of 1146 psig plus a 1% 
tolerance and line flow losses. For this test, the HPCI turbine is 
supplied with steam at the nominal operating reactor pressure of 920 
+140/-20 psig. Therefore, the test requires the HPCI pump/turbine to 
produce an output that exceeds that which would be commensurate with 
the input conditions. Stated differently, HPCI would be required to 
develop a pump discharge pressure associated with a steam dome 
pressure of 1187 psig (1175 plus or minus 1% psig), while being 
supplied with a steam dome pressure as low as 900 psig.
    The purpose of this specification is to demonstrate that the 
system is capable of producing the required flow at the required 
pressure. The concern with this approach is that while it 
demonstrates the required capability by achieving the actual 
Technical Specification value, it requires the pump turbine to 
``over perform''. It also reduces the margin available to compensate 
for normal wear and tear [that] occurs and is monitored under the 
ASME Section XI Pump and Valve Test Program. Power uprate will be 
further increasing the demand because of the increase in reactor 
steam dome pressure.
    The intent of Surveillance 4.5.1b.3 is to demonstrate that the 
HPCI System will produce its design flow rate at an expected reactor 
pressure during a LOCA. Confirmation of the capability to achieve 
the required flow and pressure can be satisfactorily demonstrated 
without requiring the pump/turbine to ``over perform''. This can be 
done by producing the nominal operating design pressure from the 
pump with steam supplied to the turbine at nominal reactor operating 
pressure. From these conditions extrapolation via pump affinity laws 
will show the pump discharge pressure that would be developed at 
emergency reactor operation conditions (i.e. lowest SRV setpoint). 
This value could then be compared to the calculated value required 
for assuring adequate core cooling in both SSES specific and generic 
evaluations. The HPCI System has been evaluated and shown to be 
capable of achieving the required pressure and flow conditions for 
power uprate.
    Applying the method of pump affinity laws, the new Technical 
Specification pump discharge pressure would become greater than or 
equal to 1140 psig. This value is determined based on the maximum 
allowable test steam dome pressure of 920 + 140 = 1060 psig, plus 
head losses. Through the use of pump affinity laws it has been shown 
by calculation that achieving a value of 1140 psig at nominal 
reactor operating conditions will produce the required flow and 
pressure during emergency conditions.
    Therefore, the Technical Specification HPCI pump discharge 
pressure at power uprate conditions is changed to greater than or 
equal to 1140 psig.
    2. No. The methodology and the supporting change described above 
in the response to Question 1 above do not alter the function nor 
the operation of the HPCI system. Therefore, they do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. No. The methodology and the supporting change described above 
in response to Question 1 do not involve a significant reduction in 
a margin of safety.
    Specification 5.4.2, Design Features, Reactor Coolant System, 
Volume
    This specification is being changed to show that the nominal 
Tave is being changed from 528 deg.F to 532 deg.F. This change 
is being made to reflect the higher average saturation temperature 
that results from a 30 psi increase in reactor design pressure.
    1. No. The effects of power uprate have been evaluated to ensure 
that the increase in system temperatures causes minor increases in 
thermal loadings on pipe supports, equipment nozzles, and in-line 
components. The results of analyses show that at uprated conditions 
all ASME components will satisfy design specification requirements 
and code limits when evaluated to the rules of Subsection NB-3600 of 
the ASME Boiler and Pressure Vessel Code Section III. The effects of 
thermal expansion as a result of power uprate were found to be 
insignificant. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. No. Increases in system temperatures as a result of power 
uprate have been evaluated to show that increase in thermal loadings 
on pipe supports, equipment nozzles and in-line components are 
minor. Analysis shows that at all uprated conditions all ASME 
components will satisfy design specification requirements and code 
limits when evaluated to the rules of subsection NB-3600 of Section 
IV to the Boiler and Pressure Vessel Code. The effects of power 
uprate with respect to thermal expansion were found to be 
insignificant and, therefore, not found to create the possibility of 
a new or different kind of accident.
    3. No. As stated above, the effects of thermal expansion as a 
result of power uprate were found to be insignificant. Consequently, 
the nominal increase in Tave does not involve a significant 
reduction in a margin of safety.
    Specification Table 5.7.1-1, Component Cyclic or Transient 
Limits
    This specification is being changed to raise the upper limit for 
a heat cycle from 546 deg.F to 551 deg.F. This change is being made 
to reflect the higher average saturation temperature that results 
from a 30 psi increase in reactor design pressure.
    1. No. The purpose of this specification is to limit the number 
of heatup and cooldown cycles. The effects of power uprate have been 
evaluated to ensure that the reactor vessel components continue to 
comply with the existing structural requirements of the ASME Boiler 
and Pressure Vessel Code. The analyses were performed for the 
design, normal, upset, emergency and faulted conditions. The 
increase in the temperature limitation is not significant with 
respect to the affect it has upon the RPV and associated components.
    2. No. The effects of uprating power have been evaluated for the 
design, normal, upset, emergency and faulted conditions to ensure 
that the reactor vessel components continue to comply with the 
existing structural requirements of the ASME Boiler and Pressure 
Vessel Code. The increase in the temperature limitation has been 
found not to be significant and, therefore, does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. No. This specification is intended to limit the number of 
heatup/cooldown cycles. The increase in the temperature limitation 
has not been found to be significant with respect to its effects 
upon the RPV and its associated components and, therefore, does not 
significantly reduce the margin of safety.
    Specification 6.9.3.2, Core Operating Limits Report
    Administrative Control Section 6.9.3.2 describes and lists 
topical reports that are used to determine core operating limits. 
Topical reports 15 through 19 are LOCA methodology reports and are 
being deleted. These reports describe Siemens LOCA methodology. As 
stated in Reference 1, the GE SAFER/GESTR LOCA methodology is being 
used for this uprated cycle. In addition, other minor methodology 
changes were made for power uprate transient analysis. GE topical 
report NEDC-32071P, PP&L topical report NE-092-001 and the NRC 
Safety Evaluation Report on the PP&L power uprate licensing topical 
are proposed to be added as Topical Reports No. 15, 16, and 17, 
respectively.
    1. No. These changes are editorial in nature in that only the 
references to documents are being changed. The methodology used to 
determine core limits have been previously reviewed and approved by 
the NRC.
    2. No. See the response to Question 1 above.
    3. No. See the response to Question 1 above.
    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: Osterhout Free Library, Reference 
Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: Mohan C. Thadani, Acting

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: July 27, 1994
    Description of amendment request: This amendment will change the 
definition of a CORE ALTERATION included in Technical Specification 
Section 1.0 for each unit to allow movement and replacement of local 
power range monitors and control rods in a defueled cell. The new 
definition is consistent with the Improved Standard Technical 
Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. In the submittal, the licensee stated that:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change eliminates two previous evolutions, LPRM and 
Control Rod movement from a defueled cell, from being considered 
CORE ALTERATIONS. Thus the issue is whether the elimination of these 
constraints could contribute to a significant increase in the 
probability or consequences of a reactivity event.
    Adding local power range monitors to the list of detectors which 
can be moved without invoking CORE ALTERATION requirements allows 
for the removal of these detectors for repair and replacement. 
Movement of these components does not impact the reactivity of the 
core. Therefore, allowing the movement of these detectors without 
invoking CORE ALTERATION provisions, does not contribute to a 
significant increase in the probability or consequences of a 
reactivity event.
    Removal of a Control Rod from a defueled cell results in a 
negligible increase in core reactivity. Appropriate Technical 
Specification controls and refueling interlocks are applied during 
the fuel movements preceding the control rod removal to protect from 
or mitigate a reactivity excursion event. In addition, the design of 
a control rod precludes its replacement without all fuel assemblies 
in the cell removed. Therefore, allowing the movement of control 
rods from a defueled cell without invoking CORE ALTERATION 
provisions, does not contribute to a significant increase in the 
probability or consequences of a reactivity event.
    The proposed Technical Specification change to adopt the revised 
CORE ALTERATION definition (NUREG 1433, as amended) does not effect 
the probability or consequences of an accident previously evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change eliminates two previous evolutions, LPRM and 
Control Rod movement from a defueled cell, from being considered 
CORE ALTERATIONS. Thus the issue is whether the elimination of these 
constraints could create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    For local power range monitors, Technical Specification 3/4.3.1 
defines the minimum number of LPRMs required to be maintained 
operable in OPCON 5 and during Shutdown Margin Demonstration. The 
addition of LPRMs as an exclusion under the CORE ALTERATION 
definition does not change the operability requirements for the 
LPRMs under Technical Specification 3/4.3.1. Thus the ability of the 
LPRMs to perform their monitoring function is not affected by the 
proposed CORE ALTERATION definition change. In addition, movement of 
these components does not impact the reactivity of the core. 
Therefore, allowing the movement of these detectors without invoking 
CORE ALTERATION provisions, does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    For Control Rods, in the unlikely event that the wrong control 
rod was inadvertently withdrawn from a fueled cell during evolutions 
which were not intended to be CORE ALTERATIONS, adequate protective 
measures are provided by design and core monitoring instrumentation 
required to be operable in OPCON 5. Withdrawal of a single control 
rod from a cell containing fuel is bounded by Shutdown Margin 
analysis and demonstration. However, assuming the inadvertent 
control rod withdrawal resulted in a significant reactivity 
addition, the Reactor Protection System (RPS) would respond by 
inserting all control rods via the Scram function. The RPS monitors 
for recriticality during OPCON 5 with SRMs (except during specific 
controlled evolutions), IRMs, and APRMs. The Scram circuitry is 
completely redundant from the insert and withdrawal circuitry for 
the control rods. Therefore, allowing the movement of control rods 
from a defueled cell without invoking CORE ALTERATION provisions, 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed Technical Specification change to adopt the revised 
CORE ALTERATION definition (NUREG 1433, as amended) does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    To evaluate the potential effect on safety margin, the proposed 
change was evaluated as to its effect on Shutdown Margin. Shutdown 
Margin defines the amount of reactivity by which the reactor is 
subcritical, and thus is a measure of the safety margin in avoiding 
unanticipated criticality events.
    The movement of LPRMs does not impact the reactivity of the 
core, and thus does not reduce the Shutdown Margin. Removal of a 
Control Rod from a defueled cell results in a negligible increase in 
core reactivity. Therefore, the removal of a Control Rod from a 
defueled cell will have a negligible effect on the core Shutdown 
Margin. Per Technical Specification 3/4.9.10.2(c), adequate core 
Shutdown Margin must exist during refueling when multiple control 
rods and the surrounding fuel assemblies are removed from the core. 
Appropriate Technical Specification controls and refueling 
interlocks are applied during the fuel movements preceding the 
control rod removal to protect from or mitigate a reactivity 
excursion event. In addition, the core is analyzed to maintain 
Shutdown Margin even with the withdrawal of the highest worth rod 
from a fueled cell.
    The proposed Technical Specification change to adopt the revised 
CORE ALTERATION definition (NUREG 1433, as amended) 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: Osterhout Free Library, Reference 
Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: Mohan Thadani, Acting

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

    Date of amendment request: July 20, 1994
    Description of amendment request: The amendments would raise the 
Steam Leakage Detection system set-points that isolate the High 
Pressure Coolant Injection System (HPCI) and Reactor Core Isolation 
Cooling (RCIC) system equipment on high equipment room temperature and 
high delta temperature. The amendments are supported by a Limerick 
Generating Station modification to increase the environmental 
qualifications limits of the HPCI and RCIC systems to allow the systems 
to remain operable when equipment room cooling is unavailable.
    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 Technical Specifications changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    Those accident which are potentially impacted by these changes 
are any accident or events that require the isolation of the HPCI or 
RCIC system steam supply lines. This would include gross failures 
(pipe breaks) or significant leaks (pipe cracks) in steam lines. 
Minor leaks that do not significantly affect the environment in the 
equipment compartments are only considered with regard to being 
potential precursors to the development of a larger crack or break. 
The ability to detect small steam leaks is not dependent on the 
isolation instrumentation and the proposed changes to the isolation 
instrumentation will not impact the detection methods.
    The proposed TS changes will not increase the probability of an 
accident since the changes will only increase the trip set-points of 
the instrumentation which detect increases in the temperature in the 
HPCI and RCIC equipment rooms. The physical establishment and 
setting of the proposed set-points of these accident detection and 
mitigation instruments will have no direct physical impact on the 
plant's normal operating conditions. This instrumentation is 
normally in a ``monitoring mode,'' and is not actively supporting 
normal plant operation. Therefore, the proposed set-points can have 
no impact on the operating plant that would make an accident more 
likely to occur.
    Two perspectives were evaluated regarding the potential impact 
on the consequences of accidents. One case is the impact on 
accidents which do not require HPCI or RCIC steam line isolation, 
but that may require the operation of the HPCI or RCIC Systems. The 
other case is the impact resulting from HPCI and RCIC steam line 
break accidents.
    In the first case, the proposed changes to the set-points of 
these accident mitigation instruments will have no direct physical 
impact on the plant's accident response, except during the HPCI or 
RCIC pipe break accidents. During all other pipe breaks or 
accidents, the bounding peak HPCI and RCIC equipment compartment 
temperatures will still be at least 35 deg.F below the proposed TS 
lower allowable values (i.e., 218 deg.F and 198 deg.F, 
respectively), and the isolation instrumentation will remain in a 
``monitoring mode.'' The isolation instrumentation will only be 
required to continue to passively monitor the HPCI and RCIC 
compartment temperatures and will meet the design basis by not 
inadvertently isolating the HPCI or RCIC systems.
    In the second case, the HPCI and RCIC pipe break accidents 
described in LGS, Updated Final Safety Analysis Report (UFSAR) 
Section 3.6 ``Protection Against Dynamic Effects Associated with the 
Postulated Rupture of Piping,'' determine the peak pressures and 
temperatures for the affected compartments. These peak pressures for 
the HPCI and RCIC breaks are the bounding pressures for breaks in 
these lines and, since they occur quickly, they are unaffected by 
the leak detection and isolation actuation systems. The peak 
pressures predicted in the UFSAR for the largest HPCI and RCIC steam 
line breaks, in the HPCI, RCIC and isolation valve compartments, are 
the bounding values for breaks of all sizes in these compartments. 
In addition, the peak temperatures are not affected by the proposed 
changes to the isolation actuation set-points. Therefore, the 
isolation of the HPCI and RCIC steam lines following a HPCI or RCIC 
steam line guillotine break is not dependent on the temperature trip 
functions, rather, the isolation is dependent on the high flow or 
low pressure trip functions where a delay in the response of the 
temperature isolation instrumentation will have no adverse impact on 
the consequences of the accidents described in the SAR.
    An evaluation was performed to determine the potential impacts 
due to the proposed changes affecting the room temperatures used in 
the environmental qualification program. The results of this 
evaluation determined that the postulated peak temperatures for the 
HPCI pump room and the HPCI and RCIC piping areas would be at the 
saturation temperature for the HPCI or RCIC break blow-down in these 
compartments, therefore, these compartment temperatures values will 
not be exceeded. The RCIC pump room and isolation valve compartment 
environmental qualification temperatures were not postulated to be 
at the saturation temperature. However, this does not increase the 
consequences of any of the accident described in the SAR because the 
equipment which is normally required for RCIC system operation and 
which is located in the RCIC pump compartment is not required to 
operate following breakage of the RCIC steam supply line. The only 
equipment in the RCIC pump compartment that is required to operated 
following a RCIC steam line break is the RCIC leak detection 
instrumentation which are qualified to operate at temperatures 
greater than the saturation temperature. Finally, the isolation 
valve compartment postulated peak temperatures result from a HPCI 
steam line break in the Unit 1 and 2 isolation valve compartments. 
This line break produces the highest isolation valve compartment 
temperatures which bounds the results of a RCIC steam line break in 
the isolation valve compartment and the HPCI and RCIC steam lien 
breaks in the HPCI and RCIC pump rooms and piping areas. However, 
since the leak detection and isolation actuation trip set-points for 
the instruments in the isolation valve compartment are not being 
changed, then the environmental conditions in the isolation valve 
compartment will remain unchanged. This will assure that the 
isolation valves will be able to provide isolation when required.
    For HPCI or RCIC leaks, the environmental conditions were not 
the only design basis considerations evaluated. The radiological 
affects were also considered. By increasing the upper allowable high 
ambient temperature or high delta temperature values for certain 
line break sizes there will be a larger total mass blow-down from 
the break due to the corresponding lengthening of the time to reach 
the higher temperature limit. However, the total integrated mass of 
blowdown prior to isolation of the HPCI or RCIC steam line break 
will still be bounded by the LGS UFSAR accident analysis and 
therefore, the radiological consequences of these breaks as 
described in the SAR will remain unchanged. These conclusions are 
supported by an evaluation that provided the design basis for the 
main steam line break and then examines the radiological 
consequences at the upper and lower end of the HPCI and RCIC break 
spectrum. Since the largest HPCI and RCIC breaks are isolated based 
on high flow and not based on compartment temperature increases, 
then the proposed changes in the temperature set-points have no 
impact on the radiological consequences of the design basis HPCI or 
RCIC pipe break accidents as described in the SAR.
    The impact of the proposed changes on the probability of a 
malfunction of the system isolation instrumentation, valves, or the 
HPCI or RCIC systems was evaluated. The isolation actuation 
instruments are qualified for the expected environmental conditions 
and the proposed set-points are within the normal operating range of 
the instruments. Therefore, these isolation actuation instruments 
are more likely to randomly fail than before. In addition, by 
ensuring that there is no adverse impact on the ability of the HPCI 
or RCIC systems to respond to events which are caused by 
malfunctions of equipment, then the consequences of these events are 
not increased. An adequate margin between the proposed lower 
allowable trip values and the postulated equipment room 
environmental conditions is being maintained such that an 
inadvertent actuation of the HPCI or RCIC system isolation function 
is also no more likely to occur. The increase in the temperature 
isolation allowable trip values will allow increased blow-down from 
a pipe break or crack which will result in higher pump compartment 
temperatures and pressures than before for a given break size; 
however, the overall impact is still bounded by the LGS UFSAR 
Section 3.6 ruptured piping analyses. The isolation actuation 
instruments are qualified for the expected environmental conditions, 
and the proposed set-points are also within the normal operating 
range of the isolation instruments. Therefore, the instruments are 
no more likely to randomly fail and cause the loss of the HPCI or 
RCIC system than before. In fact, by increasing the qualification 
limits of the HPCI and RCIC systems, the systems will be able to 
remain operable with an even large steam leak in the room when room 
cooling is available. Therefore, the changes will have no impact on 
the operating plant that would increase the possibility or 
consequences of a malfunction of equipment important to safety.
    Since the proposed changes will maintain the HPCI or RCIC steam 
isolation system design basis, where the consequences are bounded by 
an analysis contained in the LGS UFSAR, and will only change the 
set-points of the existing instrumentation without impacting 
equipment important to safety, the proposed Technical Specifications 
changes do not involve a significant 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 proposed TS changes will not create the possibility of a 
different type of accident or malfunction of equipment since the 
changes will only increase the trip set-points of the 
instrumentation which detect increases in the temperature in the 
HPCI and RCIC equipment rooms. The physical establishment and 
resetting of the set-points of these accident detection and 
mitigation instruments will have not direct physical impact on the 
plant's normal operating conditions and will not create any new 
accident initiators or failure modes. The severity of the potential 
piping system pressure transients caused by the isolation of the 
HPCI or RCIC steam lines at higher room temperatures remains 
unchanged since the isolation occurs after the postulated break 
blow-down has dropped to its steady state rate. Therefore, the 
changes will not result in a pipe break or result in any malfunction 
of equipment that has not previously been postulated to occur.
    Therefore, the proposed set-points will not create the 
possibility of a different type of accident or possibility of a 
different type of malfunction of equipment important to safety than 
previously evaluated in the SAR.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety for the isolation actuation instrumentation 
as defined in the TS bases is not reduced. The proposed system 
isolation TS trip set-points were selected to provide equivalent 
margins that ensure the effectiveness of the isolation systems to 
mitigate the consequences of accidents without compromising the 
operability of the HPCI and RCIC systems. The proposed trip set-
points and proposed allowable value ranges maintain adequate margins 
between these new values and the operating range of the HPCI and 
RCIC systems in order to prevent the inadvertent actuation of the 
isolation system and the loss of either the HPCI or RCIC systems. 
The differences between the trip set-points and the allowable values 
are being maintained as an allowance for instrument drift. The trip 
set-points and the allowable ranges are within the specified range 
of the instruments and therefore, the accuracy and drift will 
provide the same margin of safety as previously assumed.
    Therefore, the proposed TS change do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room: Pottstown Public Library, 500 High 
Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Mohan C. Thadani, Acting

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

    Date of amendment request: July 22, 1994
    Description of amendment request: This amendment would remove the 
surveillance frequency details which govern 10 CFR 50, Appendix J, Type 
B and C testing from 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 Technical Specifications changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes involve the removal of repetitious 
surveillance details from TS also found in 10 CFR 50, Appendix J, 
and rewording of TS. The removal and rewording involves no technical 
changes to the existing TS. The changes to the existing TS are 
proposed in order to be consistent with NUREG-1433. During the 
development of NUREG-1433, certain wording preferences or English 
language conventions were adopted. The proposed changes to this TS 
section are administrative in nature and do not impact initiators of 
analyzed events. They also do not impact the assumed mitigation of 
accidents or transient events. Therefore, the changes do not involve 
a significant 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 proposed changes do not involve a physical alteration of the 
plant or changes in methods governing normal plant operation. The 
proposed changes will not impose any new or different requirements 
or eliminate any existing requirements. Therefore, the 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 changes are administrative in nature and will not involve 
any technical changes. The proposed changes will not reduce a margin 
of safety because they have no impact on any safety analysis 
assumptions. In addition, because the changes are administrative in 
nature, no question of safety is involved. Therefore, the changes do 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room: Pottstown Public Library, 500 High 
Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Mohan C. Thadani, Acting

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

    Date of amendment request: August 19, 1994
    Description of amendment request: This change would reduce the 
minimum setpoints and allowable values for the Steam Generator Level - 
Low-Low and Low reactor protection system signals. The bases would also 
be modified to expand the description of the relationship between 
setpoints, allowable values and the plant safety analysis.
    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. do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The Steam Generator Water Level--Low-Low signal and the Low 
Steam Generator Level coincident with Steam Flow/Feed Flow Mismatch 
signal are designed to mitigate design basis transients involving 
significant reductions of steam generator inventory (e.g., Loss of 
Normal Feedwater, Turbine Trip, Loss of Offsite Power, Feedwater 
Line Break). The setpoints and allowable values for these protection 
signals are prescribed by Technical Specifications such that 
performance of the signals is consistent with the plant safety 
analyses, considering the effects of channel uncertainties. The 
proposed reductions to the setpoints and allowable values for the 
low-low and low steam generator level signals would not affect the 
probability of any transient that the protection signals are 
designed to mitigate. The changes would reduce the probability of 
unnecessary reactor trips and Auxiliary Feedwater (AFW) system 
actuations by providing greater operating margin for plant 
evolutions involving steam generator level changes (e.g., plant 
startup). Therefore, the proposed changes do not involve any 
increase in probability of an accident previously evaluated.
    The changes to the Steam Generator Water Level--Low-Low signal 
would not result in any increase in consequences of a previously 
analyzed accident because the proposed setpoint and allowable value 
would continue to ensure the safety analysis assumptions remain 
valid. As described in the accompanying changes to the Technical 
Specifications Bases, the channel uncertainty calculations performed 
to establish the relationships between the setpoints, allowable 
values and safety analyses are consistent with NRC Regulatory Guide 
1.105, Revision 2. Low Steam Generator Level coincident with Steam 
Flow/Feed Flow Mismatch signal is not credited in the UFSAR Chapter 
15 safety analyses. The proposed changes to the low steam generator 
level setpoint and allowable value would continue to provide 
reliable backup to the low-low level trip signal, consistent with 
IEEE-279-1971. Therefore, the proposed changes would not involve an 
increase in consequences of any previously analyzed accident.
    2) do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes would continue to ensure the appropriate 
reactor protection system functions (reactor trip and AFW 
initiation) are initiated in the event that steam generator water 
level decreases to the value used in the plant safety analyses. The 
proposed changes would not involve any changes in protection system 
logic or function, and do not involve any plant configurations that 
could adversely affect the initiation or progression of any accident 
sequence. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3) do not involve a significant reduction in a margin of safety.
    The proposed setpoints and allowable values would continue to 
ensure that the assumptions in the safety analyses remain valid, 
with appropriate consideration of protection system channel 
uncertainties. Therefore, the proposed 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: Salem Free Public library, 112 West 
Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: Mohan C. Thadani, Acting

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

    Date of amendment request: July 20, 1994
    Description of amendment request: The proposed change would modify 
the Virgil C. Summer Nuclear Station (VCSNS) Technical Specification 
(TS) Tables 2.2-1, ``Reactor Trip System Instrumentation Setpoints,'' 
and 3.3-4, ``Engineered Safety Features Actuation System 
Instrumentation Trip Setpoints,'' and several associated bases. The 
proposed change would remove three columns from the Tables. The columns 
contain specific rack and sensor allowable drift values.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of VCSNS in accordance with the proposed license 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    This change does not alter or delete any setpoints or Allowable 
Values, and as such, has no affect on any assumptions used for 
accident analysis. No hardware or software changes are involved, so 
no common mode or common cause failures can occur as a result of 
this change. This change has no impact on the daily operation of 
VCSNS. The performance of periodic calibrations and channel checks 
will assure the setpoints remain within tolerance. Since this 
amendment request affects only information that is no longer used in 
the daily operation of the plant and has no impact on accident 
analysis, the probability or consequences of an accident previously 
evaluated are not increased.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This change revises two TS tables which contain both setpoints 
and Allowable Values as well as other information for safety trip 
functions. However, the revision only deletes three columns of data 
that were used in determining the operability of one channel of the 
safety function. These values are also used in determining the 
setpoints and are based on measured or published tolerances and 
uncertainties. Although these columns are being deleted, no changes 
to any hardware, software, or setpoints will occur. Since these 
changes do not have any plant impact, no new failure mechanisms are 
introduced. Only the information not used on a daily basis is being 
removed from these tables; this will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    This change revises the format of TS Tables 2.2-1 and 3.3-4 
which list the setpoint and Allowable Values for safety trip 
functions. The data that is being removed from these tables was used 
to establish clear reportability requirements for any portion of one 
channel of any of the listed safety trip functions. Since the 
reporting requirements have changed and an LER is not required if 
one coincident channel is inoperable, this data is no longer used in 
daily operations. The margin of safety was established when 
setpoints and Allowable Values were determined, and no changes to 
these values are involved. There is no reduction in a margin of 
safety that could affect the plant, SCE&G employees, or the public.
    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: Fairfield County Library, Garden and 
Washington Streets, Winnsboro, South Carolina 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 20, 1994
    Description of amendment request: The proposed change would modify 
the Virgil C. Summer Nuclear Station, Unit 1, (VCSNS) Technical 
Specifications (TS) to allow alternative, equivalent testing of diesel 
fuel used in the emergency diesel generators (EDG). These alternative 
methods are necessary due to recent changes in Environmental Protection 
Agency (EPA) regulations that are designed to limit the use of high 
sulfur fuels.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The probability or consequences of an accident previously 
evaluated is not significantly increased.
    The change in testing methods for the EDG fuel oil has no impact 
on the probability or consequences of any design basis accident. 
These tests have been determined to be equivalent to the previously 
approved testing methods and are needed due to changes in the EPA's 
regulations regarding sulfur in motor vehicle fuels. The dye used to 
identify high sulfur fuels will have no adverse affect on the 
performance of the EDG's. The proposed testing assures a continued 
high level of quality of the diesel fuel received and stored on 
site.
    The change in revision level of a reference in TS section 
6.9.1.11 has no impact on the probability of occurrence or 
consequences of any design basis accident. All design and 
performance criteria will continue to be met and no new single 
failure mechanisms will be created. The change in revision level for 
WCAP-10216-P-A does not involve any alterations to plant equipment 
or procedures which could affect any operational modes or accident 
precursors. This change only incorporates by reference, the 
methodology for determining the penalty to be used in calculating 
Core Operating Limits. This methodology allows the penalty to be 
cycle specific and is primarily affected by the core configuration. 
This penalty is used for normal operation and provides more 
conservatism to the core operation for the cycle.
    2. [The proposed license amendment does not] create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The change in testing methods for the EDG fuel oil will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. These tests have been determined 
by the EPA and other organizations to be equivalent to the 
previously approved testing methods. The effect of the blue dye, 
used to identify high sulfur fuels, on the performance of the EDGs 
has been evaluated and determined to be insignificant. The testing 
proposed assures a continued high level of quality for the diesel 
fuel received and stored on site.
    The change of revision level of a reference in TS section 
6.9.1.11 has no impact on the probability of occurrence or 
consequences of any design basis accident. All design and 
performance criteria will continue to be met and no new single 
failure mechanisms will be created. The change in revision level for 
WCAP-10216-P-A does not involve any alterations to plant equipment 
or procedures which could affect any operational modes or accident 
precursors. This change only incorporates, by reference, the 
methodology for determining the penalty to be used in calculating 
Core Operating Limits. This methodology allows the penalty to be 
cycle specific and is primarily affected by the core configuration. 
This penalty is used for normal operation and provides more 
conservatism to the core operation for the cycle.
    3. [The proposed license amendment does not] involve a 
significant reduction in a margin of safety.
    The change in testing methods for the EDG fuel oil will not 
involve a significant reduction in a margin of safety. The proposed 
testing methods have been determined to be equivalent to the 
previously approved testing methods. The test for sulfur assures 
that the sulfur content is within the allowable range for weight-
percent. The test for color and clarity assures that the fuel is 
relatively free of water and particulate contaminants. The proposed 
tests provide at least an equivalent level of quality and 
repeatability for the fuel oil analysis, thus assuring that the 
margin of safety is not reduced.
    The change in revision level of a reference in TS section 
6.9.1.11 does not change the proposed reload design or safety 
analysis limits for each cycle reload core. The associated change to 
WCAP-10216-P-A due to the revision will be specifically evaluated 
using approved reload design methods. The larger penalty actually 
provides for an increase in margin during certain burnup ranges. 
Since the safety analysis limits are unaffected, and the cycle 
specific analysis will show that the analysis limits are met, the 
change proposed will have no adverse impact on a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room: Fairfield County Library, Garden and 
Washington Streets, Winnsboro, South Carolina 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: David B. Matthews

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

    Date of amendment request: August 19, 1994 (TS 93-09)
    Description of amendment request: The proposed change would revise 
the implementation schedule for Amendment Nos. 182 and 174 from that 
stated in the amendments when they were approved by the Commission by 
letter dated May 24, 1994. As issued, the amendments reflected the 
licensee's plans to implement the changes during the Unit 2 Cycle 6 
refueling outage. However, the licensee has determined that 
implementation would be more appropriate following the refueling outage 
when both units are operating in 1995. No changes to the technical 
specification pages other than those approved when the amendments were 
issued are needed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
determined that the no significant hazards consideration exists. This 
analysis was provided in the original submittal for the amendment from 
the licensee dated October 1, 1993, and was used in the preparation of 
the amendments. The licensee has determined that this analysis remains 
valid for the proposed revision to the implementation dates and that 
the changes do not constitute a significant hazard. The staff 
previously issued the proposed finding in the Federal Register (59 FR 
4947) and there were no public comments on the finding. This analysis 
is reproduced as follows:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revision supports the implementation of design 
logic and setpoint changes to the loss-of-power relaying. This 
relaying is designed to ensure adequate voltage is available to 
safety-related loads in order to enhance their operability and 
support accident mitigation functions and to provide for auxiliary 
feedwater (AFW) pump starts. The design changes alter relay logic 
and delete unnecessary relaying, but do not change the diesel 
generator (D/G) start and load-shedding actuations that result from 
loss-of-power conditions. Therefore, no new actuations or functions 
have been created; and because the existing and proposed functions 
provide for accident mitigation considerations that are not the 
source of an accident, the probability of an accident is not 
increased. The deletion of the 6.9-kilovolt shutdown board normal-
feeder undervoltage relays actually reduces the potential for 
inadvertent shutdown board blackouts as a result of short-duration 
voltage transients or instrument failures.
    The setpoints and time delays for loss-of-power functions have 
been modified based on the guidelines developed by the Electrical 
Distribution System Clearinghouse as evaluated and determined 
through detailed analysis by TVA. This design is documented in TVA 
Calculations SQN-EEB-MS-TI06-0008, 27DAT, and DS-1-2 and is 
available for NRC review at the SQN site. The assigned values are 
conservative settings that will ensure adequate voltage is supplied 
to safety-related loads for accident mitigation and safety functions 
under normal, degraded, and loss-of-offsite-power voltage conditions 
with appropriate time delays to prevent damage to electrical loads 
and minimize premature or unnecessary actuations. The identification 
of loss-of-voltage conditions is enhanced by the design changes to 
ensure the timely sequencing of loads onto the D/G and the 
initiation of AFW pump starts for accident mitigation. Because there 
are no reductions in safety functions resulting from the design 
logic, setpoint, and time-delay changes to the loss-of-power 
instrumentation and offsite dose levels for postulated accidents 
will not be increased, the consequences of an accident are not 
increased.
    The applicable mode addition, TS 3.0.4 exclusion deletion, and 
response time measurement clarification incorporated in the proposed 
change do not affect plant functions. These changes reflect the 
requirements that SQN has been maintaining and serve to clarify the 
requirements to provide consistency of application and easier 
understanding. The AFW footnote addition and bases revision only 
clarify operability conditions that are consistent with the plant 
design for the AFW pump and loss-of-power instrumentation. Because 
there are no changes to plant functions or operations, these 
revisions have no impact on accident probabilities or consequences.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    As described above, the loss-of-power instrumentation ensures 
adequate voltage to safety-related loads by initiating D/G starts 
and load shedding and provides for AFW pump starting, but is not 
considered to be the source of an accident. Although the design 
logic, setpoint, and time-delay actuation criteria have changed, the 
output functions to various plant systems that actuate for load 
shedding and D/G starts remain the same. Therefore, actuation 
criteria have been affected, but not safety functions, and the TVA 
evaluation has confirmed that the new design enhances the ability to 
maintain adequate voltage to support safety functions. Since safety 
functions have not changed and the new loss-of-power instrumentation 
design continues to support operability of safety-related equipment, 
no new or different accident is created.
    The applicable mode addition, TS 3.0.4 exclusion deletion, and 
response time measurement clarification, as well as the AFW 
operability clarifications, do not affect plant functions and will 
not create a new accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed loss-of-power TS changes support design logic, 
setpoint, and time-delay requirements that have been verified by TVA 
analysis to provide acceptable voltage levels for safety-related 
components. In determining the acceptability of these voltage 
levels, the minimum voltage for operation as well as detrimental 
component heating resulting from sustained degraded-voltage 
conditions were considered. This design ensures that safety-related 
loads will be available and operable for normal and accident plant 
conditions. The applicable mode addition, TS 3.0.4 exclusion 
deletion, response time measurement clarification, and AFW 
operability clarifications provide enhancements to TS requirements 
and do not affect plant functions. Therefore, no safety functions 
are reduced by these changes and there is no reduction in the margin 
of safety.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room: Chattanooga-Hamilton County Library, 
1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

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

Philadelphia Electric Company, Public Service Electric and Gas 
Company, Delmarva Power and Light Company, and Atlantic City 
Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania

    Date of amendment request: June 23, 1993
    Brief description of amendment request: The amendments would revise 
the licenses and the technical specifications to change the maximum 
core power limit from 3293 MWt to 3458 MWt.Date of publication of 
individual notice in Federal Register: August 29, 1994 (59 FR 44432) 
Expiration date of individual notice: September 28, 1994
    Local Public Document Room: Government Publications Section, State 
Library of Pennsylvania, (REGIONAL DEPOSITORY) Education Building, 
Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of amendments request: August 9, 1994
    Brief description of amendments request: These amendments revise 
the Technical Specifications (TS) 5.3.4, ``Steam and Power Conversion 
Systems,'' and 15.3.7, ``Auxiliary Electrical Systems,'' to increase 
the allowed outage times for one motor driven auxiliary feedwater pump 
and for the standby emergency power for the Unit 1, Train B4160 Volt 
safeguards bus (A06) from 7 to 12 days. The proposed amendments would 
also modify TS 15.3.3, ``Emergency Core Cooling System, Auxiliary 
Cooling Systems, Air Recirculation Fan Coolers, and Contained Spray,'' 
to provide the clarification that the service water pump (P-32E) 
operating with power supplied by the Alternative Shutdown System is 
operable from offsite power. The changes are one-time extensions of 
specific allowed outage times.Date of publication of individual notice 
in the Federal Register: August 19, 1994 (59 FR 42870).
    Local Public Document Room: Joseph P. Mann Library, 1516 Sixteenth 
Street, Two Rivers, Wisconsin 54241.

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 20555, and at the local public document 
rooms for the particular facilities involved.

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

    Date of application for amendments: November 3, 1993
    Brief description of amendments: The amendments revise the Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Technical Specifications 
(TSs) by removing the TSs that are applicable to the incore instrument 
(ICI) system. The limitations on the use of the ICI system will be 
relocated to the Updated Final Safety Analysis Report. The core power 
distribution limits, which the ICI system is used to verify, remain in 
the TSs which is consistent with 10 CFR 50.36.Date of issuance: August 
24, 1994Effective date: As of the date of issuance to be implemented 
within 30 days.
    Amendment Nos.:  191 and 168
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64601) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated August 24, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room:  Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: November 3, 1993
    Brief description of amendments: The amendments modify the 
surveillance requirements to reflect the removal of the auto-closure 
interlock from the shutdown cooling system and revises the setpoint for 
the open permissive interlock.
    Date of issuance: August 24, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 192 and 169
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64600) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated August 24, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: May 27, 1994
    Brief description of amendments: The amendments revise the 
Technical Specification surveillance test intervals from monthly to 
quarterly for several channel functional tests for the Reactor 
Protection System and the Engineered Safety Feature Actuation System. 
In addition, an administrative change was made to remove an out-of-date 
footnote concerning the Emergency Diesel Generator logic circuit 
modifications.
    Date of issuance: August 24, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 193 and 170
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37062) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated August 24, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room:  Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: November 5, 1993, as 
supplemented March 11, 1994
    Brief description of amendments: The amendments consist of two 
related changes. The first change revises the containment penetration 
Technical Specifications (TSs) to resemble the containment penetration 
TSs in NUREG-1432, ``Standard Technical Specifications for Combustion 
Engineering Pressurized Water Reactors.'' The second revises the TSs to 
allow the containment personnel airlock to be open during fuel movement 
and core alterations. The TS Bases have also been revised to reflect 
the changes as the result of issuing these amendments.
    Date of issuance: August 31, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 194 and 171
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64602) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated August 31, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room:  Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendment: June 16, 1994
    Brief description of amendment: The amendment removes from 
Technical Specification 3/4.8.3, ``Onsite Power Distribution,'' a 
footnote applicable for Cycle 18 only, and adds surveillance 
requirement 4.8.3.1.2, to test the MCC-5 automatic bus transfer feature 
once per refueling.
    Date of Issuance: August 23, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 176
    Facility Operating License No. DPR-61. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37067) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated August 23, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room: Russell Library, 123 Broad Street, 
Middletown, Connecticut 06457.

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

    Date of amendment request: April 28, 1994
    Brief description of amendments: The amendments revised Technical 
Specification 4.6.1.3.e to add an option that will allow the personnel 
airlock pneumatic system leak test to be completed in 8 hours with a 
pressure drop of 0.50 psi.
    Date of issuance: August 29, 1994
    Effective date: August 29, 1994
    Amendment Nos.:  Unit 1 - Amendment No. 64; Unit 2 - Amendment No. 
53
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27057) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 29, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room: Wharton County Junior College, J. M. 
Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: November 15, 1993
    Brief description of amendments: The amendment revises the 
Technical Specifications to extend the surveillance interval for the 
chemical analysis, inventory, and flow area of the ice condenser from 9 
to 18 months.
    Date of issuance: August 23, 1994
    Effective date: August 23, 1994
    Amendment Nos.: 180 & 164
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67849) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 23, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room: Maud Preston Palenske Memorial Library, 
500 Market Street, St. Joseph, Michigan 49085.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: September 24, 1992 and 
supplemented March 2, 1994.
    Brief description of amendments: The amendment removes the list of 
containment isolation valves and associated references to the list from 
the Technical Specifications.
    Date of issuance: August 29, 1994
    Effective date: August 29, 1994
    Amendment Nos.:  181 and 165
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 17, 1993 (58 
FR 8773) The March 2, 1994, letter provided supplemental information 
that was not outside the scope of this initial notice. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated August 29, 1994.No significant hazards consideration 
comments received: No.
    Local Public Document Room: Maud Preston Palenske Memorial Library, 
500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: July 1, 1994
    Brief description of amendment: The amendment revises the secondary 
containment drawdown time testing requirement of Technical 
Specification (TS) 4.6.5.1.c.1 and the secondary containment inleakage 
testing requirement of TS 4.6.5.1.c.2. The amendment supports a revised 
design basis radiological analysis which supports an increase in 
secondary containment drawdown time from 6 to 60 minutes by taking 
credit for fission product scrubbing and retention in the suppression 
pool which were not assumed in the original radiological analysis but 
are currently assumed in the NRC's Standard Review Plan (NUREG-0800). 
The revised analysis also takes credit for additional mixing of primary 
containment and engineered safety feature system leakage with 50 
percent of the secondary containment free air volume prior to the 
release of radioactivity to the environment. The revised radiological 
evaluation has determined that the radiological doses remain below 10 
CFR Part 100 guideline values and General Design Criterion 19 criteria.
    Date of issuance: August 30, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 56
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37074) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 30, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room: Reference and Documents Department, 
Penfield Library, State University of New York, Oswego, New York 13126.

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

    Dates of application for amendment: November 30, 1993 and June 30, 
1994.
    Brief description of amendment: The proposed amendment would delete 
the requirements for a chlorine detection system from the following 
sections of Technical Specifications: 3.2.I, 3.17.A, 4.17.A, tables 
4.2.1 and Technical Bases 3.2 and 3.17.A. Due to design changes at the 
Monticello Nuclear Generating Plant, chlorine is no longer stored 
onsite as a liquified gas and regulations requiring early warning of an 
onsite chlorine release do not apply.
    Date of issuance: August 25, 1994
    Effective date: August 25, 1994
    Amendment No.: 89
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10010) The June 30, 1994, letter provided documents cited in the 
amendment application and did not affect the staff's initial no 
significant hazards determination. The Commission's related evaluation 
of the amendment is contained in a Safety Evaluation dated August 25, 
1994.No significant hazards consideration comments received: No.
    Local Public Document Room: Minneapolis Public Library, Technology 
and Science Department, 300 Nicollet Mall, Minneapolis, Minnesota 
55401.

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

    Date of amendment request: May 21, 1993, as supplemented by letters 
dated September 10, 1993, and May 25, 1994
    Brief description of amendment: The amendment changed the Technical 
specifications to reflect the relocation of the old 10 CFR 20.106 
requirements to the new 10 CFR 20.1302, and to implement administrative 
changes.
    Date of issuance: August 24, 1994
    Effective date: August 24, 1994
    Amendment No.: 164
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36442) The additional information contained in the supplemental letters 
dated September 10, 1993, and May 25, 1994, was clarifying in nature 
and thus, within the scope of the initial notice and did not affect the 
staff's proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 24, 1994.No significant hazards 
consideration comments received: No.
    Local Public Document Room: W. Dale Clark Library, 215 South 15th 
Street, Omaha, Nebraska 68102

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

    Date of amendment request: February 12, 1993, as supplemented by 
letters dated August 20, 1993, and June 6, 1994
    Brief description of amendment: This amendment revised Technical 
Secification 2.1.4, ``Reactor Coolant System Leakage Limits,'' to 
implement the reactor coolant system leak-before-break methodology 
detection criteria. Additionally, administrative changes were made.
    Date of issuance: August 25, 1994
    Effective date: August 25, 1994
    Amendment No.: 165
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37076) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 25, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room: W. Dale Clark Library, 215 South 15th 
Street, Omaha, Nebraska 68102.

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

    Date of application for amendments: December 8, 1993 (Ref. LAR 93-
07)
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit 
Nos. 1 and 2 to revise TS 3/4.8.1, ``A.C. Sources'' to increase the 
required quantity of emergency diesel generator (EDG) fuel oil stored 
in the engine-mounted tank (day tank) from 200 gallons to 250 gallons. 
The amendment also revises TS 3/4.7.11, ``Area Temperature 
Monitoring,'' and 3/4.8.1 to remove references to a five EDG 
configuration, based on the installation of a sixth EDG.
    Date of issuance: August 23, 1994
    Effective date: August 23, 1994
    Amendment Nos.: 93 and 92
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7694) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 23, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

Philadelphia Electric Company, Docket No. 50-352, Limerick 
Generating Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: June 6, 1994
    Brief description of amendment: This amendment removes the controls 
for a remote shtudown system control valve and deletes the isolation 
signal for certain primary containment isolation valves from TS Tables 
3.3.7.4-1 and 3.6.3-1 respectively, as a result of eliminating the 
steam condensing mode of the Residual Heat Removal system.
    Date of issuance: August 23, 1994
    Effective date: August 23, 1994
    Amendment Nos. 74
    Facility Operating License No. NPF-39: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37076) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 23, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room: Pottstown Public Library, 500 High 
Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendments: January 10, 1994, as 
supplemented by letter dated July 20, 1994
    Brief description of amendments: The amendments relocate the 
seismic monitoring instrumentation Limiting Condition for Operation, 
Surveillance Requirements, and associated tables and Bases contained in 
TS Sections 3.3.7.2 and 4.3.7.2 to the Updated Final Safety Analysis 
Report, Section 3.7.4.
    Date of issuance: August 29, 1994
    Effective date: August 29, 1994
    Amendment Nos. 75 and 36
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12364) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 29, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room: Pottstown Public Library, 500 High 
Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Public Service Electric and Gas 
Company Delmarva Power and Light Company, and Atlantic City 
Electric Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic 
Power Station,Unit Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: March 28, 1994, as supplemented 
on June 27, 1994 and July 8, 1994
    Brief description of amendments: These amendments relocate the fire 
protection requirements from the Technical Specifications to the 
Updated Final Safety Analysis Report in accordance with the guidance in 
Generic Letter (GL) 86-10, ``Implementation of Fire Protection 
Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements 
from Technical Specifications.''
    Date of issuance: August 24, 1994
    Effective date: August 24, 1994
    Amendments Nos.: 194 and 198
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications and the licenses.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22012) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 24, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room:  Government Publications Section, State 
Library of Pennsylvania, (REGIONAL DEPOSITORY) Education Building, 
Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of application for amendment: February 3, 1994
    Brief description of amendment: The licensee commenced operating on 
a 24-month fuel cycle, instead of the previous 18-month fuel cycle, 
with fuel cycle 9. Fuel cycle 9 started in August 1992; however, the 
facility has been shut down since February 1993 for a ``Performance 
Improvement Outage'' and a restart date has not yet been established. 
In order to accommodate operation on a 24-month cycle after the 
facility restarts, the following Engineered Safety Features (ESF) 
instrument calibration intervals have been extended:
    (1) Reactor coolant temperature instrument channels (specified in 
TS Table 4.1-1)
    (2) Steam generator level instrument channels (specified in TS 
Table 4.1-1)
    (3) Containment pressure instrument channels (specified in TS Table 
4.1 1)
    (4) Steam line pressure instrument channels (specified in TS Table 
4.1-1)
    (5) Turbine first stage pressure instrument channels (specified in 
TS Table 4.1-1)
    (6) Turbine trip low auto stop oil pressure instrument channels 
(specified in TS Table 4.1-1)
    (7) 480V bus undervoltage and alarm relays (specified in TS Table 
4.1-1)
    These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle,'' as applicable.Additionally, the 
following changes were also incorporated:
    (8) A limiting conditions for operation requirement for a wide 
range containment pressure variable was added to TS Table 3.5-5 to 
ensure consistency with Regulatory Guide 1.97 commitments and the IP3 
Emergency Operating Procedures (EOPs).
    (9) A quarterly functional test surveillance requirement for the 
low average temperature actuation circuits of the reactor coolant 
temperature channels was added to Item 4 of TS Table 4.1-1.
    (10) Item 14 of TS Table 4.1-1 was expanded to specify surveillance 
requirements for the wide range containment pressure instrumentation 
channels.
    (11) Item 20 to TS Table 4.1-1 was revised to clarify that both the 
reactor trip and the ESF actuation relay logic channels are 
functionally tested.
    Date of issuance: September 1, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 150
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14894) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 1, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room: White Plains Public Library, 100 
Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: June 17, 1993, as supplemented 
February 24, 1994, and June 13, 1994
    Brief description of amendment: The amendment adds Section 3/
4.2.J., ``Remote Shutdown Capability,'' and associated Table 3.2-10, 
``Remote Shutdown Capability Instrumentation and Controls,'' to the 
Technical Specifications (TSs) to provide Limiting Conditions for 
Operation and surveillance requirements for the remote/alternate 
shutdown equipment. The amendment also adds an associated Bases section 
to the TSs. These additions to the TSs were based on NUREG-1433, 
``Standard Technical Specifications - General Electric Boiling Water 
Reactors (BWR/4).'' Several administrative changes were also made to 
accommodate the additions to the TSs.
    Date of issuance: August 31, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 216
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41511) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 31, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room: Reference and Documents Department, 
Penfield Library, State University of New York, Oswego, New York 13126.

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

    Date of application for amendments: March 4, 1994, as supplemented 
on June 14, 1994 and by phone on July 22, 1994
    Brief description of amendments: These amendments modify Section 
5.3.1 of the Technical Specifications (TS) to allow the use of 
Westinghouse Vantage+ fuel with ZIRLO cladding. The previous TS 
required the fuel cladding to be Zircaloy-4, which is used in the 
Westinghouse Standard and Vantage 5H fuel designs.
    Date of issuance: August 22, 1994
    Effective date: August 22, 1994
    Amendment Nos. 154 and 134
    Facility Operating License Nos. DPR-70 and DPR-75. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14896) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 22, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room: Salem Free Public Library, 112 West 
Broadway, Salem, New Jersey 08079

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

    Date of application for amendment: December 13, 1993, as 
supplemented February 2, 1994, and March 11, 1994.
    Brief description of amendment: The amendment changes the Technical 
Specifications to allow for the storage of fuel with an enrichment not 
to exceed a nominal 5.0 weight percent (w/o) U-235 in the VCSNS new 
(fresh) and spent fuel storage racks. The changes would also allow 
UO2 with a maximum nominal enrichment up to 5.0 w/o U-235 to be 
used as fuel in the VCSNS core.
    Date of issuance: August 23, 1994
    Effective date: August 23, 1994
    Amendment No.: 116
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12365) The March 11, 1994, letter provided clarifying information that 
did not change the initial determination of no significant hazards 
consideration as published in the Federal Register. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated August 23, 1994. No significant hazards consideration comments 
received: No
    Local Public Document Room: Fairfield County Library, Garden and 
Washington Streets, Winnsboro, South Carolina 29180.

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

    Date of application for amendment: May 20, 1994
    Brief description of amendment: The proposed amendment would remove 
Core Spray High Sparger Instrumentation from the Vermont Yankee 
Technical Specifications for Emergency Core Cooling System Actuation 
Instrumentation. In addition, an unrelated administrative change is 
also made.
    Date of issuance: August 22, 1994
    Effective date: August 22, 1994
    Amendment No.: 140
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34669) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 22, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room: Brooks Memorial Library, 224 Main 
Street, Brattleboro, Vermont 05301.

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: June 9, 1994
    Brief description of amendments: These amendments revise the NA-1&2 
Technical Specifications (TS) by removing the Reactor Trip System and 
the Engineered Safety Features Actuation System response times from the 
TS to station-controlled documents.
    Date of issuance: August 24, 1994
    Effective date: August 24, 1994
    Amendment Nos.: 187 and 168
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37088) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 24, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of application for amendment: January 6, 1994
    Brief description of amendment: This amendment relocates the 
requirements related to seismic monitoring instrumentation from the 
Technical Specifications (TS) to the Final Safety Analysis Report 
(FSAR) and plant procedures. The existing requirements will be 
maintained and controlled in accordance with the requirements of 10 CFR 
50.59 and TS 6.8.1.
    Date of issuance: August 22, 1994
    Effective date:  August 22, 1994, to be implemented within 30 days 
of issuance
    Amendment No.: 131
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14902) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 22, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room: Richland Public Library, 955 Northgate 
Street, Richland, Washington 99352.

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

    Date of application for amendments: September 29, 1993.
    Brief description of amendments: The amendments changed the 
inservice test frequency of the safety injection pumps, residual heat 
removal pumps, and containment spray pumps from monthly to quarterly. 
Also, the amendments added the administration of the inservice testing 
program to TS 15.4.2. The amendments added requirements to verify the 
containment sump suction is not blocked and to verify on a monthly 
basis, valve alignments of the emergency core cooling system and 
containment cooling systems.
    Date of issuance: August 25, 1994
    Effective date: Date of issuance to be implemented within 45 days
    Amendment Nos.: 150 and 154
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4949) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 25, 1994.No significant hazards 
consideration comments received: No.
    Local Public Document Room: Joseph P. Mann Library, 1516 Sixteenth 
Street, Two Rivers, Wisconsin 54241.

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

    Date of application for amendments: October 6, 1992
    Brief description of amendments: The amendments changed all 
references of rod position in the Technical Specifications to units of 
steps rather than inches. The amendments also changed Figure 15.3.10-1 
by referencing rod position in units of steps instead of percent 
withdrawn. Further, the amendments revised the basis for Section 
15.3.10 by clarifying the definition of ``fully withdrawn'' as it 
concerns Rod Cluster Control Assemblies, and modified the basis for 
Section 15.3.10 to be consistent with the above changes.
    Date of issuance: August 26, 1994
    Effective date: Immediately, to be implemented within 45 days.
    Amendment Nos.: 151 and 155
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1993 (58 FR 
16234) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 26, 1994No significant 
hazards consideration comments received: No.
    Local Public Document Room: Joseph P. Mann Library, 1516 Sixteenth 
Street, Two Rivers, Wisconsin 54241.

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

    Date of amendment request: June 7, 1994
    Brief description of amendment: The amendment revises Technical 
Specification Table 2.2-1, ``Reactor Trip System Instrumentation 
Setpoints,'' to change the over-temperature-delta-temperature (OTDT) 
axial flux difference (AFD) limits to reflect the results of the Cycle 
8 core maneuvering analysis.
    Date of issuance:  August 25, 1994
    Effective date:  August 25, 1994
    Amendment No.: 79
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34672) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated Augusty 25, 1994.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.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 
and at the local public document room for the particular facility 
involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By October 14, 1994, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC 20555 and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, 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 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendment: June 9, 1994, as supplemented 
August 10, 1994
    Brief description of amendment: This amendment increases the 
allowed out-of-service time from 7 days to 14 days for the automatic 
depressurization system, the high pressure coolant injection system, 
and the reactor core isolation cooling system. A change is also made to 
Section 4.5.H, ``Maintenance of Filled Discharge Pipe'' to reflect 
Amendment 149 issued September 28, 1993.
    Date of issuance: August 22, 1994
    Effective date: August 22, 1994
    Amendment No.: 156
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No The Commission's related 
evaluation of the amendment, consultation with the State, and final 
determination of no significant hazards consideration are contained in 
a Safety Evaluation dated
    Local Public Document Room: Plymouth Public Library, 11 North 
Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Walter R. Butler

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley 
Power Station, Unit 2, Shippingport, Pennsylvania

    Date of application for amendment: August 17, 1994
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) by revising Surveillance Requirement (SR) 4.6.2.2.d 
of Limiting Condition For Operation (LCO) 3.6.2.2, entitled 
``Containment Recirculation Spray System,'' by adding a new footnote 
number (1) pertaining to 2RSS*P21A pump performance requirements. In 
addition, SR 4.6.2.2.e.2 is revised by deleting the footnote, denoted 
by a single asterisk, which pertains to an extension to the 18-month 
surveillance interval for first fuel cycle.
    Date of issuance: August 22, 1994
    Effective date: As of the date of issuance.
    Amendment No: 62
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No. On August 17, 1994, the staff 
issued enforcement discretion, which was immediately effective and 
remained in effect until the staff's review of this amendment was 
completed.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, consultation with the Commonwealth of 
Pennsylvania and final no significant hazards considerations 
determination are contained in a Safety Evaluation dated August 22, 
1994.
    Local Public Document Room: B. F. Jones Memorial Library, 663 
Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Dated at Rockville, Maryland, this 7th day of September 1994.
    For The Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 94-22593 Filed 9-13-94; 8:45 am]
BILLING CODE 7590-01-F