94-27613. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

[Federal Register Volume 59, Number 216 (Wednesday, November 9, 1994)]
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[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-27613]


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[Federal Register: November 9, 1994]



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NUCLEAR REGULATORY COMMISSION
 

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from October 17, 1994, through October 28, 1994. 
The last biweekly notice was published on October 26, 1994 (59 FR 
53834).

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 December 9, 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.

Carolina Power & Light Company, et al.

Docket Nos. 50-325 and 50-324

    Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, 
North Carolina.
    Date of amendments request: September 30, 1994.
    Description of amendments request: The amendments would revise the 
Technical Specifications to eliminate the scram and isolation trip 
functions from the main steam line radiation monitor (MSLRM). This 
change would specifically remove the reactor scram, main steam line 
isolation valve closure, main steam line drain valve closure, reactor 
water sample line isolation, and mechanical vacuum pump line isolation 
actuated on a MSLRM High-High Radiation signal. The actuation signal 
for isolation of the reactor water sample line will be replaced with a 
low condenser vacuum signal. The isolation of the mechanical vacuum 
pump line will be changed to a signal from the main stack radiation 
monitor.
    The MSLRMs will have both High Radiation and High-High Radiation 
alarms. The setpoint for the MSLRM High Radiation alarm will be set at 
or below 1.5 times the nominal full power background radiation adjusted 
for Hydrogen water chemistry operation. The setpoint for the condenser 
off-gas radiation monitor will be set at a value of 1.5 times 
background radiation, but not less than 1.5 Rem per hour.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The deletion of the MSLRM trip function from the reactor scram and the 
Group 1 isolation initiation logic removes a potential transient 
initiation and therefore decreases the probability of plant transients 
occurring due to inadvertent scrams resulting from this system.
    The deletion of the MSLRM trip function from the Main Steam Drain 
Valve, the Reactor Water Sample Isolation Valve, and the Mechanical 
Vacuum Pump line isolation logic, does not affect the initiators of any 
accident previously evaluated in the Safety Analysis Report. Therefore, 
the proposed change does not involve an increase in the probability of 
occurrence of any accident previously evaluated.
    The NRC staff acceptance criterion for the Control Rod Drop 
Accident is that the doses from the accident fall significantly below 
the limits given in 10 CFR Part 100. The releases calculated for 
accident during plant operations when the Steam Jet Air Ejectors (SJAE) 
are operating and when the Mechanical Vacuum Pumps are operating are 
within these acceptance limits.
    In NEDO-31400, GE shows that the occurrence of a CRDA, with the MSL 
high radiation isolation removed, and SJAE in operation, results in 
offsite radiological exposures that are small fractions of 10CFR100 
guidelines. Since the Brunswick specific CRDA doses are lower than the 
[sic] calculated by GE and the GE dose parameters envelope those used 
for the Brunswick analysis, it is concluded that the NRC's findings 
that the radiological release consequence is within the staff's 
acceptance criteria, even without the automatic MSIV trip, is 
applicable to Brunswick.
    While not specifically addressed in the GE evaluation, Carolina 
Power and Light also proposes to eliminate the Main Steam Line Drain 
valves, the Reactor Water Sample Line isolation valves, and the 
mechanical vacuum line isolation valves from the MSLRM isolation logic. 
Main Steam Line Drain Valves B21-F016 and B21-F019 drain to the main 
condenser, which is the same flow path as the MSIVs. The discharge of 
both the MSIV and MSL drain flow paths is processed through the offgas 
system. Any radiation released through the drain valves during a 
control rod drop accident will be negligible and, for Brunswick, is 
bounded by the NEDO analysis.
    The reactor water sample line provides a small amount of reactor 
water to the Reactor Building Sample Panel. The discharge of the 
Reactor Building Sample Panel is routed through the floor drain sump to 
the liquid radwaste system. Any releases through this path would be 
negligible and, for Brunswick, is bounded by the NEDO analysis.
    The mechanical vacuum pumps are used only when the reactor is at 
low power (less than 5%) and there is insufficient steam flow to 
operate the Steam Jet Air Ejectors. The increase in radiation will be 
detected by the MSLRMs and annunciated in the Main Control Room. 
Operators will be instructed, in the annunciator response procedures, 
to take action to stop the Mechanical Vacuum Pump(s) and isolate the 
Mechanical Vacuum Pump line. The amount of radiation released prior to 
isolating the line would represent the most limiting case for this 
accident. However, it will still be well within 10 CFR Part 100 limits. 
Additionally, the dose received in the Main Control Room as a result of 
this accident is within General Design Criteria 19 (SRP 6.4) limits.
    Therefore, since elimination of the MSIV [sic, MSLRM] scram and 
isolation functions would not result in an increase in exposure above 
NRC acceptance limits, the proposed changes will not significantly 
increase the consequences of a previously evaluated accident.
    2. The proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The function of a MSLRM trip is to detect abnormal fission 
product release and isolate the steam lines, thereby stopping the 
transport of fission products from the reactor to the main condenser. 
The monitors do not perform a prevention function for any kind of 
accident. The existence of a MSLRM trip does not prevent the occurrence 
of a fuel failure event or any other type of event. The elimination of 
these signals, which served only in a mitigative function, does not 
create the possibility of a new or different kind of accident from 
those previously evaluated. Also, radiation monitors with alarm 
functions will remain installed in the plant to warn the operators of a 
high radiation condition in the main steam lines, or in the off-gas 
system. Thus no new or different accident can be postulated by the 
proposed changes.
    3. The proposed amendments do not involve a significant reduction 
in a margin of safety. As shown in the topical report, the changes 
represent an overall improvement in plant safety. Safe operation of the 
plant is further enhanced by elimination of the unnecessary scram and 
isolation of the reactor vessel. With implementation of these changes, 
1) the primary heat sink (main condenser) remains available, 2) large 
transients on the reactor vessel, as well as challenges to the ESF, are 
avoided, and 3) the Offgas system remains available to control the 
pathway of potential releases. As such, the margin of safety is 
enhanced by the proposed changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Acting Project Director: Michael L. Boyle.

Carolina Power & Light Company

Docket No. 50-261

    H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, 
South Carolina.
    Date of amendment request: October 7, 1994.
    Description of amendment request: The proposed amendment would 
revise the introduction to TS Section 6.9.3.3 to require the approved 
revision number for the referenced analytical methods be listed in the 
Core Operating Limits Report. The methodology referenced in 6.9.3.3.b.f 
(XN-NF-82-49(A)) will be updated to clarify that all supplements are 
included. New methodologies ANF-89-151(A) and EMF-92-081(A) will be 
added to TS Section 6.9.3.3.b.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated. 
The proposed changes will have no influence on the probability of an 
accident previously evaluated. No changes will be made to any safety 
related equipment, systems, or setpoints used in determining the 
probability of an evaluated accident. The plant design basis will not 
be altered. Therefore, there will be no significant increase in the 
probability of an accident previously evaluated.
    Consequences are dependent on the type of accident and the 
mitigating response of safety related equipment. Furthermore, the 
magnitude of consequences are calculated, directly or through 
supporting calculations, by use of NRC approved methodologies. The 
proposed license amendment will not alter the function of safety 
related equipment designed to mitigate the consequences of an accident 
previously evaluated or allow operation of the facility outside any 
current limitations or restrictions. Also, this amendment will not 
alter the requirement that evaluation of the consequences of an 
accident previously evaluated by determined/supported with NRC reviewed 
and approved methodologies. The change to TS Section 6.9.3.3.b's 
introductory wording satisfies an administrative commitment and the 
requirements it adds are administrative in nature. Accordingly the 
proposed license amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated. 
The addition of and update to NRC previously reviewed and approved 
methodologies in TS Section 6.9.3.3.b will not result in any design or 
function changes to any safety related equipment designed to prevent 
and/or mitigate accidents, to any setpoints or systems, or to any 
portion of the plant design basis. Operation of the facility will 
remain within all required limitations and/or restrictions. The change 
to TS Section 6.9.3.3.b's introductory wording satisfies an 
administrative commitment and the requirements it adds are 
administrative in nature. Therefore, the proposed amendment will not 
create the possibility of a new kind of accident from any accident 
previously evaluated.
    The addition of and update to NRC previously reviewed and approved 
methodologies in TS Section 6.9.3.3.b will not result in any design or 
function changes to any safety related equipment designed to prevent 
and or mitigate accidents, to any setpoints or systems, or to any 
portion of the plant design basis. Operation of the facility will 
remain within all required limitations and/or restrictions. The changes 
to TS Section 6.9.3.3.b's introductory wording satisfies an 
administrative commitment and the requirements it adds are 
administrative in nature. Therefore, the proposed amendment will not 
create the possibility of a different kind of accident from any 
accident previously evaluated.
    3. The proposed amendment does not involve a significant reduction 
in the margin of safety. The proposed license amendment is defined as 
administrative in nature. No current operational limits, restrictions, 
or operating modes of the facility and its equipment, safety related or 
otherwise, designed to preserve the margin of safety will be changed or 
affected by the proposed amendment. There will be no changes to 
setpoints or to the plant design basis. The methodology proposed for 
addition to TS Section 6.9.3.3.b and the methodology that will be 
updated has been previously reviewed and approved by the NRC. The 
change to TS Section 6.9.3.3.b's introductory wording satisfies an 
administrative commitment and the requirements it adds are 
administrative in nature. Accordingly the proposed license amendment 
will not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: R.E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: William H. Bateman.

Entergy Operations, Inc., et al.

Docket No. 50-416

    Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi.
    Date of amendment request: October 12, 1994.
    Description of amendment request: The proposed amendment requests 
the closure and deletion of License Condition 2.C.(26) related to 
turbine disk integrity.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. No significant increase in the probability or consequences of an 
accident previously evaluated results from this change.
    The proposed change would close and delete License Condition 
2.C.(26). The approved methodology currently used to evaluate the 
probability of rotor failure and the inspection interval will not be 
changed. The closure and deletion of the license condition is an 
administrative change and will affect any accident previously 
evaluated.
    The bounding accident for the turbine-generator as analyzed in the 
Grand Gulf Nuclear Station (GGNS) Updated Final Safety Analysis Report 
(UFSAR) is the occurrence of an external missile resulting from the 
failure of a low pressure (LP) turbine disc. The probability of this 
incident occurring is less than 1 x 10-5 per year, which is the 
NRC acceptable failure criterion for probability.
    Any extension to the service interval in the future will be 
evaluated in accordance with the current methodology. The original 
acceptable levels of failure will be maintained. Therefore, no 
significant increase in the probability or consequences of a previously 
evaluated accident results from this change.
    2. The change would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed change does not involve a change to the control logic 
or operating procedures for the turbine but rather transfers the 
control of the LP turbine disc inspection interval from the Operating 
License to administrative control. The current approved methodology 
will continue to be used when determining future inspection intervals.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The change would not involve a significant reduction in a margin 
of safety.
    Closing and deleting the current license condition for LP turbine 
disc inspections and controlling the inspection interval 
administratively has no adverse effects to the margin of safety. The 
current approved methodology for failures will continue to be used and 
any changes to future inspection intervals will be evaluated by the 
methodology. This change does not affect any previous safety analysis 
presented in the UFSAR and does not affect the criteria used to 
establish safety limits, the basis for limiting safety system settings, 
the basis for limiting conditions of operation, a change to the 
technical specifications or a change in plant operations.
    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, 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.

Florida Power and Light Company

Docket Nos. 50-250 and 50-251

    Turkey Point Plant, Units 3 and 4, Dade County, Florida.
    Date of amendment request: October 20, 1994.
    Description of amendment request: The licensee proposes to change 
Turkey Point, Units 3 and 4 Technical Specifications (TS) by revising 
TS 1.9, Definitions--CORE ALTERATIONS to only address activities which 
may, in actuality, affect core reactivity. In addition, the licensee 
proposes to revise TS 3.9.4, Containment Building Penetrations to allow 
both containment personnel airlock (PAL) doors to be open during core 
alterations and movement of irradiated fuel in containment provided (a) 
that at least one PAL door is capable of being closed; (b) the plant is 
in Mode 6 with at least 23 feet of water above the fuel; and (c) a 
designated individual is available outside the PAL to close the door. 
The licensee also proposes a revision to the footnote of TS 3.9.4, to 
remove the description of the purpose for imposing administrative 
controls.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The change in the definition of CORE ALTERATIONS would allow the 
movement of a temporary source range detector or other small 
components, such as cameras, tools, etc., within the reactor vessel 
without the activity being considered CORE ALTERATIONS. The potential 
exists, however small, that an object can be dropped into the reactor 
vessel. However, the justification for this change, is that the 
insertion of small components into the reactor vessel will have no 
effect on core reactivity since these items displace a small volume of 
borated water, and sufficient borated water will surround the 
components and provide the necessary neutron absorption to 
neutronically isolate the components from the reactor. The consequences 
of dropping one of these small components into the vessel are bounded 
by the In-Containment Fuel Handling Accident Analysis discussed in 
Chapter 14.2.1 of the Turkey Point Updated Final Safety Analysis Report 
(UFSAR). Therefore, the proposed change is bounded by the current and 
the proposed In-Containment Fuel Handling Accident Analyses and will 
not involve a significant increase in the probability or consequences 
of an accident previously evaluated.
    The proposed change to TS 3.9.4 would allow the containment 
personnel airlock (PAL) doors to be open during fuel movement and core 
alterations. Currently, a single PAL door is closed during fuel 
movement and core alterations to prevent the escape of radioactive 
material in the event of a in-containment fuel handling accident. The 
PAL is not an initiator of an accident. Whether the PAL doors are open 
or closed during fuel movement and core alterations has no affect on 
the probability of any accident previously evaluated.
    Allowing the PAL doors to be open during fuel movement and core 
alterations does not increase the consequences from a fuel handling 
accident. The calculated offsite doses are well within the limits of 10 
CFR Part 100. In addition, the calculated doses are larger than the 
expected doses because the calculation does not incorporate the closing 
of the PAL door after the containment is evacuated. The proposed change 
should significantly reduce the dose to workers in containment in the 
event of a fuel handling accident by reducing the time required to 
evacuate the containment. The proposed change will also significantly 
decrease the wear on the PAL doors and, consequently, increase the 
availability of the PAL doors in the event of an accident.
    The proposed change to the footnote of TS 3.9.4 is administrative 
in nature, and does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The changes being proposed do not affect assumptions contained in 
plant safety analyses or the physical design of the plant, nor do they 
affect Technical Specifications that preserve safety analysis 
assumptions. Therefore, operation of the facility in accordance with 
the proposed amendments would not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The change in the definition of CORE ALTERATIONS would allow the 
movement of a temporary source range detector or other small 
components, such as cameras, tools, etc., within the reactor vessel 
without the activity being considered CORE ALTERATIONS. The potential 
exists however small, that an object can be dropped into the reactor 
vessel. However, the justification for this change, is that the 
insertion of small components into the reactor vessel will have no 
effect on core reactivity since these items displace a small volume of 
borated water, and sufficient borated water will surround the 
components and provide the necessary neutron absorption to 
neutronically isolate the components from the reactor. The consequences 
of dropping one of these small components into the vessel are bounded 
by the In-Containment Fuel Handling Accident Analysis discussed in 
Chapter 14.2.1 of the Turkey Point UFSAR. Therefore the proposed change 
is bounded by the current and the proposed In-Containment Fuel Handling 
Accident Analyses and will not create the possibility of a new or 
different kind of accident.
    The proposed change to Specification 3.9.4 affects a previously 
evaluated accident, i.e., in-containment fuel handling accident. Both 
the current and the proposed In-Containment Fuel Handling Accident 
Analysis assume that all of the iodines and noble gases that become 
airborne within the containment escape and reach the site boundary and 
low population zone with no credit taken for the containment building 
barrier or for decay or deposition taken. Since the proposed change 
does not involve the addition or modification of equipment nor does it 
alter the design of plant systems and the revised analysis is 
consistent with the current In-Containment Fuel Handling Accident 
Analysis, the proposed change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed change to the footnote of TS 3.9.4 is administrative 
in nature and does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The change in the definition of CORE ALTERATIONS would allow the 
movement of a temporary source range detector or other small 
components, such as cameras, tools, etc., within the reactor vessel 
without the activity being considered CORE ALTERATIONS. The potential 
exists however small, that an object can be dropped into the reactor 
vessel. However, the justification for this change, is that the 
insertion of small components into the reactor vessel will have no 
effect on core reactivity since these items displace a small volume of 
borated water, and sufficient borated water will surround the 
components and provide the necessary neutron absorption to 
neutronically isolate the components from the reactor. The consequences 
of dropping one of these small components into the vessel are bounded 
by the Fuel Handling Accident Analysis discussed in Chapter 14.2.1 of 
the Turkey Point UFSAR. Therefore, the proposed change is bound by the 
current In-Containment Fuel Handling Accident Analyses and as a result 
will not involve a significant reduction in a margin of safety.
    The margin of safety as defined by 10 CFR Part 100 has not been 
reduced. There is no increase in calculated offsite dose resulting from 
a fuel handling accident in containment and the calculated dose is a 
small fraction of the limits given in 10 CFR Part 100. The proposed 
changes do not alter the bases for assurance that safety-related 
activities are performed correctly or the basis for any Technical 
Specification that is related to the establishment of or maintenance of 
a safety margin. Therefore, operation of the facility in accordance 
with the proposed amendments would not involve a significant reduction 
in a margin of safety.
    The proposed change to the footnote of TS 3.9.4 is administrative 
in nature and does not relate to or modify the safety margins defined 
in, and maintained by, the Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer, 
P.C., 1615 L Street, NW., Washington, DC 20036.
    NRC Project Director: Mohan C. Thadani, (Acting)

Florida Power and Light Company

Docket Nos. 50-250 and 50-251

    Turkey Point Plant, Units 3 and 4, Dade County, Florida.
    Date of amendment request: October 20, 1994.
    Description of amendment request: This supersedes the licensee's 
original request dated July 19, 1994, and noticed in the Federal 
Register on August 3, 1994 (59 FR 39588). The licensee proposes to 
change Turkey Point, Units 3 and 4 Technical Specifications (TS) 
4.8.1.1.2e. and 4.8.1.1.2f., which address Emergency Diesel Generator 
(EDG) fuel oil testing, by replacing the specific EDG fuel oil 
Surveillance Requirements with the requirement to verify new and stored 
EDG fuel oil in accordance with the Diesel Fuel Oil Testing Program. In 
addition, the licensee proposes the addition of ACTION statements g. 
and h., to TS 3.8.1.1, to address the required action in the event the 
diesel fuel oil does not meet the Diesel Fuel Oil Testing Program 
limits. The Diesel Fuel Oil Testing Program will be described in both 
TS 6.8.4 and the BASES Section to the Technical Specifications. In 
addition, FPL proposes revising TS 6.8.1 to include the requirement 
that written procedures shall be established, implemented and 
maintained for implementation of the Diesel Fuel Oil Testing Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed changes to the Technical Specifications will permit 
the Technical Specification required testing of Emergency Diesel 
Generator (EDG) fuel oil in accordance with the Turkey Point, Units 3 
and 4 Diesel Fuel Oil Testing Program. The proposed change will permit 
FPL to use more recent editions of the American Society for Testing and 
Materials (ASTM) standards currently listed in Technical Specification 
Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f. Prior to changing 
the Diesel Fuel Oil Testing Program, the proposed change will be 
evaluated pursuant to Title 10 Code of Federal Regulations Sec. 50.59 
(10 CFR Sec. 50.59), ``Changes, tests, and experiments.'' Title 10 CFR 
Sec. 50.59 permits a licensee to make changes in the procedures as 
described in the safety analysis report without prior Commission 
approval, provided that the proposed changes does not involve an 
unreviewed safety question.
    Title 10 CFR Sec. 50.59(a)(2) states that a proposed change 
involves an unreviewed safety question (i) if the probability of 
occurrence or the consequences of an accident or malfunction of 
equipment important to safety previously evaluated in the safety 
analysis report may be increased. Consequently, since any change to the 
Diesel Fuel Oil Testing Program, including the ASTM standard or ASTM 
edition standard to be used to evaluate EDG fuel oil acceptability, the 
change must be evaluated relative to the more restrictive evaluation 
criterion of 10 CFR Sec. 50.59, then operation of the facility in 
accordance with the proposed amendments would not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The EDG fuel oil TS Surveillance Requirements will be 
replaced with a requirement to test the EDG fuel oil in accordance with 
the Turkey Point Units 3 and 4 Diesel Fuel Oil Testing Program.
    ACTION statement g. of TS 3.8.1.1 is added to address the required 
action in the event the new fuel oil properties do not meet the Diesel 
Fuel Oil Testing Program limits. A failure to meet the American 
Petroleum Institute (API) gravity, kinematic viscosity, flash point or 
clarity limits is cause for rejecting the new fuel oil prior to the 
addition to the Diesel Fuel Oil Storage Tanks, but does not represent a 
failure to meet the Limiting Condition for Operation (LCO) of TS 
3.8.1.1, since the new fuel oil has not been added to the storage 
tanks. Provided these new fuel oil properties are met subsequent to the 
addition of the new fuel oil to the storage tanks, 30 days is provided 
to complete the analyses of the other fuel oil properties specified in 
Table 1 of ASTM-D975-81, except sulfur which may be performed in 
accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the other 
new fuel oil properties specified in Table 1 of ASTM-D975-81 are not 
met, ACTION statement g. of TS 3.8.1.1 provides an additional 30 days 
to meet the Diesel Fuel Oil Testing Program limits. This additional 30 
day period is acceptable because the fuel oil properties of interest, 
even if they are not within limits, would not have an immediate effect 
on EDG operation.
    ACTION statement h. of TS 3.8.1.1 is added to address the required 
action in the event the stored fuel oil total particulates do not meet 
the Diesel Fuel Oil Testing Program limits. Fuel oil degradation during 
long term storage shows up as an increase in particulate, due mostly to 
oxidation. The presence of particulate does not mean the fuel oil will 
not burn properly in a diesel engine. The frequency for performing 
surveillance on stored fuel oil is based on stored fuel oil degradation 
trends which indicate that particulate concentration is unlikely to 
change significantly between surveillances.
    Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil 
Testing Program, FPL will need to determine if the proposed program 
change is at least as, if not more, effective, in detecting 
unsatisfactory fuel oil. The EDGs will thus continue to function as 
designed and the probability or consequences of previously evaluated 
accidents will be unaffected.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes to the Technical Specifications will permit 
the Technical Specification required testing of Emergency Diesel 
Generator fuel oil using more recent editions of the American Society 
for Testing and Materials (ASTM) standards currently listed in 
Technical Specification Surveillance Requirements 4.8.1.1.2e. and 
4.8.1.1.2f. Prior to changing the edition of the previously approved 
ASTM standard being used to evaluate the EDG fuel oil, the proposed 
edition standard will be evaluated pursuant to 10 CFR Sec. 50.59, 
``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits a 
licensee to make changes in the procedures as described in the safety 
analysis report without prior Commission approval, provided that the 
proposed changes does not involve an unreviewed safety question. Title 
10 CFR Sec. 50.59(a)(2) states that a proposed change involves an 
unreviewed safety question (ii) if a possibility for an accident or 
malfunction of a different type than any evaluated previously in the 
safety analysis report may be created. Consequently, since any change 
to the edition of the ASTM standard to be used to evaluate EDG fuel oil 
acceptability must be evaluated relative to the more restrictive 
evaluation criterion of 10 CFR Sec. 50.59, then operation of the 
facility in accordance with the proposed amendments would not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    ACTION statement g. of TS 3.8.1.1 is added to address the required 
action in the event the new fuel oil properties do not meet the Diesel 
Fuel Oil Testing Program limits. A failure to meet the API gravity, 
kinematic viscosity, flash point or clarity limits is cause for 
rejecting the new fuel oil prior to the addition to the Diesel Fuel Oil 
Storage Tanks, but does not represent a failure to meet the Limiting 
Condition for Operation (LCO) of TS 3.8.1.1, since the new fuel oil has 
not been added to the storage tanks. Provided these new fuel oil 
properties are met subsequent to the addition of the new fuel oil to 
the storage tanks, 30 days is provided to complete the analyses of the 
other fuel oil properties specified in Table 1 of ASTM-D975-81, except 
sulfur which may be performed in accordance with ASTM-D1552-79 or ASTM-
D2622-82. In the event the other new fuel oil properties specified in 
Table 1 of ASTM-D975-81 are not met, ACTION statement g. of TS 3.8.1.1 
provides an additional 30 days to meet the Diesel Fuel Oil Testing 
Program limits. This additional 30 day period is acceptable because the 
fuel oil properties of interest, even if they are not within limits, 
would not have an immediate effect on EDG operation.
    ACTION statement h. of TS 3.8.1.1 is added to address the required 
action in the event the stored fuel oil total particulates does not 
meet the Diesel Fuel Oil Testing Program limits. Fuel oil degradation 
during long term storage shows up as an increase in particulate, due 
mostly to oxidation. The presence of particulate does not mean the fuel 
oil will not burn properly in a diesel engine. The frequency for 
performing surveillance on stored fuel oil is based on stored fuel oil 
degradation trends which indicate that particulate concentration is 
unlikely to change significantly between surveillances.
    Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil 
Testing Program, FPL will need to determine if the proposed program 
change is at least as, if not more, effective, in detecting 
unsatisfactory fuel oil. Since the proposed changes do not involve a 
change in the design of any plant system or component, and since the 
proposed changes will need to evaluate the effect of any ASTM standard 
edition change on the level of EDG reliability, the change proposed 
will not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes to the Technical Specifications will permit 
the Technical Specification required testing of Emergency Diesel 
Generator (EDG) fuel oil using more recent editions of the American 
Society for Testing and Materials (ASTM) standards currently listed in 
Technical Specification Surveillance Requirements 4.8.1.1.2e. and 
4.8.1.1.2f. Prior to changing the edition of the previously approved 
ASTM standard being used to evaluate the EDG fuel oil, the proposed 
edition standard will be evaluated pursuant to 10 CFR Sec. 50.59, 
``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits a 
licensee to make changes in the procedures as described in the safety 
analysis report without prior NRC approval, provided that the proposed 
changes does not involve an unreviewed safety question. Title 10 CFR 
Sec. 50.59(a)(2) states that a proposed change involves an unreviewed 
safety question (iii) if the margin of safety as defined in the basis 
for any technical specification is reduced. Consequently, since any 
change to the edition of the ASTM standard to be used to evaluate EDG 
fuel oil acceptability must be evaluated relative to the more 
restrictive evaluation criterion of 10 CFR Sec. 50.59, then operation 
of the facility in accordance with the proposed amendments would not 
involve a significant reduction in a margin of safety.
    ACTION statement g. of TS 3.8.1.1 is added to address the required 
action in the event the new fuel oil properties do not meet the Diesel 
Fuel Oil Testing Program limits. A failure to meet the API gravity, 
kinematic viscosity, flash point or clarity limits is cause for 
rejecting the new fuel oil prior to the addition to the Diesel Fuel Oil 
Storage Tanks, but does not represent a failure to meet the Limiting 
Condition for Operation (LCO) of TS 3.8.1.1, since the new fuel oil has 
not been added to the storage tanks. Provided these new fuel oil 
properties are met subsequent to the addition of the new fuel oil to 
the storage tanks, 30 days is provided to complete the analyses of the 
other fuel oil properties specified in Table 1 of ASTM-D975-81, except 
sulfur which may be performed in accordance with ASTM-D1552-79 or ASTM-
D2622-82. In the event the other new fuel oil properties specified in 
Table 1 of ASTM-D975-81 are not met, ACTION statement g. of TS 3.8.1.1 
provides an additional 30 days to meet the Diesel Fuel Oil Testing 
Program limits. This additional 30 day period is acceptable because the 
fuel oil properties of interest, even if they are not within limits, 
would not have an immediate effect on EDG operation.
    ACTION statement h. of TS 3.8.1.1 is added to address the required 
action in the event the stored fuel oil total particulates does not 
meet the Diesel Fuel Oil Testing Program limits. Fuel oil degradation 
during long term storage shows up as an increase in particulate, due 
mostly to oxidation. The presence of particulate does not mean the fuel 
oil will not burn properly in a diesel engine. The frequency for 
performing surveillance on stored fuel oil is based on stored fuel oil 
degradation trends which indicate that particulate concentration is 
unlikely to change significantly between surveillances.
    Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil 
Testing Program, FPL will need to determine if the proposed program 
change is at least as, if not more, effective, in detecting 
unsatisfactory fuel oil. Since the proposed changes will require a 
safety evaluation to assure that the reliability of the EDGs using fuel 
oil tested in accordance with the different ASTM standard edition 
maintains the current margin of safety, 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer, 
P.C., 1615 L Street, NW., Washington, DC 20036.
    NRC Project Director: Mohan C. Thadani, Acting.

Florida Power Corporation, et al.

Docket No. 50-302

    Crystal River Nuclear Generating Plant, Unit No. 3, Citrus County, 
Florida.
    Date of amendment request: September 30, 1994.
    Description of amendment request: The proposed amendment would 
revise the Crystal River 3 (CR3) Nuclear generating Plant Technical 
Specifications (TS) to allow an increase in the rated thermal power 
(RTP) for CR-3 from the current 2544 level to 2568 Megawatt thermal 
(Wt). Accordingly, in TS 1.1, ``Definitions,'' would be revised to 
indicate the new power level of 2568 MWt. The proposed change would not 
require any hardware modifications.
    Basis for proposed no significant hazards consideration 
determination: Currently, CR-3 is operating at a maximum RTP of 2544 
MWt. The licensee proposes to operate at a maximum RTP of 2568 MWt, an 
increase of 24 MWt over the current licensed power of 2544 MWt.
    The licensee states that the Babcock and Wilcox (B&W) 177 Fuel 
Assembly (FA) Nuclear Steam Supply System (NSSS) in the CR3 design is 
capable of operating at a thermal power level of 2772 MWt. Due to 
limitations in the secondary area of the plant, the licensee requests 
authorization to operate at 2568 MWt which is less than the design 
level of 2772 MWt. The licensee performed a detailed engineering study 
on this power increase.
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the probability 
of occurrence or consequences of an accident previously evaluated. The 
thermal-hydraulic and nuclear characteristics of the reactor core were 
originally designed for a rated thermal power of 2568 MWt or higher. 
Therefore, the proposed thermal power increase to the reference power 
level of 2568 MWt does not change the original design assumptions and 
analyses for the reactor core. Most of the design basis accidents and 
transients were originally evaluated at the proposed power level. As 
described more fully in this submittal, those transients and accidents 
that were not originally evaluated at 2568 MWt were re-evaluated using 
CR-3 FSAR [Final Safety Analysis Report] Chapter 14 accident sequence 
of events, reactor protection criteria, and approved calculational 
methods. Based on this evaluation and initial plant design evaluations, 
FPC [Florida Power Corporation, the licensee for CR3] has determined 
that the probability and consequences of design basis transients and 
accidents are not significantly increased and that the radiological 
consequences from the design basis transients and accidents remain well 
below 10 CFR 100 limits.
    FPC has also reviewed CR-3 balance of plant and safety related 
systems to determine which systems and components could be affected by 
the proposed power increase. The changes to the reactor coolant system 
and secondary conditions and parameters are discussed in this 
submittal. These changes are minor in nature. The only Technical 
Specification change is to revise the reference power to 2568 MWt. No 
facility modifications will be required. FPC evaluated the systems and 
components and concluded that these systems and components will 
continue to perform within their design parameters with the unit 
operating at 2568 MWt.
    Based on the foregoing, the proposed amendment does not 
significantly increase the probability or consequences of an accident 
previously evaluated.
    2. The proposed thermal power increase does not create the 
possibility of a new or different kind of accident from previously 
evaluated accidents. As noted above, the thermal-hydraulic and nuclear 
characteristics of the reactor core were originally designed for 
operation at the proposed thermal power. Therefore, operation at the 
proposed power level does not introduce new or different performance 
characteristics that create the possibility of a new or different kind 
of accident.
    FPC has also reviewed CR-3 safety-related systems and balance of 
plant systems to determine which systems could be affected by the 
proposed power increase and the resultant minor changes in plant 
parameters and operating conditions. Systems that could be affected 
were evaluated using the licensing basis criteria described in the CR-3 
FSAR to assure their adequacy at the increased power level. Included in 
these evaluations were plant features that are not power level related 
or directly affected by an increase in power level, as well as, 
associated issues such as environmental considerations. Equipment 
performance and plant operation were evaluated with respect to actual 
performance versus projected operating conditions to identify any 
hardware modifications required to achieve the upgraded power. Based on 
this evaluation, FPC has determined that all systems will continue to 
perform within their design parameters at 2568 MWt and that no physical 
modifications to these systems will be necessary to accommodate a 2568 
MWt rating. Only minor re-calibration of plant instrumentation to 
reflect the increased power will be needed. The proposed power level 
does not introduce any new performance characteristics or modes of 
operation for plant systems and components, and does not introduce any 
new failure modes.
    Based on the foregoing, the proposed amendment does not create the 
possibility of a new or different kind of accident.
    3. The proposed amendment does not involve a significant reduction 
in a margin of safety. The thermal-hydraulic and nuclear 
characteristics of the reactor core were originally designed for 
operation at the proposed power level. Most of the design basis 
transients and accidents were originally analyzed assuming a power 
level of 2568 MWt or higher. As described more fully in this submittal, 
those transients and accidents that were not originally analyzed at 
2568 MWt were re-evaluated using CR-3 FSAR Chapter 14 accident sequence 
of events, reactor protection criteria, and approved calculational 
methods. FPC has determined that operation with the proposed thermal 
power will be bounded by the original analyses. In addition, FPC's 
evaluation of affected plant systems and components revealed that plant 
systems and components will continue to operate within their design 
parameters with no significant change in a margin of safety.
    Based on the foregoing, the proposed amendment does not involve a 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629
    Attorney for licensee: A. H. Stephens, General Counsel, Florida 
Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida 
33733.
    NRC Project Director: Mohan C. Thadani, (Acting).

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 amendment request: August 3, 1994.
    Description of amendment requests: The proposed amendments would 
allow the radiological effluent technical specifications (TS) to be 
relocated to other controlled documents. Procedural details contained 
in the current radiological effluents TS have been relocated to either 
the Offsite Dose Calculation Manual (OCDM) or the Process Control 
Program (PCP), as applicable. Proposed revisions to the OCDM and PCP 
have been prepared in accordance with the proposed changes to the 
administrative controls section of the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    The changes described above in no way negatively impact the 
requirements of the T/Ss. Separating the turbine room sump releases 
from the others is purely a clarification of the method we handle 
releases. The six ground monitoring wells added to the T/S table 
updates our current monitoring practice. With the six extra wells to 
monitor, we exceed the monitoring requirements of the T/Ss. Therefore, 
it is concluded that the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    Criterion 2
    No changes to the LCOs for either T/S are proposed as part of this 
amendment request. The proposed change does not involve any physical 
changes to the plant or any changes to plant operations. The changes 
merely propose to update our methods of implementing the T/S with our 
current practices. Thus, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Criterion 3
    The changes described above in no way negatively impact the 
requirements of the T/Ss. Separating the turbine room sump releases 
from the others is purely a clarification of the method we handle 
releases. The six ground monitoring wells added to the T/S table 
updates our current monitoring practice. With the extra wells to 
monitor, we exceed the monitoring requirements called for in the T/Ss. 
Therefore, it is concluded that the proposed changes do not involve a 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

Niagara Mohawk Power Corporation

Docket No. 50-410

    Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.
    Date of amendment request: October 5, 1994.
    Description of amendment request: The proposed license amendment 
would revise the applicability requirements of Technical Specification 
(TS) 3.7.3 to require operability of the Control Room Outdoor Air 
Special Filter Train System in Operational Conditions 1, 2, 3 and ** 
(when irradiated fuel is being handled in the reactor building and 
during CORE ALTERATIONS and operations with a potential for draining 
the reactor vessel and uncovering irradiated fuel) rather than in all 
Operational Conditions and * * *. The applicability requirements for 
Action Statement b of TS 3.7.3 and for the Radiation Monitoring 
Instrumentation required operable by TS Tables 3.3.7.1-1 and 4.3.7.1-1 
would be changed in a similar manner. The proposed amendment would also 
add a notation to Action Statement b.1 of TS 3.7.3 stating that the 
provisions of Specification 3.0.4 are not applicable provided an 
operable control room filter train is in the emergency pressurization 
mode of operation. The licensee stated that these proposed changes are 
consistent with the requirements of the NRC's Improved Standard 
Technical Specifications (NUREG-1433) and with Generic Letter 87-09, 
``Section 3.0 and 4.0 of the Standard Technical Specifications (STS) on 
the Applicability of Limiting Conditions for Operation and Surveillance 
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 operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The Control Room Outdoor Air Special Filter Train System is not an 
initiator or precursor to an accident. The Control Room Outdoor Air 
Special Filter Train System responds to a release of radioactivity to 
the outside environment as detected in the air supply to the control 
room by providing a radiologically controlled environment within the 
control room. In operational conditions 4 and 5, the probability and 
consequences of a design basis accident are reduced due to the pressure 
and temperature limitations in these operational conditions. Therefore, 
maintaining the chiller subsystem operable is not required in 
operational conditions 4 and 5, except for the * * * operational 
condition. Therefore, a change to applicability and action statements 
of LCO [Limiting Condition For Operation] 3.7.3 cannot affect the 
probability of a previously evaluated accident.
    All accidents which take credit for operation of the Control Room 
Outdoor Air Special Filter Train System in the emergency pressurization 
mode of operation are analyzed and presented in Chapter 15 of the USAR 
[Updated Safety Analysis Report]. These accidents can only occur in 
operational conditions 1, 2, 3 and * * *.
    Accordingly, the proposed change in the applicability of LCO 3.7.3 
from all operational conditions (i.e., 1, 2, 3, 4, 5 and * * *) to 
operational conditions 1, 2, 3 and * * * does not significantly 
increase the consequences of an accident previously evaluated. The 
proposed change to action statement b of LCO 3.7.3 and to Tables 
3.3.7.1-1 and 4.3.7.1-1 of LCO 3.3.7.1 is consistent with the above 
change.
    Sections 15.7.4 and 15.7.5 of the USAR evaluate a fuel handling 
accident and a spent fuel cask drop accident, respectively. The 
radiological evaluation of these accidents considers the unfiltered 
radioactivity that enters the control room prior to the automatic 
operation of the Control Room Outdoor Special Filter Train System in 
the emergency pressurization mode of operation. The radiological 
consequences of these accidents are within the limits of GDC [General 
Design Criterion]-19.
    With one control room filter train inoperable and prior to entering 
the operational condition, the proposed change to action statement b.1 
of LCO 3.7.3 would require an operable control room filter train be 
placed in the emergency pressurization mode of operation. During an 
accident involving the release of radioactivity to the environment, an 
operable control room filter train would already be running in the 
emergency pressurization mode and performing its safety function, 
thereby preventing the entry of unfiltered radioactivity into the 
control room. Therefore, if a fuel handling accident or a spent fuel 
cask drop accident were to occur and release radioactivity, the control 
room personnel radiological doses would be less than the doses depicted 
in the USAR. Accordingly, the Technical Specification change to action 
statement b.1 does not significantly increase the consequences of a 
previously evaluated accident.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This amendment does not involve any accident precursors or 
initiators. In addition, this amendment does not require any changes to 
plant equipment.
    During an accident involving the release of radioactivity to the 
environment an operable control room filter train would already be 
running in the emergency pressurization mode and performing its safety 
function. Furthermore, the operating status of a running control room 
filter train would be unaffected by the receipt of an automatic start 
signal due to high radiation in either air intake to the Control Room 
Outdoor Air Special Filter Train System. Therefore, the proposed 
amendment will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed change in the applicability of LCO 3.7.3 from all 
operational conditions (i.e., 1, 2, 3, 4, 5 and * * *) to operational 
conditions 1, 2, 3 and * * * is consistent with the safety analysis 
contained in the USAR. The proposed changes to action statement b of 
LCO 3.7.3 and to Tables 3.3.7.1-1 and 4.3.7.1-1 of LCO 3.3.7.1 is 
consistent with the above change.
    Entry into the ** operational condition for LCO 3.7.3 with one 
control room filter train inoperable and the other control room filter 
train operable and operating in the emergency pressurization mode 
provides a comparable level of safety to two operable non-running 
control room filter trains. The remedial measure prescribed by 
Technical Specification action statement b.1 (placing an operable 
control room filter train in the emergency pressurization mode of 
operation) for which the exception to LCO 3.0.4 is proposed provides a 
sufficient level of protection to permit operational mode changes and 
safe long-term operation of NMP2 [Nine Mile Point Unit 2] consistent 
with the licensing basis described in the USAR. Therefore, the proposed 
change to action statement b.1 is consistent with Generic Letter 87-09, 
``Sections 3.0 and 4.0 of the Standard Technical Specifications (STS) 
on the Applicability of Limiting Conditions for Operation and 
Surveillance Requirements.'' Accordingly, this change will not 
significantly reduce the margin of safety.
    This proposed amendment is consistent with the Improved Standard 
Technical Specifications, NUREG-1433. Accordingly, as determined by the 
analysis above, this proposed amendment involves no significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Ledyard B. Marsh.

Niagara Mohawk Power Corporation

Docket No. 50-410

    Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.
    Date of amendment request: October 21, 1994.
    Description of amendment request: The proposed amendment would add 
a footnote to Technical Specification (TS) 4.8.1.1.2.e.8 which would 
permit performance of the 24-hour functional test of the emergency 
diesel generators (EDGs) during power operation. TS 4.8.1.1.2.e.8 
currently requires the 24-hour functional test of the EDGs be performed 
at least once per 18 months during shutdown; the proposed amendment 
would permit this testing to be performed during power operation 
provided the other two EDGs are operable. If either of the other two 
EDGs become inoperable, the test would be aborted.
    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 operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to permit the 24 hour functional test of the 
diesels to be performed during power operation does not increase the 
chances for a previously analyzed accident to occur. The function of 
the diesels is to supply emergency power in the event of a loss of 
offsite power. Operation of the diesels is not a precursor to any 
accident. Furthermore, the diesel generator being tested will remain 
operable and will be available to supply emergency loads within the 
required time. In addition, the two remaining diesel generators will be 
operable during the test. Consequently, if an offsite disturbance were 
to occur that affected the operability of the diesel being tested, the 
two remaining diesels would be capable of feeding the loads necessary 
for safe shutdown of the plant. This addresses the concerns raised in 
Information Notice 84-69 regarding the operation of emergency diesel 
generators connected in parallel with offsite power. In summary, the 
proposed changes do not adversely affect the performance or the ability 
of the diesel generators to perform their intended function.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed amendment to the 24 hour functional surveillance test 
will not affect the operation of any safety system or alter its 
response to any previously analyzed accident. The diesel will 
automatically transfer from the test mode if necessary to supply 
emergency loads in the requried time. The test mode is used for the 
monthly surveillance of the diesel generators as well, therefore, no 
new plant operating modes are introduced. In the event the diesel fails 
the functional test it will be declared inoperable and the actions 
required for an inoperable diesel will be performed. The remaining two 
diesel generators will be operable and are capable of feeding the loads 
necessary for safe shutdown of the plant.
    Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed amendment will not reduce availability of the diesel 
generator being tested to provide emergency power in the event of a 
loss of offsite power. If a loss of offsite power or a loss of coolant 
accident occurs during the surveillance test, the emergency bus would 
de-energize and shed load. The diesel generator would then transfer 
from the test mode to the emergency mode. It would then be available to 
automatically supply emergency loads. In addition, the two remaining 
generators will be operable during the test. Consequently, if an 
offsite disturbance were to occur that affected the operability of the 
diesel begin tested, the two remaining diesels would be capable of 
feeding the loads necessary for safe shutdown of the plant. The time 
required for the diesel being tested to pick up emergency loads will 
not be affected by performing the 24 hour functional test during power 
operation.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Ledyard B. Marsh.

Northeast Nuclear Energy Company et al.

Docket No. 50-336.

    Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut.
    Date of amendment request: October 18, 1994.
    Description of amendment request: The proposed amendment would 
require three type A overall Integrated Containment Leakage Tests be 
conducted at approximately equal intervals during shutdowns during each 
10-year service period. For the third Type A test for the second 10-
year period, it would be conducted during the thirteenth refueling 
outage extending the second 10-year service period to the end of the 
thirteenth refueling outage. The amendment would also change the 
Containment Leakage Bases by reflecting the conditions of a proposed 
exemption to 10 CFR 50, Appendix J, that would remove the requirement 
that the third Type A test for each 10-year period be conducted when 
the plant is shutdown for the 10-year plant inservice inspection.
    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 a SHC [significant 
hazards consideration] because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    Type A tests are performed to ensure that the total leakage from 
containment does not exceed the maximum allowable primary containment 
leakage rate at the design pressure. This ensures compliance with the 
dose limits of 10 CFR 100.
    The proposal to revise Surveillance Requirement 4.6.1.2.a of the 
Millstone Unit No. 2 Technical Specifications will increase the 
flexibility for scheduling the Type A tests. It does not modify the 
maximum allowable leakage rate at the design containment pressure, does 
not impact the design basis of the containment, and does not make any 
physical or operational changes to existing plant structures, systems, 
or components.
    The first two Type A tests of the second 10-year service period for 
Millstone Unit No. 2 have been conducted. The results of these tests 
demonstrate that Millstone Unit No. 2 has maintained control of 
containment integrity by maintaining margin between the acceptance 
criterion and the ``As-Found'' and ``As-Left'' leakage rates.
    Historically, Type A tests have a relatively low failure rate where 
Type B and C testing (local leakage rate tests) could not detect the 
leakage path. Most Type A test failures are attributed to failures to 
Type B or C components (containment penetrations and isolation valves). 
Type B and C components are tested per Surveillance Requirement 
4.6.1.2.d for the Millstone Unit No. 2 Technical Specifications. These 
tests are required to be conducted at intervals no greater than 24 
months, and the acceptance criterion for the combined leakage rate for 
all penetrations and valves subject to the Type B and C tests is 0.6 
La. These local leakage rate tests provide assurance that 
containment integrity is maintained. The relatively low ``As-Left'' 
Type B and C total leakage resulting from the past outage indicates 
that the leakage has been maintained within the technical specification 
acceptance criterion. The Type B and C tests will continue to be 
performed in accordance with the requirements of Surveillance 
Requirement 4.6.1.2.d. However, on September 26, 1994, NNECO submitted 
a request for a one-time technical specification change, request for 
enforcement discretion, and a request for a scheduler exemption from 
Appendix J to 10 CFR 50 regarding the Schedule for Type B and C 
testing. The NRC verbally granted enforcement discretion on September 
24, 1994, and written enforcement discretion on September 30, 1994. The 
schedular exemption request was granted on October 12, 1994.
    The previous Type A, B, and C tests demonstrate that Millstone Unit 
No. 2 has maintained control of containment integrity by maintaining a 
conservative margin between the acceptance criterion and the ``As-
Found'' and ``As-Left'' leakage results. Based on this, the Millstone 
Unit No. 2 containment is considered to be in sound condition. No 
operations are known to have occurred which would suggest any 
substantial degradation of these results.
    Based on the above, the proposal to revise Surveillance Requirement 
4.6.1.2.a of the Millstone Unit No. 2 Technical Specifications does not 
involve a significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposal to revise Surveillance Requirement 4.6.1.2.a of the 
Millstone Unit No. 2 Technical Specifications will increase the 
flexibility in scheduling the Type A tests. It does not make any 
physical or operational changes to existing plant structures, systems, 
or components. In addition, the proposal does not modify the acceptance 
criterion for the Type A tests. Maintaining the leakage through the 
containment boundary to the atmosphere within a specific value ensures 
that the plant complies with the requirements of 10 CFR 100. The 
containment boundary serves as an accident mitigator; it is not an 
accident initiator. Therefore, the proposal to revise Surveillance 
Requirement 4.6.1.2.a does not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The proposal to revise Surveillance Requirement 4.6.1.2.a of the 
Millstone Unit No. 2 Technical Specifications will increase the 
flexibility for scheduling the Type A tests. It does not modify the 
maximum allowable leakage rate at the design containment pressure, does 
not impact the design basis of the containment, and does not make any 
physical or operational changes to existing plant structures, systems, 
or components.
    The first two Type A tests of the second 10-year service period for 
Millstone Unit No. 2 have been conducted. The results of these tests 
demonstrate that Millstone Unit No. 2 has maintained control of 
containment integrity by maintaining margin between the acceptance 
criterion and the ``As-Found'' and ``As-Left'' leakage rates. 
Additionally, the results of the last Type B and C tests had 
significant margin with respect to the acceptance criterion. Based on 
the previous Type A, B, and C tests, the Millstone Unit No. 2 
containment is considered to be in sound condition. No operations are 
known to have occurred which would suggest any substantial degradation 
of these results.
    Based on the above, the proposal 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Learning 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: Phillip F. McKee.

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: September 30, 1994.
    Description of amendment request: The licensee has proposed to 
revise the Technical Specifications (1) to clarify the definition of 
core alterations, (2) to change the verbiage in the Limiting Condition 
For Operation (LCO) addressing the refueling operations, (3) to make 
changes to three surveillance requirements involving source range 
instrumentation, and (4) to change the LCO regarding the Residual heat 
Removal and coolant circulation water levels to be consistent with the 
guidance provided in NUREG-1431.
    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.
    Boron Dilution in Mode 6--A boron dilution in Mode 6 is precluded 
by technical specification requirements to close and lock all dilution 
source valves. There is a provision for dilution valves to be opened 
under administrative controls; in this case, cautionary measures will 
be taken to control and monitor the reactivity addition. Deletion of 
the source range analog operational test prior to core alterations will 
not impact an accident previously evaluated since the sources range 
monitors are verified operable prior to entry into Mode 6 and every 7 
days thereafter. The change in definition for a core alteration means 
that components which do not effect reactivity may be moved within the 
reactor vessel without any additional condition such as direct 
supervision of an SRO.
    Since a boron dilution would not be initiated by movement of 
nonfuel components within the reactor vessel, it is not impacted by the 
change in definition of a core alteration.
    Inadvertent Loading of a Fuel Assembly--Movement of a fuel assembly 
would be performed as a core alteration under the supervision of an 
SRO, therefore, it would not be impacted by the change to the 
definition of a core alteration. The change to the source range 
monitors also will not affect the probability of occurrence of a 
misloaded fuel assembly since this accident is precluded by 
administrative controls, as well as the source range monitors. Also, 
there will be no degradation in the reliability or accuracy of the 
source range monitors due to this change. The deletion of the 
requirement to perform the analog channel operational test within eight 
hours prior to core alterations will not impact performance of the 
monitors, since they have to be checked prior to entry into Mode 6 and 
every 7 days thereafter.
    Fuel Handling Accident--Movement of fuel will not affect this 
accident, because it will still be considered a core alteration. 
Therefore, there is no effect on the probability of a fuel handling 
accident. The source range monitors are not involved in the occurrence 
of a fuel handling accident. The fuel handling accident is the only 
accident considered here with radiological consequences. It will not be 
impacted by the proposed changes.
    Loss of RHR in Mode 6--The probability of this accident will not be 
changed since the new requirement is the same as before. As before, RHR 
may be secured for up to one hour per eight-hour period and boron 
dilution operations may not be performed with RHR secured (although 
this requirement is being added to the notes, the requirement is also 
given elsewhere in the technical specifications). Additionally, the 
existing reactor coolant system (RSC) temperature limits must still be 
met.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    All required systems will continue to operate as before. Therefore, 
there is no possibility of a new or different kind of accident. The 
deletion of the source range analog channel operational test within 
eight hours prior to core alterations will not affect the performance 
of the monitors since they will have had this test completed prior to 
entry into Mode 6 and every 7 days thereafter. The change in definition 
of a core alteration cannot create the possibility of a new type of 
accident because those initiating events for accidents will remain 
classified as core alterations.
    3. Involve a significant reduction in the margin of safety.
    The margin of safety for the above listed accidents will remain as 
before.
    a. Boron dilution in Mode 6--This accident calculates the time from 
receipt of a shutdown margin monitor dilution alarm until the core 
reaches criticality. Since this time is not changed, there is no 
reduction in the margin of safety. In this case, the dilution is 
precluded by administrative controls which will not be impacted by the 
proposed changes.
    b. Inadvertent Loading of a Fuel Assembly--Technical Specification 
3.9.1.1 protects against this accident by requiring sufficient boron in 
the RCS to prevent criticality for any core configuration including two 
stuck RCCAs [rod cluster control assemblies] in the fully withdrawn 
position. Since this requirement will not change, the margin of safety 
will not change.
    c. Fuel Handling Accident--The margin of safety for the 
radiological limits is not changed.
    d. Loss of RHR--Changes are editorial due to the revised definition 
of a core alteration. There is no change to the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Phillip F. McKee.

Northern States Power Company

Docket Nos. 50-282 and 50-306

    Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County, Minnesota.
    Date of amendment requests: October 3, 1994.
    Description of amendment requests: The proposed amendment would 
revise Prairie island Nuclear Generating Plant Technical Specification 
4.6, ``Periodic Testing of Emergency Power Systems.'' Specifically, the 
proposed amendment would modify the emergency diesel generator (EDG) 
24-hour load test requirements to provide a indicated load range of 
103-110% of the continuous rating. The proposed amendment would also 
rephrase various EDG test requirements to provide clarity and delete 
the requirement to verify that the auto-connected loads do not exceed 
3000 kw (Unit 2 5100kw).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    Changing the specification from ``unit'' to ``diesel generator'' 
does not change the intent of the specification, it merely clarifies 
the original intent and therefore cannot involve a change in the 
probability or consequences of an accident.
    Changing the 22-hour lower range limit from a load of 90% to an 
indicated load of 92% removes possible ambiguity from the specification 
but does not change the actual requirement, therefore it cannot involve 
a change in the probability or consequences of an accident.
    Removing the 22-hour upper range limit from the specification does 
not reduce the conservatism of the test since operating at a higher 
load provides more evidence of the ability of the machine to carry the 
accident loads. For this reason, this change will not involve any 
increase in the consequences of an accident. Also, increasing the load 
at which the diesel generator is tested cannot affect the probability 
of an accident.
    The NRC staff has pointed out, in Generic Letter 88-15, the hazards 
of testing the Diesel Generator at a load greater than the design 
rating. The proposed change is intended to ensure that the design 
rating is not inadvertently exceeded. Since the recent installation of 
two additional emergency diesel generators, the highest anticipated 
event loads are: Unit 1-2414kW, Unit 2-3813 kW. For these diesel 
generators, then, 103% of the continuous ratings:
     Unit 1, 103% of 2750 kW (continuous rating) = 2832.5 kW 
represents 117.3% of the highest anticipated event load and;
     Unit 2, 103% of 5400 kW (continuous rating) = 5562 kW 
represents 145.9% of the highest anticipated event load.
    A test load of 103%, therefore would still be significantly greater 
than the load required during accident conditions. Since an adequate 
level of electrical load carrying capacity of the diesel generators 
(and thus their accident mitigating functions) would still be 
demonstrated by the surveillance test, the consequences of an accident 
would be unaffected by the proposed change. The probability of 
occurrence of a previously evaluated accident would be unaffected since 
testing a diesel generator at load between 103 and 110 percent instead 
of at load between 105 and 110 percent could not cause or contribute to 
the initiation of an accident. For these reasons, this change could 
have no effect on the probability or consequences of an accident 
previously evaluated.
    Allowing momentary transients outside of the test band does not 
affect the conduct of the test, it merely allows momentary swing 
outside the specified band to not invalidate the test. Not allowing 
momentary transients would not prevent them, it would only require 
conducting the test longer until the specified time period was achieved 
without moving outside the band. Since the machine will not be operated 
any differently, this specification change cannot affect the 
probability or consequences of an accident previously evaluated.
    Proposed changes A, B, C, D, and the first part of E [identified as 
such in the submittal] are intended to clarify the meaning of the 
existing specifications without changing the requirements. For this 
reason, these proposed changes to the Technical Specifications will not 
change the manner in which the plant is operated or maintained. These 
administrative changes, therefore, will effect on the probability or 
consequences of an accident previously evaluated.
    The second part of E (verification of the bypass of diesel 
generator trips during a simulated safety injection signal vs 
concurrent safety injection and loss of offsite power signals) does not 
change the intended function which is to be tested but, rather, reduces 
the special conditions (temporary electrical jumpers to simulate the 
loss of offsite power) in which the plant needs to be placed in order 
to perform the test.
    Proposed change F (removal of the verification that the auto-
connected load do not exceed 3000 or 5100 kW) does not reduce the 
assurance of the ability of the diesel generators to perform the 
accident mitigation functions since this verification is performed by 
other, more pertinent, means.
    Therefore, these changes cannot increase the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a new 
or different king of accident from any accident previously analyzed.
    Changing the specification from ``unit'' to ``diesel generator'' 
does not change the intent of the specification, it merely clarifies 
the original intent and therefore cannot create the possibility of a 
new or different kind of accident.
    Changing the 22-hour lower range limit from a load of 90% to an 
indicated load of 92% removes possible ambiguity from the specification 
but does not change the actual requirement.
    Removing the 22-hour upper range limit from the specification does 
not change the manner in which the surveillance is performed. It only 
affects whether the time spent above 100% load can be counted toward 22 
hours in the 22-hour portion of the test. This change would not allow 
any new modes of operation nor does it allow any modification to the 
plant.
    As stated above, testing a diesel generator at a load between 103 
and 110% instead of between 105 and 110% could not cause or contribute 
to the initiation of an accident.
    Allowing momentary transients outside of the test band does not 
affect the conduct of the test, it merely allows momentary swings 
outside the specified band to not invalidate the test. Not allowing 
momentary transients would not prevent them, it would only require 
conducting the test longer until the specific time period was achieved 
without moving outside the band.
    Therefore, for these reasons, operation of the facility in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
analyzed.
    As stated above [for changes A-F], the proposed changes will not 
cause a change in the way in which the plant is operated or maintained, 
excepted for the reduction of the special conditions in which the plant 
needs to be placed in order to test the bypass of the diesel generator 
trips. Therefore, these administrative changes will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. The proposed amendment will not involve a significant reduction 
in a margin of safety.
    Changing the specification from ``unit'' to ``diesel generator'' 
does not change the intent of the specification, it merely clarifies 
the original intent and therefore cannot affect the margin of safety.
    Changing the 22-hour lower range limit from a load of 90% to an 
indicated load of 92% removes possible ambiguity from the specification 
but does not change the actual requirement and therefore cannot affect 
the margin of safety.
    The margin of safety is not affected by removal of the 22-hour 
upper range limit on the operation of the diesel generators during 
surveillance testing since the margin of safety is related to the 
magnitude of the accident loads and the maximum capacity of the machine 
to carry load and this margin would be unaffected by this change.
    The capacity of each diesel generator to carry electrical load can 
not be diminished by being tested at a lower load. Also, load testing 
to less than 105% but more than 103% does not lessen the confidence in 
the ability of the diesel generators to carry adequate load for this 
facility since these diesel generators have significantly greater load 
capacity than required by Standard Review Plan guidance in this regard 
(the guidance allows peak accident load up to 100% of the continuous 
rating versus Unit 1 diesel generators peak accident load of 87.8% and 
Unit 2 diesel generators peak accident load of 70.6%). Therefore, this 
change will not involve a significant reduction in the margin of 
safety.
    Allowing momentary transients outside of the test band does not 
affect the conduct of the test, it merely allows momentary swings 
outside the specified band to not invalidate the test. Not allowing 
momentary transients would not prevent them, it would only require 
conducting the test longer until the specified time period was achieved 
without moving outside the band. Since the machine will not be operated 
any differently per the new specification, the margin of safety is 
unaffected.
    As stated above [for changes A-F], the proposed changes will not 
cause a change in the way in which the plant is operated or maintained, 
except for the reduction of the special conditions in which the plant 
needs to be placed in order to test the bypass of the diesel generator 
trips. Therefore, these administrative 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 requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

Omaha Public Power District

Docket No. 50-285

    Fort Calhoun Station, Unit No. 1, Washington County, Nebraska.
    Date of amendment request: October 7, 1994.
    Description of amendment request: The proposed amendment to the 
Technical Specifications (TSs) would (1) delete the surveillance 
requirements contained in TS 3.6(3)a for the raw water backup valves to 
the containment cooling coils, (2) delete the surveillance requirements 
contained in TS 3.2, Table 3-5, item 6, for raw water valves, and (3) 
revise the basis of TS 2.4 to reflect these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The deletion of surveillance requirements contained in Technical 
Specifications (TS) 3.2, Table 3-5, Items 6 and 3.6(3)a does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    TS 3.6(3)a requires the Raw Water (RW) backup valves to the 
containment air coolers to be tested each refueling outage. In 1990, 
during the process of reviewing several open items created by the 
design basis reconstitution project, an engineering analysis determined 
that RW direct cooling of the containment air cooling coils should not 
be used after an accident that has created elevated temperature 
conditions inside containment. The high containment air temperatures, 
in conjunction with the low back pressure in the containment cooling 
coils when in the RW direct cooling mode, introduces the possibility of 
vaporization inside the coils. Therefore, the use of RW direct cooling 
for the containment air coolers has been discontinued in post-Loss of 
Coolant Accident (LOCA) or post-Main Steam Line Break (MSLB) 
situations. The issue of not being able to utilize RW direct cooling to 
the containment air cooling coils was reported to the NRC in LER-90-25, 
dated October 29, 1990 and LER-90-25 Revision 1, dated December 17, 
1990.
    Raw water direct cooling of the containment air coolers is possible 
if the containment atmospheric temperatures are less that 150 deg.F. If 
RW direct cooling of the containment air coolers was utilized after a 
LOCA or MSLB accident, it could only be used for long-term containment 
atmospheric cooling. These conditions are essentially equivalent to 
that associated with conditions in containment during normal plant 
operation. RW direct cooling of the containment air coolers is not a 
required post-accident function to maintain containment pressure below 
60 psig. Since these valves are not required to perform a post-accident 
function, deletion of the requirements to test these valves does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    TS 3.2, Table 3-5, Item 6 requires that valves in the RW system be 
tested every refueling outage. The valves tested by this surveillance 
that could perform a safety function are already tested in accordance 
with TS 3.3(1). Therefore testing of these valves under TS 3.2, Table 
3-5, Item 6 is redundant to TS 3.3(1)a.
    (2) The proposed changes do not create the possibility of a new or 
different kind of accident from any previously analyzed.
    There will be no physical alterations to the plant configuration, 
changes to setpoint values, or changes to the implementation of 
setpoints or limits as a result of this proposed change. Valves that 
are required to be repositioned during an accident to mitigate the 
consequences will still be tested on a refueling frequency. The 
proposed change only deletes unnecessary or redundant testing 
requirements from the TS. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from any 
previously analyzed.
    (3) The proposed changes do not involve a significant reduction in 
a margin of safety.
    The proposed changes delete unnecessary or redundant surveillance 
requirements within the TS. The deletion of TS 3.2, Table 3-5 Item 6, 
only deletes testing requirements that are already required to be 
conducted by TS 3.3(1)a. The deletion of the requirement to test the RW 
backup valves to the containment air coolers in TS 3.6(3) only deletes 
an unnecessary surveillance. RW direct cooling of the containment air 
coolers is not required to maintain containment pressure below the 
design limit of 60 psig. Therefore, the proposed changes do not involve 
a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875 
Connecticut Avenue, N.W., Washington, D.C. 20009-5728.
    NRC Project Director: Theodore R. Quay.

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 September 26, 1994.
    Description of amendment request: The amendment would remove the 
requirement for operability of the Average Power Range Monitors (APRMs) 
while the plant is in Operational Condition 5. However, the requirement 
for the APRMs to be operable during a shutdown margin demonstration, 
when the mode switch is in Startup, will remain unchanged.
    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 
construction, which is presented below:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Not requiring APRMs to be OPERABLE in OPCON 5 will not increase the 
probability of inadvertent reactor critically during refueling 
operations. Refueling Interlocks, NMS [Neutron Monitoring System] (SRMs 
[Source Range Monitor], IRMs [Intermediate Range Monitor]), and 
procedural restrictions provide assurance that inadvertent criticality 
does not occur due to the simultaneous withdrawal or removal of two 
control rods or due to the inadvertent insertion of a fuel bundle into 
a core location with a control blade removed.
    The FSAR [Final Safety Analysis Report] Section 15.4.1 discusses 
the potential for a control rod withdrawal error during refueling and 
start-up operations. The discussion concludes that the withdrawal of 
one control rod does not require a safety action because the total 
worth of one control rod is not sufficient to cause criticality. The 
attempted withdrawal of two control rods, assuming an operator error 
and a single active failure, would result in a control rod block 
initiated by the Refueling Interlocks. The safety-related IRM 
subsystem, which is required by Technical Specifications to be OPERABLE 
while in OPCON 5, is designed to generate a rod block or reactor scram 
on high neutron flux and is therefore a backup protective system for 
the Refueling Interlocks during refueling.
    The Safety-related IRM subsystem of the NMS is required by 
Technical Specifications to be OPERABLE during OPCON 5 to support the 
safety design bases of the NMS and RPS [Reactor Protection System]. The 
SRM is not a safety-related subsystem but is important to plant safety 
and is required by Technical Specifications to be OPERABLE in OPCON 5. 
The SRM subsystem provides the plant operator with neutron flux levels 
from startup conditions to the IRM operating range. The SRMs and IRMs 
are designed to respond to local core conditions and would indicate and 
respond (control rod block or scram) to an accident condition to 
mitigate the transient. Thus, the APRMS are not necessary to be 
OPERATOR in OPCON 5. The proposed Technical Specification change will 
not alter the current requirements that the APRMs be OPERABLE during 
shutdown margin demonstrations in OPCON 5 when the mode switch is in 
Startup.
    The proposed Technical Specification change would reduce the APRM 
operability requirement in OPCON 5 and would not affect the FSAR 
evaluation of the inadvertent criticality due to the withdrawal or 
removal of the highest worth control rod or due to the insertion of 
fuel bundles in uncontrolled cells. The FSAR concludes that the 
Refueling Interlocks and plant procedures provide assurance that 
inadvertent criticality does not occur during refueling.
    The consequences of an accident will not be increased by the 
proposed Technical Specification change because of the existing lines 
of defense which prevent an inadvertent criticality event during 
refueling, e.g., administrative restrictions, refueling procedures, 
licensed plant operators, SRMs, Refueling Interlocks, and IRMs. 
Furthermore, should the number of operator IRM or SRM channels be less 
than that required by Technical Specifications, the Technical 
Specifications require that core alteration activities be suspended and 
all insertable control rods be inserted into the core.
    Therefore, the proposed changes do not result in an increase in 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 changes to the Technical Specifications will remove 
the APRM operability requirement while in OPCON 5 (except for shutdown 
margin demonstration testing); however, the SRMs and IRMs will still be 
required to be OPERABLE in OPCON 5.
    The IRMs are safety-related and are designed to detect and respond 
to increases in neutron flux within the local core regions. Any 
inadvertent increases in neutron flux during refueling would originate 
at a local core location, i.e., rod withdrawal error or fuel bundle 
insertion. Technical Specifications require IRM operability and will 
generate an RPS scram or control rod block if neutron flux increased to 
the setpoint. Therefore, removing the APRMs operability requirement in 
OPCON 5 would not effect any safety related equipment or equipment 
important to safety.
    The APRMs provide core power information to the control room 
operator and also provide trip signals to the RMCS [Reactor Manual 
Control System] and RPS as required. The absence of an APRMs input 
signal will not affect these systems during refueling operations.
    Removing the APRMs operability in OPCON 5 will not affect the 
response of safety-related equipment as previously evaluated in the 
FSAR. The proposed changes to the Technical Specifications do not 
affect any safety-related equipment or equipment important to safety.
    The proposed changes to the Technical Specifications would remove 
the APRMs operability requirement during refueling operations. 
Technical Specifications require IRM operability and will generate an 
RPS scram or control rod block if neutron flux increased to the 
applicable setpoint.
    No new types of accidents would be introduced since the SRMs and 
IRMs are available and required to be OPERABLE in OPCON 5. Both SRMs 
and IRMs would indicate and provide a control rod block or scram 
signal, as appropriate, to an increase in neutron flux to mitigate a 
transient event. Furthermore, should the number of OPERABLE IRM or SRM 
channels be less than that required by Technical Specifications, the 
Technical Specifications require that core alteration activities be 
suspended and all insertable control rods be inserted into the core.
    Therefore, the proposed Technical Specification changes do 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.
    For the reasons discussed in items 1 and 2 above and because the 
Technical Specification Bases do not discuss or require APRMs 
operability during OPCON 5, Refueling, the proposed Technical 
Specification changes do not involve a significant reduction in a 
margin of safety.
    The NRS staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. Local 
Public Document Room location: Osterhout Free Library, Reference 
Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701 
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Philadephia Electric Company

Docket Nos. 50-352 and 50-353

    Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania.
    Date of amendment request: August 22, 1994.
    Description of amendment request: The amendment consists of five 
(5) sections of Technical Specifications changes which reflect the 
Improved Standard Technical Specifications (NUREG-1433):

Section 1: Control Rod Block Instrumentation,
Section 2: Standby Liquid Control System Operability in Mode 5,
Section 3: Scram Discharge Volume Valve Testing,
Section 4: Optional Method of Scram Timing, and
Section 5: Definition of Core Alteration.

    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:
Section 1: Control Rod Block Instrumentation
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed TS changes can be divided into two general categories, 
the deletion of the ``S/U'' requirements, and the change in frequency 
of the SRM [Source Range Monitor] and IRM [Intermediate Range Monitor] 
Calibration and Functional Tests. In each case in which the ``S/U'' 
requirement has been deleted, the normal surveillance frequency 
specified for the required Operating Condition remains. The equipment's 
associated probability of failure remains unchanged. In the case of the 
surveillance frequency changes proposed for the SRMs and IRMs, the 
probability of an accident evaluated in the SAR [Safety Analysis 
Report] occurring does not increase since there is no credit taken in 
the SAR for those Control Rod Block functions with respect to an 
accident. As such, the proposed changes will not result in an increase 
in the probability of occurrence of an accident previously evaluated in 
the SAR. The proposed TS changes do not alter the method of operation 
or performance of the equipment in carrying out associated Control Rock 
Block functions. Thus, the consequences of an accident previously 
evaluated in the SAR are not increased.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed TS changes do not alter the configuration of the plant 
or the way that the plant is operated. The equipment can perform no 
other function than it is presently capable of, or cause or permit any 
other accident than is now possible. Thus, the possibility of an 
accident of a different type than previously evaluated in the SAR 
cannot be created.
    Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    Since the proposed TS changes affect only the surveillance 
frequency intervals and do not change the plant configuration or 
associated instrument setpoints, there is no quantitative or 
qualitative reduction in the margin of safety. Thus, the margin of 
safety as defined in the bases of any Technical Specification is not 
reduced.
    Therefore, the proposed TS changes do not involve a reduction in a 
margin of safety.
Section 2: Standby Liquid Control System Operability in Mode 5
    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed TS change will remove the SLCS operability requirement 
in OPCON 5. The purpose of the SLCS is to bring the reactor to and 
maintain it in a cold shutdown condition from normal power operations 
following failure to scram during power operations. Initiation of the 
SLCS is not a precursor to any accident. Therefore, inoperability of 
the SLCS in OPCON 5 cannot increase the probability of an accident 
previously evaluated.
    The proposed TS change does not involve a physical change in any 
system's configuration and no new modes of operation are introduced. 
The SLCS has not analyzed function OPCON 5. The probability of fuel 
failure will not be increased by this change. Shutdown margin, in 
conjunction with TS requirements and procedural controls, will assure 
that an inadvertent criticality event will not occur during refueling. 
In addition, the Reactor Protection System (RPS) and Control Rod System 
will provide protection in the unlikely event that an inadvertent 
criticality should occur.
    Therefore, the proposed TS change does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed TS changes does not involve a physical change in any 
system's configuration and no new modes of operation are introduced. 
The SLCS's only purpose is to mitigate the consequences of a failure to 
scram during power operation. In OPCON 5, the SLCS has no analyzed 
function, therefore, the proposed TS change will not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    The purpose of the SLCS is to bring the reactor to and maintain it 
in a cold shutdown condition from normal power operations following a 
failure to scram during power operations. The SLCS is not designed to 
terminate an inadvertent criticality during OPCON 5. Shutdown margin, 
either demonstrated or analytically determined, in conjunction with 
Technical Specifications and procedural controls, will assure that an 
inadvertent criticality event will not occur during refueling 
operations. In addition, the RPS and Control Rod System, which are 
extremely reliable, will provide protection in the unlikely event that 
an inadvertent criticality does occur. Therefore, the proposed TS 
change does not involve a reduction in a margin of safety.
Section 3: Scram Discharge Volume Valve Testing
    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The Scram Discharge Volume (SDV) is not an accident initiator. 
Deletion of the requirement that the SDV be determined OPERABLE by 
testing the SDV vent and drain valves when control rods are scram 
tested from a normal control and configuration of less than or equal to 
50% rod density at least once per 24 months, as proposed, will have no 
effect on the probability or consequences of an accident previously 
evaluated.
    This proposed TS will have a negligible impact on the conditions 
experienced by the vent and drain valves as they stroke closed, since 
the SDV is initially vented to the atmosphere, and the valves close 
before the SDV becomes pressurized, even during a scram at full reactor 
power. Reactor pressure and Control Rod Drive (CRD) discharge flow 
conditions do not influence the SDV vent and drain closure rates, since 
the SDV is of sufficient volume and initially vented such that peak 
pressure prior to the SDV complete isolation will not be substantial. 
In addition, lower coolant temperatures expected during testing at 
shutdown conditions will also have a negligible impact on the 
performance of the test. Although, there could be some variation in the 
performance [of] the SDV vent and drain valves to re-open when 
performing the test during shutdown conditions, as opposed to 
conducting the test during power operation, the ability of the valves 
to re-open is demonstrated after each reactor scram during power 
operation.
    In the event and SDV vent or drain valve failed to open, increasing 
SDV level during reactor operation would cause 1) an alarm in the Main 
Control Room (MCR), 2) a control rod block, and finally a reactor scram 
initiated by the Reactor Protection System (RPS) if action is not taken 
to drain the SDV. Therefore, the ability to shut down the reactor is 
not impaired. If a SDV vent or drain valve fails to close, the 
redundant valve's closure would provide the required function. If both 
valves failed to close, a loss of reactor coolant in the form of water 
discharged from the CRD system would occur. The amount of water 
discharged will be relatively small, and is more of a concern from the 
standpoint of contamination to the Secondary Containment rather than a 
loss of reactor water inventory. A structural failure of the SDV, which 
bounds this case of an open SDV vent or drain line, has been previously 
evaluated in NUREG-0808, ``Generic Safety Evaluation Report Regarding 
Integrity of BWR Scram System Piping.'' In this evaluation, the NRC 
concluded that, for a bounding leakage case corresponding to a rupture 
of the SDV, the offsite doses would be well within the limits of 
10CF100, and that adequate core cooling would be maintained.
    Deletion of the requirement that the SDV be determined OPERABLE by 
testing the SDV vent and drain valves, as proposed in this TS Change 
Request, will have an insignificant effect on the probability of 
occurrence of malfunction of any plant equipment. The conditions in the 
SDV at the time of vent and drain valve closure are not appreciably 
different whether a scram is initiated from power operation or during 
shutdown conditions. In addition, this proposed TS change eliminates 
the potential need for an additional startup and shutdown cycle, along 
with the associated challenges to all systems and components, that 
would be required to satisfy the current TS requirements in the event a 
unit were to trip off-line shortly before a planned outage when the 
surveillance was scheduled to be performed. Furthermore, this proposed 
TS changes does not affect the testing frequency for the valves.
    This proposed TS change will not result in appreciably different 
conditions experienced by the valves as they close, and their ability 
to re-open is confirmed following each reactor scram from power 
conditions. The consequences resulting from a failed closed or failed 
open SDV vent or drain line have been evaluated and determined not to 
result in offsite doses that would exceed the limits specified in 
10CFR100, or jeopardize adequate reactor core cooling capability. 
Therefore, the consequences of a malfunction of equipment important to 
safety previously evaluated is not increased.
    Therefore, the proposed TS change does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The SDV is not an accident initiator. Deletion of the requirement 
that the SDV be determined OPERABLE by testing the SDV vent and drain 
valves from a configuration of less than or equal to 50% rod density, 
as proposed, will not create the possibility of a different type [of] 
accident than any previously evaluated.
    No plant equipment is added or deleted as a result of this proposed 
change. Since the initial conditions of pressure, temperature, and CRD 
system discharge flowrate have no appreciable effect on the SDV vent 
and drain valve performance, no different type of malfunction of any 
equipment important to safety is created.
    Therefore, the proposed TS change does not create the possibility 
of a new different kind of accident from any previously evaluated.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    Since the initial test conditions of pressure, temperature, and CRD 
discharge flowrate will have no appreciable effect on the SDV vent and 
drain valve performance, conducting the surveillance test during 
shutdown conditions, as specified in this proposed TS change, will not 
affect the validity of the surveillance results with respect to the 
operability of the SDV to perform its intended safety function. 
Furthermore, every reactor scram is a serious plant transient and a 
potential challenge to safety-related systems and equipment. The 
potential decrease in future scrams which could result from this 
proposed TS change will represent an improvement in overall safety.
    Therefore, the proposed TS change does not involve a reduction in a 
margin of safety.
Section 4: Optional Method of Scram Timing
    1. The proposed Technical Specification (TS) changes involves a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    Scram testing control rods at zero reactor coolant pressure will 
not increase the probability of any control rod related transient or 
accident discussed in the UFSAR [Updated Final Safety Analysis Report]. 
UFSAR Sections 15.4.1.1 and 15.4.1.2 discuss the consequences of 
inadvertent reactivity insertion errors due to the withdrawal of one or 
more control rods. The probability of one of these events occurring is 
a function of operator error and equipment malfunction and is not 
related to scram insertion times.
    An inadvertent reactivity insertion error is prevented by existing 
system hardware interlocks and procedural controls that are not 
affected by scram time testing, e.g., core design, control and design, 
one-rod-out interlocks, refueling interlocks, control rod sequence 
designations, and neutron monitoring systems.
    USFAR Section 15.4.9 discusses the control rod drop accident 
(CRDA). The CRDA assumes that a control rod suddenly drops out of the 
core due to equipment malfunction. The probability of occurrence of 
this accident is based on an equipment malfunction and is not affected 
by scram testing.
    Engineering analysis and control rod scram test data demonstrate 
that a control rod drive that will meet the 2.0 second, scram insertion 
time, test criteria at zero reactor coolant pressure will also meet all 
scram insertion criteria during reactor startup and up to 40% rated 
thermal power.
    The 2.0 second criterion was chosen to conservatively envelop scram 
time criteria and reactivity insertion criteria during reactor startup 
and up to 40% rated power conditions. Therefore, scram testing affected 
control rods at zero reactor pressure will not increase the 
consequences of an accident previously evaluated.
    UFSAR Sections 15.4.1.1 and 15.4.1.2 evaluate reactivity insertion 
transients at low power conditions due to inadvertent control rod 
withdrawal errors. The UFSAR concludes that rod withdrawal errors at 
low power are adequately precluded by refueling interlocks, rod worth 
minimizer, operating procedures, core design, and control rod hardware 
design. However, should operator errors followed by equipment 
malfunctions result in an inadvertent criticality event, the IRMs would 
provide the necessary rod blocks or reactor scram to preclude the 
operational transient. Scram insertion time limits for the continuous 
rod withdrawal error during startup is 5.0 seconds. This scram time 
criterion will be met by a control rod that scrams within 2.0 seconds 
at zero reactor pressure. The 2.0 second scram criterion was 
established to ensure that affected control rods will meet scram 
requirements from zero reactor pressure up to 40% core thermal power.
    Also, during low power operation (UFSAR Subsection 15.4.1.2) the 
rod worth minimizer (RWM) prevents the operator from selecting and 
withdrawing an out-of-sequence control rod. During reactor operation in 
the power range (UFSAR subsection 15.4.2) the rod block monitor (RBM) 
prevents a rod withdrawal error by inhibiting inadvertent control rod 
withdrawal. The RWM and RBM do not rely on a scram function to 
mitigation the consequences of a rod withdrawal error, and therefore 
the consequences of an accident evaluated in the UFSAR will not be 
affected by the proposed changes to the Technical Specifications.
    The consequences of a control rod drop accident (UFSAR Section 
15.4.9) would not be affected by scram testing a control rod at zero 
reactor pressure. The design basis accident of the rod drop accident 
assumes that control rods scram within 5.0 seconds. This 5.0 second 
scram test requirement will be met by control rods that meet the 2.0 
second criterion at zero reactor pressure.
    Therefore, the proposed TS changes do not involve an increase in he 
probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The changes to the Technical Specifications will allow control rods 
to be scram tested at zero reactor pressure and then again at rated 
reactor pressure prior to achieving 40% rated reactor power. No new 
types of accidents will be introduced since control rods that meet the 
2.0 second scram criterion at zero reactor pressure will also meet all 
scram test criteria during reactor startup and at rated reactor 
pressure.
    Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The basis for shutdown margin (TS Bases 3/4.1.1) states that the 
reactor shall be made subcritical by all certain margin in all 
operating and shutdown conditions. The proposed changes to the 
Technical Specifications will not affect the shutdown margin 
requirements. Adequate shutdown margin is assured by core design, the 
one-rod-out interlock, and administrative controls.
    The basis for the control rod insertion times (TS Bases 3/4.1.3) 
states that the scram times are to be consistent with those used in the 
transient and accident analysis. The proposed Technical Specifications 
changes will add an additional scram test verification for affected 
control rods at zero reactor pressure. The zero reactor pressure scram 
limit (2.0 seconds) was designed to ensure that the scram times assumed 
in the transient analysis will remain bounding from zero reactor 
pressure up to 40% rated core thermal power.
    The basis for the control rod drop accident (TS Bases 3/4.1.3) 
states that the potential effects of a CRDA are limited. The proposed 
Technical Specifications changes will not effect the control rod drop 
results as the changes do not affect the reactivity of the rod or the 
rod drop velocity. The CRDA analysis is based on a 5.0 second scram 
insertion time criterion. The 2.0 second time criterion was established 
to ensure that the 5.0 second scram time criterion was valid from zero 
reactor pressure to 950 psig reactor pressure.
    The basis for MCPR limits (TS Bases 3/4.1.3 and 2.3) states the CRD 
system must bring the reactor subsubcritical at a rate fast enough to 
prevent MCPR from becoming less than the fuel cladding safety limit 
during the limiting power transient analyzed in the UFSAR. The proposed 
changes to the Technical Specifications will not affect the scram 
insertion rates that are used as input to the transient analysis. The 
zero reactor pressure scram limit of 2.0 seconds was developed to 
ensure that the control rods would meet their design scram insertion 
times from zero reactor pressure up to 40% rated power.
    The proposed changes to the Technical Specifications will not 
increase the probability of inadvertent criticality because the changes 
do not affect the reactivity worth of control rods.
    Therefore, the proposed TS changes do not involve a reduction in a 
margin of safety.
Section 5: Definition of Core Alteration
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed definition change removes the requirement to have a 
SRO or LSRO supervise control rod withdrawal in an off-loaded cell 
(i.e. no fuel assemblies). The evaluated accident potentially affected 
by this change is a control rod movement error during refueling 
resulting in inadvertent criticality. The supervision by a SRO or LSRO 
does not solely preclude inadvertent criticality and was not relied 
upon in the accident analysis contained in Section 15.4 of the LGS 
Updated Final Safety Analysis Report (UFSAR). The LGS reactor core is 
designed to have adequate shutdown margin with the highest-reactivity-
worth control rod withdrawn. The withdrawal of a second rod with fuel 
assemblies loaded in the associated cell is prevented by a combination 
of the refueling, one-rod-out interlock, and the Limiting Conditions 
for Operation (LCO) requirement of TS 3.9.10.2. The LCO requirements 
ensure adequate shutdown margin is present prior to control rod 
withdrawal. This is accomplished by testing during startup following a 
refueling outage or by analytical calculations during refueling. The 
refueling interlock will provide a rod block upon an attempt to 
withdraw a second control rod and is required to be operable in 
accordance with TS 3.9.10.2 except for rods which have no fuel 
assemblies in the associated cell. The removal of the fuel assemblies 
from a cell eliminates the need for the reactivity control function of 
the associated rod. The physical removal of a control blade from the 
core by means of the refueling floor, first requires the removal of the 
four associated fuel assemblies in the cell. This design inherently 
prevents inadvertent criticality. Finally, this change is consistent 
with NUREG-1433 ``Standard Technical Specifications.'' Since current 
analysis permits the withdrawal of a control rod blade, provided the 
associated cell is unloaded, and refueling mode interlocks, 
administrative TS requirements and the physical design of the control 
blade and fuel cell, which preclude inadvertent criticality, will 
remain unchanged, this proposed change to the TS definition of CORE 
ALTERATION will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The LGS UFSAR currently permits control rod withdrawal and or 
removal, provided there are no fuel assemblies in the associated fuel 
cell. The definition change removes the requirement to have a SRO or 
LSRO supervise rod withdrawal in an off-loaded cell. The change 
potentially [a]ffects a control rod movement error during refueling 
resulting in inadvertent criticality which has been previously 
evaluated. In addition, the proposed change will make no physical 
changes to equipment or remove administrative controls which solely 
preclude inadvertent criticality. Therefore, this change will not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    The LGS TS bases address reactivity concerns, radiological 
releases, control rods, and monitoring of the facility related to this 
change. With the four fuel assemblies removed from a cell, the control 
rod/blade in the associated cell has no reactivity function. The 
reactivity issues addressed by TS are therefore unaffected. The rod/
blade coupling integrity is maintained by the requirement to perform a 
coupling check following maintenance. Section 15.4 of the UFSAR states 
that there are no radiological releases in association with a rod 
withdrawal error during refueling. This conclusion is maintained by the 
administrative requirements of TS 3.9.10.2, the refueling interlocks 
for one-rod-out, and the physical design of the blade and cell. Lastly, 
the TS requirements for Emergency Core Cooling, Plant System, 
Containment, and Electrical Power Distribution System, which provide 
the systems necessary to mitigate the effects of a radiological release 
during control rod movement in an unloaded cell were reviewed and were 
found not to be adversely [a]ffected by the proposed change. Therefore, 
this change will not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: John F. Stolz.

Philadelphia Electric Company

Docket Nos. 50-352 and 50-353

    Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania.
    Date of amendment request: August 31, 1994.
    Description of amendment request: The proposed amendments, which 
are consistent with the Improved Standard Technical Specifications 
(NUREG-1433), involve the following six (6) sections of TS changes:

Section 1: Relocation of Turbine Overspeed Protection System 
Requirements;
Section 2: Relocation of Primary Containment Conductor Protection 
Devices Requirements;
Section 3: Feedwater/Main Turbine Trip System Actuation Instrumentation 
Requirements;
Section 4: Permit Operability of Low Pressure Coolant Injection While 
Aligned to Shutdown Cooling;
Section 5: Remove Temperature Requirement for Operational Condition 
[OPCON] 5; and
Section 6: Reduce Frequency of Alternate Decay Heat Demonstration

    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:
Section 1: Relocation of Turbine Overspeed Protection System 
Requirements
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed change relocates requirements from the TS, to licensee 
controlled documents. The licensee controlled documents containing the 
relocated requirements will be maintained using the provisions of 10 
CFR 50.59 and are subject to the change control process in the 
Administrative Controls Section 6.0 of the TS. Since changes to 
licensee controlled documents will be evaluated per 10 CFR 50.59, no 
increase (significant or insignificant) in the probability or 
consequences of an accident previously evaluated will be allowed. 
Therefore, this change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident previously evaluated.
    This change relocates requirements to licensee controlled 
documents. This change will not alter the plant configuration (no new 
or different type of equipment will be installed) or make changes in 
methods governing normal plant operation. This change will not impose 
different requirements and adequate control of information will be 
maintained. This change will not alter assumptions made in the safety 
analysis and licensing basis. Therefore, this change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    This change relocates requirements from the TS to licensee 
controlled documents. This change will not reduce a margin of safety 
since it has no impact on any safety analysis assumptions. In addition, 
the requirements to be transferred from the TS to licensee controlled 
documents are the same as the existing Technical Specifications. Since 
any future changes to these licensee controlled documents will be 
evaluated per the requirements of 10 CFR 50.59, no reduction 
(significant or insignificant) in [a] margin of safety will be allowed. 
Therefore, this change will not involve a significant reduction in a 
margin of safety.
    The existing requirements for NRC review and approval of revisions, 
in accordance with 10 CFR 50.59, to these details and requirements 
proposed for relocation, does not have a specific margin of safety upon 
which to evaluate. However, since the proposed change is inconsistent 
with the BWR [boiling-water reactor] Improved Standard Technical 
Specifications (NUREG-1433 approved by the NRC Staff) and the change 
controls for proposed relocated details and requirements provide an 
equivalent level of regulatory authority, revising the TS to reflect 
the approved level of detail and requirements ensures no reduction to 
the margin of safety.
Section 2: Relocation of Primary Containment Conductor Protection 
Devices Requirements
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    This proposed change relocates requirements from the TS to licensee 
controlled documents. The licensee controlled documents containing the 
relocated requirements will be maintained using the provisions of 10 
CFR 50.59 and are subject to the change control process in the 
Administrative Controls Section 6.0 of the TS. Since changes to these 
licensee controlled documents will be evaluated per 10 CFR 50.59, no 
increase (significant or insignificant) in the probability or 
consequences of an accident previously evaluated will be allowed. 
Therefore, this change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    This change relocates requirements to licensee controlled 
documents. This change will not alter the plant configuration (no new 
or different type of equipment will be installed) or make changes in 
methods governing plant operation. This change will not impose 
different requirements and adequate control of information will be 
maintained. This change will not alter assumptions made in the safety 
analysis and licensing basis. Therefore, this change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    This change relocates requirements from the TS to licensee 
controlled documents. This change will not reduce a margin of safety 
since it has no impact on any safety analysis assumptions. In addition, 
the requirements to be transferred from the TS to the licensee 
controlled documents are the same as the existing TS. Since any future 
changes to these licensee controlled documents will be evaluated per 
the requirements of 10 CFR 50.59, no reduction (significant or 
insignificant) in [a] margin of safety will be allowed. Therefore, this 
change will not involve a significant reduction in a margin of safety.
    The existing requirements for NRC review and approval of revisions, 
in accordance with 10 CFR 50.59, to these details and requirements 
proposed for relocation, does not have a specific margin of safety upon 
which to evaluate. However, since the proposed change is consistent 
with the BWR Improved Standard TS (NUREG-1433 approved by the NRC 
Staff) and the change controls for proposed relocated details and 
requirements provide an equivalent level of regulatory authority, 
revising the TS to reflect the approved level of detail and 
requirements ensures no reduction to the margin of safety.
Section 3: Feedwater/Main Turbine Trip System Actuation Instrumentation 
Requirements
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    For the proposed TS change, in the event of a Reactor Vessel Water 
Level--High Level 8 transient, operator action per existing plant 
procedures would terminate the event and prevent damage to the Main/RFP 
[reactor feed pump] Turbine due to water carry over. The Main/RFP 
Turbine do not serve a safety function, also at <25% [rated="" thermal="" power]="" rtp="" a="" level="" 8="" transient="" event="" will="" not="" cause="" a="" reactor="" scram.="" an="" analysis="" of="" information="" in="" the="" bases="" for="" aplhgr="" [average="" planar="" linear="" heat="" generation="" rate]="" and="" mcpr="" [minimum="" critical="" power="" ratio]="" has="" shown="" that="" a="" sufficient="" margin="" to="" core="" safety="" limit="" exist,="" so="" fuel="" integrity="" levels="" are="" not="" violated.="" therefore,="" the="" proposed="" ts="" change="" does="" not="" involve="" an="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" 2.="" the="" proposed="" ts="" change="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" should="" the="" feedwater/main="" turbine="" trip="" system,="" reactor="" vessel="" water="" level-high="" level="" 8,="" not="" actuate="" in="" opcon="" 1="" at=""><25% rtp,="" operator="" action="" per="" existing="" plant="" startup="" procedures="" would="" protect="" the="" main/rfp="" turbines.="" if="" operator="" action="" is="" not="" performed,="" damage="" to="" balance="" of="" plant,="" non-safety="" related="" equipment="" could="" occur.="" high="" reactor="" vessel="" water="" level="" is="" not="" a="" concern="" for="" reactor="" core="" safety="" at=""><25% rtp.="" therefore,="" the="" proposed="" ts="" change="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" 3.="" the="" proposed="" ts="" change="" does="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" proposed="" ts="" change,="" which="" revises="" the="" feedwater/main="" turbine="" trip="" system="" actuation="" instrumentation,="" reactor="" vessel="" water="" level-high="" level="" 8,="" operability="" requirements,="" does="" not="" affect="" the="" ts="" bases.="" the="" trips="" are="" designed="" to="" protect="" balance="" of="" plant="" equipment="" at="" all="" rate="" power="" levels.="" the="" reactor="" vessel="" water="" level-high="" level="" 8="" trips="" also="" protects="" fuel="" integrity="" at="">25% RTP. Therefore, the proposed TS change 
to the operability requirements for the feedwater/main turbine trip 
system actuation instrumentation does not involve a reduction in a 
margin of safety.
Section 4: Permit Operability of Low Pressure Coolant Injection While 
Aligned to Shutdown Cooling
    1. The proposed Technical Specifications change does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    The LPCI [low pressure coolant injection] mode of RHR is an 
accident mitigator, not an initiator. Currently, the LPCI mode of RHR 
is an automatic Emergency Core Cooling System (ECCS) function during 
OPCONs 4 and 5. However, shutdown cooling has been an accident 
initiator in many industry events. Reliance on this loop of RHR for 
LPCI does not increase the probability of an accident in shutdown 
cooling, but the alignment for LPCI will, in itself, terminate the 
draindown event by exiting the shutdown cooling mode. This proposed 
change will permit the operability of one LPCI subsystem while the 
components of that subsystem are aligned and operating in the Shutdown 
Cooling mode of RHR, provided all other components of that subsystem 
are operable and can be manually realigned from the Main Control Room, 
if required. The required number of operable Emergency Core Cooling 
Systems (ECCS) remains unchanged, thus maintaining the TS required 
subsystem redundancy (TS Section 3.5.2 requires two operable ECCS 
subsystems with exception for Reactor level). With this change, the 
required number of LPCI subsystems are capable of performing their 
function of limiting and/or mitigating the consequences of an accident, 
by allowing the manual alignment of one LPCI subsystem, during OPCONs 4 
and 5. This allowance is justified since the change only applies to 
OPCONs 4 and 5, when reactor temperature, and associated heat loads are 
sufficiently low to provide the operator sufficient time to perform the 
manual realignment, from the Main Control Room, of the RHR pump suction 
valves and restart of the pump following LPCI injection conditions. 
Similar allowances for LPCI are currently permitted during OPCON 3, 
since the decay heat loads are significantly reduced compared to OPCON 
1, which is the mode of operation under which ECCS capability is 
analyzed (Section 6.3 of the LGS [Limerick Generating Station] Updated 
Final Safety Analysis Report (UFSAR)). The change will not increase the 
probability of occurrence or consequences of a malfunction of equipment 
since there will be no physical changes made to plant equipment nor the 
method of their operation that would result in an unanalyzed condition. 
PECO Energy [Philadelphia Electric Company] evaluated the need for 
manual realignment of the pump minimum flow path since operating in 
Shutdown Cooling typically results in the isolation of the pump minimum 
flow path to prevent inadvertent draining of the reactor vessel. The 
associated pump is still operable since this change is limited to 
OPCONs 4 and 5, when reactor pressure is sufficiently low to allow 
immediate injection to the reactor vessel without a minimum flow path. 
In situations, while in OPCON 4, where reactor pressure may not be 
sufficiently low to allow injection, the RHR system will not be aligned 
for Shutdown Cooling, since the reactor vessel pressure will be greater 
than the RHR ``cut-in'' permissive pressure. In addition, 
Administrative Controls are currently in place to realign RHR to the 
LPCI mode for planned pressure increases. Finally, this change is 
consistent with NUREG-1433 ``Standard Technical Specifications.'' 
Therefore, these changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The LPCI mode of RHR is an accident mitigator, not an initiator. 
This change will not reduce the number of required ECCS during OPCONs 4 
and 5. This change will permit the operability of one LPCI subsystem 
while the components of that subsystem are aligned and operating in the 
Shutdown Cooling mode of RHR. The change does not alter current methods 
of plant operation nor does the change make a physical change to plant 
equipment resulting in an unanalyzed malfunction of equipment. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    The basis of TS Section 3.5.2 is to ensure sufficient ECCS capacity 
to maintain core cooling in OPCONs 4 and 5. This proposed change does 
not affect the required number of ECCS during OPCONs 4 and 5; 
therefore, adequate capability through subsystem redundancy is 
maintained. The amount of time required to obtain rated LPCI conditions 
is increased due to the manual realignment, from the Main Control Room, 
of the suction valves and restart of the RHR pump following LPCI 
injection conditions. This change is in conformance with the current TS 
bases, since the operator has sufficient time to perform the manual 
realignment, during OPCONs 4 and 5, ensuring sufficient ECCS capability 
to maintain core coverage. In addition, NUREG-1433 BASES states, in 
part, ``One LPCI subsystem may be aligned for decay heat removal and 
considered OPERABLE for the ECCS function, if it can be manually 
realigned (remote or local) to the LPCI mode and is not otherwise 
inoperable. Because of low pressure and low temperature conditions in 
MODES 4 and 5, sufficient time will be available to manually align and 
initiate LPCI subsystem operation to provide core cooling prior to 
postulated fuel uncover.'' Therefore, this change will not involve a 
significant reduction in a margin of safety.
Section 5: Remove Temperature Requirement for Operational Condition 5
    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed TS change does not involve a physical change in the 
configuration of any systems important to safety. The elimination of a 
temperature requirement from the definition of OPCON 5 was reviewed for 
potential effect on reactor coolant system materials and for potential 
effect on reactivity. This TS change does not result in system 
temperature and pressure change or reactivity changes not previously 
analyzed. The reactor pressure vessel will still be restricted to the 
temperature and pressure limits of TS Section 3/4.4.6 which includes 
heatup/cooldown rates and minimum boltup limits. The reactor pressure 
vessel temperature and pressure limits will still ensure proper 
protection of the reactor coolant system materials. Therefore, this TS 
change does not increase the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed TS change does not involve any physical change in 
plant configuration, and reactor coolant system temperature and 
pressure are still restricted per TS Selection 3/4.4.6. The decrease in 
moderator density corresponding to the potential change in temperature 
(i.e., above 140 deg.F and below 200 deg.F) would have a negligible, 
however conservative effect on shutdown margin. Therefore, this TS 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    This proposed TS change does not change the reactor coolant system 
material restrictions as defined in TS Section 3/4.4.6. Therefore, the 
reactor pressure vessel will still be maintained under the current 
temperature and pressure restrictions as well as the current boltup 
limits.
    The decrease in moderator density corresponding to the potential 
temperature change from 140 deg.F to 200 deg.F is insignificant and 
would afford approximately the same moderator effect. Therefore, 
shutdown margin could only be improved (although marginally) at these 
evaluated temperatures. The actual coolant temperature will be 
administratively controlled to provide for personnel safety. Therefore, 
this change will not involve a reduction in a margin of safety.
Section 6: Reduce Frequency of Alternate Decay Heat Demonstration
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed TS change does not involve any physical changes to 
plant systems or equipment. This proposed TS change will allow the use 
of either an ``analytical approach'' (i.e., calculation) or 
``demonstration'' to ensure the operability of an alternate decay heat 
removal method. This proposed TS change does not involve any physical 
changes to plant systems or components, nor does it affect the 
capability, availability, or operability of any decay heat removal 
systems/methods (e.g., Shutdown Cooling). The Shutdown Cooling mode of 
operation of the Residual Heat Removal (RHR) system, and Residual Heat 
Removal Service Water (RHRSW) system, are not impacted by this proposed 
TS change, and will continue to function as designed to remove decay 
heat loads from the reactor primary coolant system. The RHRSW system 
and various modes of operation of the RHR system, e.g., Low Pressure 
Coolant Injection (LPCI) are not accident initiators, since these 
systems function to mitigate the consequences of an accident. This 
proposed TS change is consistent with the criteria delineated in the 
Improved Standard TS (i.e., NUREG-1433, ``Standard Technical 
Specifications, General Electric Plants, BWR/4,'' dated September 28, 
1992).
    Therefore, the proposed TS change does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    This proposed TS change does not involve any physical changes to 
plant systems or equipment. The proposed TS change will allow the use 
of a ``calculation'' or ``demonstration'' as the means for determining 
the operability of an alternate decay heat removal method. The proposed 
TS change does not involve any physical changes to plant systems or 
equipment. This proposed TS change will not affect the operation of the 
Shutdown Cooling mode of the RHR system. This mode of operation will 
continue to function as designed to remove decay heat loads from the 
reactor primary coolant system. This proposed TS change will not impact 
the operation of the other modes of operation of the RHR system (e.g., 
LPCI), nor will it affect the operation of the RHRSW system. These 
systems will continue to function as designed, which is to mitigate the 
consequences of an accident. This proposed TS change will not introduce 
the potential for equipment malfunctions or failures. This proposed TS 
change is consistent with the criteria delineated in the Improved 
Standard TS (i.e., NUREG-1433).
    Therefore, the proposed TS change does not create the possibility 
of a new or different kind of accident from any previously evaluated.
    3. The proposed TS change does not involve a significant reduction 
in a margin of safety.
    The proposed change to the TS does not involve any physical changes 
to plant systems or equipment. This proposed TS change does not make 
any physical modifications to plant systems or equipment, and is 
consistent with the criteria delineated in the Improved Standard TS 
(i.e., NUREG-1433). The proposed TS change will not impact any mode of 
operation of the RHR system or the RHRSW system.
    This proposed TS change involves revising TS ACTION statements, and 
associated supporting Bases sections, to allow for the use of a 
``calculation'' or ``demonstration'' to ensure the operability of an 
alternate decay heat removal method. The bases for the TS sections 
affected by this proposed change indicate that sufficient heat removal 
capability, system redundancy, and coolant circulation will be 
available to facilitate decay heat removal and mixing to assure 
accurate temperature indication.
    This proposed TS change does not affect the function or 
availability of any decay heat removal system or method.
    Therefore, the proposed TS change does not involve a reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: John F. Stolz.

Power Authority of the State of New York

Docket No. 50-333

    James A. FitzPatrick Nuclear Power Plant, Oswego County, New York.
    Date of amendment request: October 3, 1994.
    Description of amendment request: The proposed amendment would 
extend the functional test intervals and allowable out-of-service times 
for some of the instruments subject to requirements of the Technical 
Specifications (TSs). These proposed changes are based upon NRC-
approved Licensing Topical Reports prepared under the direction of the 
Boiling Water Reactors Owners Group and intended to enhance plant 
safety by reducing the potential for test related scrams, excessive 
test cycles on equipment, and operator errors. The proposed amendment 
would also: (1) Remove the Average Power Range Monitor (APRM) downscale 
scram function from the TSs, remove instrument response time values 
from the TSs in accordance with Generic Letter 93-08, and incorporate 
various editorial changes and clarifications into the TSs. The proposed 
amendment involves reactor protection system, primary containment 
isolation, emergency core cooling, control rod block, and anticipated 
transient without scram recirculation pump trip instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the proposed 
Amendment would not involve a significant hazards consideration as 
defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated because:

a. Incorporate STI [Surveillance Test Interval] and AOT [Allowable Out-
Of-Service Time] Improvement--Category 1

    The proposed changes are limited to an extension of the 
surveillance testing intervals and allowable out-of-service times of 
plant instrumentation. The changes do not introduce any new modes of 
plant operation, make any physical changes, or alter any operational 
setpoints. Therefore, the changes do not degrade the performance of any 
safety system assumed to function in the accident analysis. 
Consequently, there is no effect on the probability of occurrence of an 
accident.
    Regarding the consequences of an accident, the GE [General Electric 
Company] Licensing Topical Reports (References 1 through 7) concluded 
that the proposed extensions in the STI and AOT for the safety system 
instrumentation results in an insignificant change in the core damage 
frequency. The extension of the STI/AOTs results in a slight increase 
in the unavailability of the safety functions. However, this effect is 
offset by a reduction in the probability of inadvertent plant trips due 
to reduced testing. The overall effect on the probability of an 
accident is negligible. While the effects of reducing unnecessary 
cycles on safety system instrumentation is not quantifiable, the effect 
will be to further reduce the core damage frequency. The NRC concurred 
in their SERs [Safety Evaluation Reports] (References 8 through 15) 
with these conclusions. Consequently, there is not a significant 
increase in the consequences of an accident.

b. Relocation of the Instrument Response Time Limits--Category 2

    The change involves the use of an alternate regulatory process for 
controlling the instrument response time limits. The change does not 
introduce any new modes of plant operation, make any physical changes, 
alter any operational setpoints, or change the surveillance 
requirements.

c. Delete APRM Downscale Scram--Category 3

    The design basis accident applicable to the startup power region is 
the Control Rod Drop Accident (CRDA). The FSAR [Final Safety Analysis 
Report] does not credit the APRM downscale scram in the termination of 
this accident. Accident mitigation is provided by the APRM fixed high 
neutron flux scram. Therefore, elimination of this scram functions has 
no adverse affect on previously evaluated accidents.

d. Miscellaneous Changes--Category 4

    The changes do not introduce any new modes of plant operation, make 
any physical changes, or alter any operational setpoints. The changes 
involve enhancements that clarify the Technical Specification 
requirements.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated because:

a. Incorporate STI and AOT Improvements--Category 1

    The proposed changes do not introduce any new accident initiators 
or failure mechanisms since the changes do not introduce any new modes 
of plant operation, make any physical changes, or alter any operational 
setpoints. The changes reduce the probability of accidents initiated by 
test-induced plant trips.

b. Relocation of the Response Time Limits--Category 2

    The change involves the use of an alternate process for controlling 
the instrument response time limits. The change does not introduce any 
accident initiators since it does not involve any new modes of plant 
operation, make any physical changes, alter any operational setpoints, 
or change the surveillance requirements.

c. Delete APRM Downscale Scram--Category 3

    Scram functions are intended to shutdown the reactor following 
transients or accidents and their removal does not introduce an 
accident initiator. The limiting accident evaluated in the FSAR for the 
startup power region is the control rod drop accident. This accident is 
assumed to occur irrespective of the scram functions provided to 
terminate this accident.

d. Miscellaneous Changes--Category 4

    The changes do not introduce any new accident initiators or failure 
mechanisms since the changes do not alter the physical characteristics 
of any plant system or component. The changes involve enhancements that 
clarify the Technical Specification requirements.
    3. Involve a significant reduction in the margin of safety because:

a. Incorporate STI and AOT Improvements--Category 1

    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The affected instrumentation setpoints 
already account for the effects of drift and include sufficient 
allowance for an extension in the STIs. The evaluations presented in 
the referenced Licensing Topical Reports concluded that the overall 
effect of the proposed changes provides a net increase in plant safety. 
The improvement is achieved by reducing the potential for: (a) Test 
related plant scrams (reduced challenges to plant shutdown systems and 
improved plant availability); (b) excessive test cycles on equipment 
(reduced wear-out potential); (c) operator errors (AOT provides 
reasonable time for making repairs and tests); (d) scrams that occur 
when inoperable channels are tripped because insufficient repair time 
exists; and (e) diversion of plant personnel and resources on 
unnecessary testing (potential safety and operational improvement).

b. Relocation of the Response Time Limits--Category 2

    The change involves the use of an alternate regulatory process for 
controlling the instrument response time limits. The change does not 
introduce any new modes of plant operation, make any physical changes, 
alter any operational setpoints, or change the surveillance 
requirements.

c. Delete APRM Downscale Scram--Category 3

    The only scram function that the UFSAR [Updated Final Safety 
Analysis Report] takes credit for in the mitigation of the limiting 
accident (control rod drop accident) is the APRM 15% power fixed high 
neutron flux scram. This scram function, as well as the IRM 
[Intermediate Range Monitor] high flux scram function in the startup 
mode which could also terminate this accident, are not affected by this 
change. Only the APRM downscale scram, for which the UFSAR takes no 
credit in the termination of any analyzed event, is eliminated by this 
change. The APRM downscale control rod block is not affected by this 
change. Elimination of the APRM downscale scram will avoid the need to 
operate the plant in a ``half scram'' condition for certain IRM/APRM 
channel bypass combinations, therefore eliminating the potential for an 
inadvertent plant transient. For these reasons, the change does not 
involve a significant reduction in the safety margin.

d. Miscellaneous Changes--Category 4

    The changes assure compliance with the Technical Specifications by 
improving its clarity and accuracy. For these reasons the changes will 
improve the plant's margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Ledyard B. Marsh.

Power Authority of the State of New York

Docket No. 50-333

    James A. FitzPatrick Nuclear Power Plant, Oswego County, New York.
    Date of amendment request: October 7, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.6.E.4 to establish that the 
manual cycling of reactor coolant system (RCS) safety/relief valves 
(SRVs) during plant startups is to be accomplished within 12 hours 
after steam pressure and flow are adequate to perform the testing. TS 
4.6.E.4 currently requires this testing to be performed within 12 hours 
of continuous power operation at a reactor steam dome pressure of at 
least 940 psig. This change was proposed to minimize the potential for 
undesirable pressure transients in the RCS. The amendment would also 
make several editorial changes to clarify the intent of TS's involving 
SRV valve testing and performance 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:
    Operation of the James A. FitzPatrick Nuclear Power Plant in 
accordance with the proposed amendment would not involve a significant 
hazards consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because the proposed 
changes do not change the test method or conditions under which valve 
testing may be performed and there is no affect on assumptions used for 
previously analyzed accidents. The original operating license for 
FitzPatrick did not specify any time limit for completing manual 
testing of the safety/relief valves.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated because the proposed amendment does not 
involve any modification of plant equipment or changes in plant 
operating conditions.
    3. Involve a significant reduction in the margin of safety because 
the proposed amendment makes no changes to the operability of 
performance requirements for the safety/relief valves including the ADS 
[Automatic Depressurization System] function. Valve lift setpoints and 
the minimum number of operable valves required are not affected.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Ledyard B. Marsh.

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: September 9, 1994.
    Description of amendment request: The proposed amendment modifies 
the visual inspection for snubbers in the Technical Specifications and 
is consistent with the guidance provided in Generic Letter 90-09.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve no hardware changes, no changes to the 
operation of snubbers, and does not change the ability of the snubbers 
to perform their intended functions. Visual inspection of snubbers is a 
separate process that complements the functional testing program. The 
NRC has concluded that functional testing of snubbers provides a 95 
percent confidence level and 90 to 100 percent of the snubbers will 
operate within the specified acceptance limits. Any change in the 
visual inspection frequency will not have any significant impact on the 
operability of the snubbers.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes will not result in an unanalyzed condition. 
Replacing the current method of determining visual surveillance 
intervals with a new method approved by the NRC in Generic Letter 90-09 
will not change the level of confidence in snubber operability. A new 
procedure for determining visual inspection frequencies will not result 
in an unreviewed failure mechanism.
    3. Will not involve a significant reduction in a margin of safety.
    The proposed changes incorporate the alternate Technical 
Specification requirements for visual inspection of snubbers identified 
in Generic Letter 90-09. The alternate visual inspection criteria 
consider the size of the category of snubbers when evaluating 
inspection intervals due to failure rates. Since the functional testing 
requirements remain unchanged and do not reduce the operability 
confidence levels, there is no resultant change in any margins of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

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: September 20, 1994
    Description of amendment request: The proposed amendment modifies 
the Technical Specifications for auxiliary feedwater to reduce the 
secondary side steam pressure required for testing the steam turbine 
driven auxiliary feedwater pump (AFW). The proposed amendment also 
clarifies the time required to perform the steam turbine driven 
auxiliary feedwater pump surveillance test when entering Mode 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed change to the minimum required test pressure for the 
steam turbine driven AFW pump does not affect the operation of the pump 
during conditions when it is required to performed its safety function. 
The clarification that the secondary side steam pressure is steam 
generator pressure is editorial. Reduced Tavg and increased steam 
generator tube plugging will affect the normal operating secondary side 
steam pressure.
    However, the zero load secondary side steam pressure is not 
affected, therefore, the conditions in which the steam turbine driven 
AFW pump will be required to perform its safety function are not 
changed.
    Providing a specific time frame in which to perform the 
surveillance test after attaining the required steam pressure ensures 
that the test will be performed in a timely manner. The time frame 
specified is consistent with NUREG-1431, Standard Technical 
Specifications--Westinghouse Plants.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident.
    The proposed changes do not change system configurations, plant 
equipment, or analysis. Therefore, the proposed changes will not 
increase the possibility of a new or different kind of accident from 
any accident previously identified.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the minimum required steam pressure will not 
affect the heat removal capability of the AFW System. Therefore, the 
value assumed in the safety analysis is not changed. The change to the 
specification 4.0.4 exemption to provide a specific time period does 
not affect any margins of safety. Therefore, these changes do not 
involve a significant reduction in any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz.

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: September 20, 1994.
    Description of amendment request: These proposed changes would 
adopt the Westinghouse Standard Technical Specifications (NUREG-1431) 
Channel Functional Test surveillance frequency for the Manual Reactor 
Trip Switches and for the Reactor Trip Breakers (RTB) and relocate RTB 
maintenance requirements from the Technical Specifications to the 
Salem, Units 1 and 2, Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does not involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes do not affect accident conditions or 
assumptions. They change the existing surveillance test and their 
frequencies to make them consistent with industry standards, and 
relocate maintenance requirements to the UFSAR [Updated Final Safety 
Analysis Report].
    The changes, for the Manual Reactor Trip Switch and Reactor Trip 
Breaker (RTB) CHANNEL FUNCTIONAL TEST frequency, incorporate the 
established Westinghouse STS surveillance frequencies. These 
surveillance frequencies have received previous NRC review and generic 
approval via the issuance of NUREG-1431. The Westinghouse STS does not 
require Channel Functional Test for the Manual Reactor Trip Switches or 
the RTB prior to each reactor startup.
    The addition of the RTB shunt trip feature for automatic reactor 
trips, the improved RTB maintenance activities developed over the past 
several years, and the implementation of 10 CFR 50.62 requirements have 
improved RTB reliability. These features are unaffected by the proposed 
changes. Excessive RTB testing results in increased component wear and 
possibly reduced component life. Testing the RTBs with associated logic 
trains reduces the potential for human errors and associated plant 
transients.
    The consequences of accidents previously evaluated are unaffected 
by the proposed changes.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes do not modify any system or equipment, nor 
alter any process function. The Manual Reactor Trip Switch and RTB 
functionality remains unchanged. Therefore these changes do not create 
a new or un-evaluated accident or operating condition.
    3. Does not involve a significant reduction in a margin of safety.
    The proposed changes adopt the NRC approved Westinghouse STS 
surveillance testing frequencies to maintain RTB reliability. Reduced 
testing at power, consistent with the associated logic train test 
frequency, reduces the potential for inadvertent actuation and 
personnel errors. Thus, the proposed changes enhance plant safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

Saxton Nuclear Experimental Corporation

Docket No. 50-146

    Saxton Nuclear Facility, Bedford County, Pennsylvania.
    Date of amendment request: August 8, 1994. This supersedes the 
request dated June 23, 1993.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to allow characterization 
activities related to the decommissioning of the Saxton Nuclear 
Facility and add administrative activities associated with the 
characterization activities.
    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 a significant hazards 
considerations because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The activities associated with characterization of the facility 
will have a minimum impact on the physical condition of the containment 
vessel as it relates to the risk of fire and has no effect on the risk 
of flooding.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    In its present condition, the only accidents applicable to the site 
are fire, flood, and radiological hazard. The possibility of a new or 
different type of accident than that previously evaluated in the FSAR 
will not be created by the implementation of activities permitted by 
the approval of this amendment request.
    3. Involve a significant reduction in a margin of safety.
    No margins of safety relevant to the equipment at the facility 
exist. Activities involved in characterization will not involve a 
reduction in a margin of safety.
    The NRC staff has reviewed the analysis of the licensee and, based 
on this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Saxton Community Library, 911 
Church Street, Saxton, Pennsylvania 16678.
    Attorney for the Licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts, and Trowbridge, 2300 N Street, NW, Washington, DC 
20037.
    NRC Project Director: Seymour H. Weiss.

Southern California Edison Company, et al.

[Docket Nos. 50-361 and 50-362

    San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego 
County, California.
    Date of amendment request: August 26, 1994.
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 3/4.7.5, ``Control Room Emergency Air 
Cleanup System.'' The proposed revision to TS 3/4.7.5 will provide a 
Limiting Condition of Operation (LCO) 3.0.4 exception for MODES 5, 6, 
or a defueled configuration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this proposed 
change involve a significant increase in the probability or 
consequences of any accident previously evaluated?
    Response: No.
    The Control Room Emergency Air Cleanup System (CREACUS) provides a 
protected environment from which operators can control the plant 
following an uncontrolled release of radioactivity or toxic gas.
    [The following are the proposed changes to Technical Specification 
3/4.7.5 ``Control Room Emergency Air Cleanup System:'']
    Proposed Change 1 [adds the following statement to the 
Applicability statement of TS 3.7.5: ``or during movement of irradiated 
fuel assemblies.''] will replace the existing wording of the 
Applicability with the following words ``ALL MODES or during movement 
of irradiated fuel assemblies.'' The requirement concerning movement of 
irradiated fuel assemblies was added because the existing Applicability 
statement does not reflect the possibility of radiation exposure to the 
operators inside the control room during this event. A fuel handling 
accident can happen during defueled operations. In this case, movement 
of the last irradiated fuel assembly from the empty core inside 
containment is not covered by the existing Applicability.
    Also, a fuel handling accident can happen inside the Fuel Handling 
Building when irradiated fuel is moved from one location to another in 
the Spent Fuel Pool (SEP). The need for the CREACUS during fuel 
handling is based on safety analysis assumptions which are specified in 
Chapter 15 of the SONGS Unit 2 and 3 Updated Final Safety Analysis 
Report (UFSAR).
    Addition of the new Applicability requirement will not involve a 
significant increase in the possibility or consequences of any accident 
previously evaluated.
    Proposed Change 2 [a new Action d): ``The provisions of 
Specification 3.0.4 are not applicable when entering MODES 5, 6, or 
defueled configuration'' is added to the Action section of TS 3.7.5] 
will add a new Action d) which reads: ``the provisions of Specification 
3.0.4 are not applicable when entering MODES 5, 6, or defueled 
configuration.'' Existing Technical Specification 3/4.7.5 prohibits 
entering MODE 6 from a defueled configuration unless both CREACUS 
trains are OPERABLE. With the addition of the statement ``or during 
movement of irradiated fuel assemblies'' to the Applicability, 
OPERABILITY of the CREACUS will be ensured prior to movement of 
irradiated fuel assemblies. Therefore, the only threshold between 
defueled configuration and MODE 6 is the position of the first 
irradiated fuel assembly--whether it is in the reactor vessel or 
external to it. This threshold has no safety significance because the 
only credible event during the transition from a defueled configuration 
to MODE 6 and from MODE 6 to defueled configuration is a Design Basis 
Fuel Handling Accident which is covered by the proposed Applicability. 
Therefore, this threshold can be expected from Limiting Condition for 
Operation (LCO) 3.0.4.
    The threshold of entering MODE 5 from MODE 6 consists of fully 
tightening the last reactor vessel head closure bolt. This evolution 
has no safety significance from the point of view of isolating the 
control room from external hazards. Therefore, this MODE change can be 
excepted from LCO 3.0.4. The threshold of entering MODE 6 from MODE 5 
consists of untightening at least one reactor vessel head closure bolt. 
If no irradiated fuel assemblies are being moved, this evolution has no 
safety significance from the point of view of isolating the control 
room from external hazards. Therefore, this MODE change can be excepted 
from LCO 3.0.4 also.
    The threshold of entering MODE 5 from MODE 4 consists of decreasing 
Reactor Coolant System (RCS) temperature from 350 deg.F > Tavg > 
200 deg.F to Tavg [less than or equal to] 200 deg.F by initiating 
shutdown cooling. If no irradiated fuel assemblies are being moved, 
this evolution has no safety significance from the point of view of 
isolating the control room from external hazards. Therefore, this MODE 
change can be excepted from LCO 3.0.4.
    The MODE changes have no significance relative to releases. 
Therefore, since CREACUS can be inoperable during each individual mode, 
it should not be required to have two OPERABLE CREACUS trains before 
mode changes.
    Therefore, addition of the new Action will not involve a 
significant increase in the probability or consequences of any accident 
previously evaluated.
    Proposed Change 3 [adds the following words ``or defueled 
configuration when moving irradiated fuel assemblies'' after the words 
``Units 2 and 3 in MODE 5 or 6'' in the Action statement of TS 3.7.5] 
will add the words ``or defueled when moving irradiated fuel 
assemblies'' to the Action statement when either Unit is in MODE 5 or 
6. These words are added for consistency with a proposed Applicability 
statement ``or during movement of irradiated fuel assemblies.'' Without 
these words it is not clear what Actions should be entered if the LCO 
requirement is not met in a defueled configuration when moving 
irradiated fuel assemblies. By adding these words Actions (a) and (b) 
became applicable in a defueled configuration when moving irradiated 
fuel assemblies. This change applies the requirement of the proposed 
Applicability to the Action when either Unit is in MODES 5 or 6. 
Therefore, addition of these words to the Action statement will not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    Proposed Change 4 [adds the following words ``or movement of 
irradiated fuel assemblies'' after the words ``suspend all operations 
involving CORE ALTERATIONS or positive reactivity changes'' in the 
Action (b) statement of TS 3.7.5] will add the words ``or movement of 
irradiated fuel assemblies'' in the Action (b) statement. These words 
are added for consistency with the proposed Applicability statement and 
proposed Action statement when either Unit is in MODES 5 or 6, or a 
defueled configuration when moving irradiated fuel assemblies. Without 
addition of these words Action (b) did not specify what should be done 
when any Unit is in a defueled configuration when moving irradiated 
fuel assemblies. Therefore, addition of these words to the Action 
statement will not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Will operation of the facility in accordance with this proposed 
change create the possibility of a new or different kind of accident 
from any previously evaluated?
    Response: No.
    The changes proposed herein do not reduce the reliability or 
performance of the Control Room Emergency Air Cleanup System (CREACUS). 
The proposed LCO 3.0.4 exception for CREACUS permits MODE 5, MODE 6, or 
defueled configuration entry with one train of CREACUS inoperable. This 
change does not affect CREACUS reliability and its capability to 
perform its intended design functions.
    Additional requirements in the Applicability to have two Control 
Room Emergency Air Cleanup Systems OPERABLE during movement of 
irradiated fuel covers the consequences of a fuel accident in the Fuel 
Handling Building and in containment when the reactor vessel is 
defueled. Operation of the facility will remain unchanged as a result 
of the proposed changes.
    Also, addition of the requirement to suspend movement of irradiated 
fuel assemblies when either Unit is in a defueled configuration when 
moving irradiated fuel is made for consistency with the proposed 
Applicability statement and Action statement. The proposed Action 
statement emphasize that Actions (a) and (b) are applicable not only 
when either Unit is in MODES 5 or 6, but also when in a defueled 
configuration when moving irradiated fuel assemblies. This change does 
not affect CREACUS reliability and its capability to perform its 
intended design functions. Therefore, the proposed changes will not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Will operation of the facility in accordance with this proposed 
change involve a significant reduction in a margin of safety?
    Response: No.
    Operation of the facility in accordance with these changes will not 
be adversely affected as a result of the changes proposed herein. The 
proposed changes include a change to the Applicability, adding the new 
Action (d), modifying the Action statement when either Unit is in MODES 
5 or 6, and modifying the Action (b). The proposed LCO 3.0.4 exception 
for CREACUS permits MODE 5, MODE 6, or defueled configuration entry 
with one train of CREACUS inoperable. Additional requirements in the 
Applicability statement to have two Control Room Emergency Air Cleanup 
Systems OPERABLE during movement of irradiated fuel, covers the 
consequences of the fuel accident in the Fuel Handling Building. Also, 
this requirement covers the movement of irradiated fuel when the 
reactor vessel is defueled. Modified Action statement for either Unit 
in MODES 5 or 6 is made for consistency with the proposed Applicability 
statement. Modified Action (b) covers the possibility of both the 
CREACUS trains being inoperable in a defueled configuration when moving 
irradiated fuel assemblies.
    The margin of safety as defined in Bases 3/4.7.5 is limiting the 
dose to control room personnel to 5.0 rem or less whole body, or its 
equivalent. As discussed above, operation of the CREACUS will be 
unchanged as a result of the proposed changes. Therefore, operation of 
the facility in accordance with this 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: Theodore R. Quay.

Southern California Edison Company, et al.

Docket Nos. 50-361 and 50-362.

    San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego 
County, California.
    Date of amendment requests: September 16, 1994.
    Description of amendment requests: The licensee proposes to revise 
the linear heat rate (LHR) limit in Technical Specification (TS) 3/
4.2.1, ``Linear Heat Rate.'' TS 3/4.2.1 requires maintaining the LHR at 
or below 13.9 kilowatts per linear foot (kw/ft) for steady-state 
operation. This amendment request is to revise this value from 13.9 kw/
ft to 13.0 kw/ft. The Bases of TS 3/4.2.1, ``Linear Heat Rate,'' are 
also being revised to reflect the new value.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this proposed 
change involve a significant increase in the probability or 
consequences of any accident previously evaluated?
    Response: No.
    The only event impacted by this Technical Specification (TS) change 
is the Large Break Loss of Coolant Accident (LBLOCA) which has been 
reanalyzed. There is a direct correlation between the magnitude of the 
TS 3/4.2.1 Linear Heat Rate (LHR) limit and the calculated peak 
cladding temperature (PCT). Since the LHR is being reduced in value, 
which is a conservative change, there will be no increase in the 
consequences of the event. The LBLOCA reanalysis, performed using the 
new LHR limit in support of an optimized fuel loading pattern, resulted 
in a reduction of the calculated LBLOCA PCT. Therefore, this change 
will not involve an increase in the probability or consequences of any 
accident previously evaluated.
    2. Will operation of the facility in accordance with this proposed 
change create the possibility of a new or different kind of accident 
from any previously evaluated?
    Response: No.
    This amendment request does not involve any change to plant 
equipment or operation. The linear heat rate limit provided in T/S 
3.2.1 is used only in the LBLOCA analysis. No change to the LBLOCA 
methodology was made. Therefore, this change does not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    3. Will operation of the facility in accordance with this proposed 
change involve a significant reduction in a margin of safety?
    Response: No.
    This amendment does not change the manner in which safety limits, 
limiting safety settings, or limiting conditions for operation are 
determined. There is no change in the PCT acceptance criterion for this 
event as a result of the proposed reduction in the LHR limit. 
Therefore, there is no reduction in the margin of safety from the 
acceptance limit to the mechanical failure point of the fuel. 
Additionally, the analysis value for the LBLOCA PCT is reduced to 2160 
deg.F. This results in an increase in the analysis margin between the 
acceptance criterion and the analysis value. Therefore, this proposed 
change does 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713.
    Attorney for licensee: T.E. Oubre, Esquire, Southern California 
Edison Company, P.O. Box 800, Rosemead, California 91770.
    NRC Project Director: Theodore R. Quay.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company

Docket No. 50-346

    Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio.
    Date of amendment request: October 7, 1994.
    Description of amendment request: The proposed amendment would 
remove the existing Surveillance Requirement (SR) 4.5.2.d.3 for the Low 
Pressure Injection (LPI) System and the existing SR 4.6.2.1.c for the 
Containment Spray (CS) System since the requirement to leak test these 
systems is programmatically covered in TS 6.8.4.a, ``Primary Coolant 
Sources Outside Containment.'' Additionally, changes are proposed to TS 
Bases 3/4.5.2 and 3/4.6.2.1 to reflect the elimination of the above 
SRs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the NRC staff has 
reviewed the licensee's analysis against the standards of 10 CFR 
50.92(c). The staff's review is presented below:
    (1) The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The change does not involve a significant increase in the 
probability of an accident previously evaluated nor does it involve a 
significant increase in the consequences of an accident previously 
evaluated because no accident initiators, conditions or assumptions are 
affected by removing the leak test requirements of LPI System SR 
4.5.2.d.3 and CS System SR 4.6.2.1.c. The purpose of these SRs is 
already encompassed by the existing program requirements of TS 6.8.4.a, 
``Primary Coolant Sources Outside Containment.'' TS 6.8.4.a requires 
integrated leak testing at refueling cycle intervals or less, for each 
system outside containment, that could contain highly radioactive 
fluids during a serious transient or accident.
    The proposed changes do not alter the source term, containment 
isolation, or allowable releases. The proposed changes, therefore, will 
not increase the radiological consequences of a previously evaluated 
event.
    These changes are consistent with NUREG-1430, Revision 0, 
``Improved Standard Technical Specifications for B&W Plants.'' The 
associated changes to TS Bases 3/4.5.2 and 3/4.6.2.1 are 
administrative.
    (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 create the possibility of any new or 
different kind of accident from any accident previously evaluated 
because no new accident initiators or assumptions are introduced by 
these proposed changes to LPI System SR 4.5.2.d.3 and CS System SR 
4.6.2.1.c. The purpose of these SRs is already encompassed by the 
existing program requirements of TS 6.8.4.a, ``Primary Coolant Sources 
Outside Containment,'' which requires leak testing to be performed on 
the LPI and CS Systems. These changes are consistent with NUREG-1430. 
The associated changes to TS Bases 3/4.5.2 and 3/4.6.2.1 are 
administrative. The proposed changes do not alter any accident 
scenarios.
    (3) The proposed changes do not result in a significant reduction 
in the margin of safety.
    The changes do not involve a significant reduction in the margin of 
safety because the proposed changes to the LPI System SR 4.5.2.d.3 and 
CS System SR 4.6.2.1.c do not reduce or adversely affect the 
capabilities of any plant structures, systems or components. The 
purpose of these SRs is already encompassed by the existing program 
requirements of TS 6.8.4.a, ``Primary Coolant Sources Outside 
Containment.'' These changes are consistent with NUREG-1430. The 
associated changes to TS Bases 3/4.5.2 and 3/4.6.2.1 are 
administrative.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Acting Project Director: C.A. Carpenter.

Virginia Electric and Power Company

Docket Nos. 50-280 and 50-281

    Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
    Date of amendment request: October 11, 1994.
    Description of amendment request: The proposed changes would modify 
the surveillance frequencies of the containment hydrogen analyzers.
    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:
    Specifically, operation of Surry Power Station in accordance with 
the proposed Technical Specifications will not:
    1. Involve a significant increase in the probability of occurrence 
or consequences of an accident previously evaluated.
    The proposed changes to the surveillance requirements for the 
hydrogen analyzers have no impact on the probability of any accident 
occurrence. The hydrogen analyzers are maintained in a standby mode 
during normal operation and are made fully operable within thirty 
minutes after a safety injection signal to provide indication of the 
hydrogen concentration in containment after a loss-of-coolant accident. 
This instrumentation is used solely post-accident to monitor 
containment conditions. Reduced testing of a post-accident monitor does 
not contribute to the probability of any previously analyzed accident. 
These monitors have no automatic safety function. Furthermore, the 
hydrogen analyzers will be operated in the same manner, and the 
operability requirements are not being altered. In addition, the Post-
Accident Sampling System serves as a diverse means to confirm post-
accident hydrogen concentration in containment. Therefore, the 
consequences of a Design Basis Accident are not being increased by the 
proposed change in surveillance test frequency of the hydrogen 
analyzers.
    Reducing the frequency of surveillance testing could however 
decrease the timeliness in identifying an inoperable hydrogen analyzer. 
However, our surveillance test experience has shown that the analyzers 
have been very stable with repeatable results, and we conclude that the 
change in test frequency should not affect the reliability or 
operability of the analyzers. Likewise, the NRC has determined in 
Generic Letter 93-05 that a reduced frequency of surveillance testing 
during power is acceptable to determine hydrogen analyzer operability.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no plant modifications or changes in methods of plant 
operation introduced by this change in hydrogen analyzer surveillance 
frequencies. The hydrogen analyzers are maintained in a standby mode 
during normal operation and are fully operable within thirty minutes 
after a safety injection signal to provide indication of the hydrogen 
concentration in containment after a loss-of-coolant accident. 
Therefore, the possibility of a new or different kind of accident than 
previously evaluated is not created by the proposed changes in 
surveillance frequency of the control rods [hydrogen analyzers 
surveillance frequencies].
    3. Involve a significant reduction in a margin of safety.
    The hydrogen analyzer surveillance requirements do not affect the 
margin of safety in that the operability requirements for the safety 
systems and containment remain unchanged. The hydrogen analyzers only 
provide indication and do not perform any direct function to mitigate 
the consequences of any previously analyzed accidents. Furthermore, the 
change in surveillance frequency is associated with a post-accident 
monitor with no automatic safety functions and a diverse means of 
confirming the parameter by the Post-Accident Sampling System. 
Therefore, the margin of safety is not altered by this proposed change 
in the surveillance frequencies of the hydrogen analyzers.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Mohan C. Thadani (Acting).

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

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

Northeast Nuclear Energy Company, et al.

Docket No. 50-336

    Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut.
    Date of amendment request: September 26, 1994.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications by adding a footnote to 
Surveillance Requirement 4.6.1.2.d that defers the performance of Type 
B and C containment leak rate tests to the end of the twelfth refueling 
outage.
    Date of publication of individual notice in Federal Register: 
October 13, 1994, (59 FR 52005).
    Expiration date of individual notice: November 14, 1994.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community--Technical College, Thames Valley Campus, 574 
New London Turnpike, Norwich, CT 06360.

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.

Arizona Public Service Company, et al.

Docket Nos. STN 50-528, STN 50-529, and STN 50-530

    Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa 
County, Arizona.
    Date of application for amendments: August 23, 1993, as 
supplemented by letter of July 21, 1994.
    Brief description of amendments: These amendments remove the Units 
1 and 3 license condition regarding an augmented reactor coolant pump 
vibration monitoring program and the confirmatory order modifying the 
Unit 2 license regarding the same issue.
    Date of issuance: October 27, 1994.
    Effective date: October 27, 1994.
    Amendment Nos.: 84, 72, and 56.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50963).
    The additional information in the letter dated July 21, 1994, was 
clarifying in nature and did not affect the staff's previously 
published no significant hazards determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 27, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

Baltimore Gas and Electric Company

Docket Nos. 50-317 and 50-318

    Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 & 2, Calvert 
County, Maryland.
    Date of application for amendments: August 4, 1994.
    Brief description of amendments: The amendments delete Technical 
Specifications 3/4.3.3.3, 6.9.2.b, 6.9.2.d, and Bases 3/4.3.3.3, which 
provide the requirements for the operation and the testing of seismic 
monitoring instrumentation, and relocates them to the Updated Final 
Safety Analysis Report and plant procedures.
    Date of issuance: October 21, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 199 and 176
    Facility Operating License No. DPR-53 and DPR-69: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47165).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated October 21, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: 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 4, 1993.
    Brief description of amendments: These amendments revise the 
Updated Final Safety Analysis Report to address the removal of orifice 
plates in the containment vent/purge lines of each unit and revise the 
maximum hypothetical accident analysis to address the increased flow as 
the result of removing the orifice plates.
    Date of issuance: October 21, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 200 and 177.
    Facility Operating License No. DPR-53 and DPR-69: Amendment revised 
the Licenses.
    Date of initial notice in Federal Register: February 25, 1994 (59 
FR 9254).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated October 21, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Commonwealth Edison Company

Docket Nos. STN 50-454 and STN 50-455

    Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.
    Date of application for amendments: August 1, 1994, as supplemented 
by your letters dated September 7, 1994, and September 17, 1994 (two 
letters), with clarifying information submitted by letters dated 
September 22, 1994, September 23, 1994, September 30, 1994, October 17, 
1994, and October 24, 1994.
    Brief description of amendments: The purpose of the amendment is to 
incorporate voltage-based repair criteria into the Byron, Unit 1, 
technical specifications, thereby permitting the use of voltage-based 
steam generator (SG) tube plugging criteria for a specific class of SG 
tube defects.
    Date of issuance: October 24, 1994.
    Effective date: October 24, 1994.
    Amendment Nos.: 66 and 66.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1994 (59 
FR 48917).
    The clarifying information in the September 22, 1994, September 23, 
1994, September 30, 1994, October 17, 1994, and October 24, 1994, 
submittals did not affect the initial determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 24, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Byron Public Library, 109 N. 
Franklin, P.O. Box 434, Byron, Illinois 61010.

Commonwealth Edison Company

Docket Nos. STN 50-454 and STN 50-455

    Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.

Docket Nos. STN 50-456 and STN 50-457

    Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois.
    Date of application for amendments: March 23, 1994, as supplemented 
on July 26, 1994.
    Brief description of amendments: The amendments change the 
Technical Specifications to reflect a reduced thermal flow to 
compensate for increased steam generator tube plugging up to 15 percent 
of the total number of tubes. The amendment also approves the use of 
higher boron concentration in the refueling water storage tank, the 
reactor coolant system accumulators, and the refueling cavity.
    Date of issuance: October 21, 1994.
    Effective date: October 21, 1994.
    Amendment Nos.: 65, 65, 56, and 55.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 15, 1994 (59 FR 
41802).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 21, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

Consumers Power Company

Docket No. 50-255

    Palisades Plant, Van Buren County, Michigan.
    Date of application for amendment: November 15, 1991, supplemented 
February 22, March 11, April 7, and August 23, 1994.
    Brief description of amendment: This amendment is a complete 
rewrite of the instrumentation operability requirements.
    Date of issuance: October 26, 1994.
    Effective date: October 26, 1994.
    Amendment No.: 162.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27052)
    The August 23, 1994, request contained editorial changes within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration findings. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 26, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

Entergy Operations, Inc.,

Docket No. 50-313

    Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas.
    Date of amendment request: January 13, 1994.
    Brief description of amendment: The amendment revised the 
specifications governing the reactor protection system (RPS) to permit 
the plant to operate indefinitely with one RPS channel in by-pass.
    Date of issuance: October 24, 1994.
    Effective date: October 24, 1994.
    Amendment No.: 174.
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10005).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 24, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

Florida Power and Light Company

Docket Nos. 50-250 and 50-251

    Turkey Point Plant Units 3 and 4, Dade County, Florida.
    Date of application for amendments: February 18, 1994, as 
supplemented by letter dated August 5, 1994.
    Brief description of amendments: These amendments delete the audit 
frequencies from the Technical Specifications (TS) and modify the TS 
administrative control requirements for emergency and security plans.
    Date of issuance: October 26, 1994.
    Effective date: October 26, 1994.
    Amendment Nos: 168 and 162.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14889). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 26, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

GPU Nuclear Corporation, et al.

Docket No. 50-219

    Oyster Creek Nuclear Generating Station, Ocean County, New Jersey.
    Date of application for amendment: August 19, 1994.
    Brief description of amendment: The amendment updates and clarifies 
the surveillance requirements for control rod exercising and standby 
liquid control pump operability testing to be consistent with Generic 
Letter 93-05.
    Date of Issuance: October 19, 1994.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 172.
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47168).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 19, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

IES Utilities Inc.

Docket No. 50-331

    Duane Arnold Energy Center, Linn County, Iowa.
    Date of application for amendment: May 28, 1992, as supplemented on 
January 6, May 27 and October 20, 1994.
    Brief description of amendment: The amendment revised the Technical 
Specifications by changing the limiting conditions for operation and 
surveillance requirements for primary containment integrity, secondary 
containment integrity, and other systems and equipment of Section 3.7, 
Containment Systems. Limiting conditions for operation and surveillance 
requirements for drywell average air temperature and secondary 
containment automatic isolation dampers were also added.
    Date of issuance: October 26, 1994.
    Effective date: October 26, 1994, to implemented within 120 days.
    Amendment No.: 201
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34665). The licensee's October 20, 1994, submittal, provided clarifying 
information at the request of the NRC staff. This submittal did not 
change the initial application or the no significant hazards 
determination as originally noticed. Therefore, renoticing was not 
warranted.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 26, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401.

Northeast Nuclear Energy Company, et al.

Docket No. 50-423

    Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut.
    Date of application for amendment: September 30, 1993, as 
supplemented July 8, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications by increasing the minimum volume of fuel oil required to 
be stored in the emergency diesel generator (EDG) day tank from 205 
gallons to 278 gallons, and clarifies the bases for the EDG fuel oil 
storage tank and day tank minimum fuel oil volume requirements.
    Date of issuance: October 17, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 97.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59753).
    The July 8, 1994, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 17, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

Pennsylvania Power and Light Company

Docket Nos. 50-387 and 50-388

    Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania.
    Date of application for amendments: May 31, 1994.
    Brief description of amendments: These amendments change the 
frequency for monitoring the Susquehanna site spray pond ground water 
level from once per month to once every 6 months.
    Date of issuance: October 20, 1994.
    Effective date: Both units; as of date of issuance and to be 
implemented within 30 days after the date of issuance.
    Amendment Nos.: 135 and 105.
    Facility Operating License Nos. NPF-14 and NPF-22. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34668).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 20, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

Philadelphia Electric Company, Public Service Electric and Gas Company 
Delmarva Power and Light Company, and Atlantic City Electric Company

Docket No. 50-277

    Peach Bottom Atomic Power Station, Unit No. 2, York County, 
Pennsylvania.
    Date of application for amendment: June 23, 1993, as supplemented 
by letters dated April 5, May 2, June 6, June 8, July 6 (two letters), 
July 7, July 20, July 28, 1994 (two letters), September 16, September 
30, and October 14, 1994. The supplemental letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    Brief description of amendment: The amendment raises the authorized 
maximum power level from 3293 MWt to a new limit of 3458 MWt.
    Date of issuance: October 18, 1994.
    Effective date: Unit 2, effective as of its date of issuance and is 
to be implemented prior to startup in Cycle 11 currently scheduled for 
October 28, 1994.
    Amendment No.: 198.
    Facility Operating License No. DPR-44: The amendment revised the 
license and Technical Specifications.
    Date of initial notice in Federal Register: August 29, 1994 (59 FR 
44432).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 18, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

Public Service Electric & Gas Company

Docket Nos. 50-272 and 50-311

    Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, 
New Jersey.
    Date of application for amendments: February 18, as supplemented by 
letter dated April 6, 1994 for Salem Unit 1 and March 28, 1994 for 
Salem Unit 2.
    Brief description of amendments: The change to Salem Unit 1 
Technical Specifications (TS) replaces the main feedwater control and 
control bypass valves with the main feedwater stop check valves for the 
Containment Isolation Function. The change to Salem Unit 2 TS adds a 
footnote to the 21-24 BF22 (main feedwater stop check valves) on Table 
3.6-1, ``Containment Isolation Valves.'' This note identifies those 
containment isolation Valves that are not subject to 10 CFR Part 50, 
Appendix J, Type C leakage testing.
    Date of issuance: October 20, 1994.
    Effective date: Units 1 and 2, effective as of date of issuance and 
shall be implemented within 60 days of the date of issuance.
    Amendment Nos.: 158 and 139.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37083).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 20, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

Southern California Edison Company, et al.

Docket Nos. 50-361 and 50-362

    San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego 
County, California
    Date of application for amendments: December 31, 1992.
    Brief description of amendments: These amendments revise the 
Technical Specifications (TS) to (1) distinguish between the core 
operating limit supervisory system (COLSS) in service and the COLSS out 
of service (OOS), (2) add surveillances to monitor departure from 
nucleate boiling ratio (DNBR) and/or linear heat rate (LHR) every 15 
minutes when the COLSS is OOS and the corresponding parameter is not 
being maintained as required, (3) increase the ACTION time from 1 hour 
to 4 hours when the COLSS is OOS and either the LHR or DNBR margin is 
not being maintained within the required limits, (4) change the power 
reduction requirements from ``HOT STANDBY'' to ``less than or equal to 
20 percent of Rated Thermal Power'' when the DNBR margin and/or the LHR 
limiting condition for operation (LCO) cannot be met within the allowed 
ACTION time, and (5) revise the Bases to the TS to reflect these 
changes.
    Date of issuance: October 27, 1994.
    Effective date: October 27, 1994.
    Amendment Nos.: 113 and 102.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 3, 1993 (58 FR 
12269).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 27, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713.

Tennessee Valley Authority

Docket Nos. 50-327 and 50-328

    Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee.
    Date of application for amendments: May 18, 1994; revised September 
9, 1994 (TS 94-05).
    Brief description of amendments: The amendments revise the action 
statement to provide a fixed duration that the control room emergency 
ventilation system may be inoperable due to actions taken as a result 
of a tornado warning.
    Date of issuance: October 17, 1994.
    Effective date: October 17, 1994.
    Amendment Nos.: 187 and 179.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32237).
    The Commission's related evaluation of the amendments are contained 
in a Safety Evaluation dated October 17, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority

Docket Nos. 50-327 and 50-328

    Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee.
    Date of application for amendments: August 19, 1994 (TS 93-09).
    Brief description of amendments: The amendments delay 
implementation of Amendments Nos. 182 and 174 from the Unit 2 Cycle 6 
refueling outage to as soon as acceptable plant conditions and 
modification activities/procedures are established in fiscal year 1995.
    Date of issuance: October 17, 1994.
    Effective date: October 17, 1994.
    Amendment Nos.: 188 and 180.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47182).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 17, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority

Docket Nos. 50-327 and 50-328

    Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee.
    Date of application for amendments: September 8, 1994 (TS 94-14).
    Brief description of amendments: The amendments incorporate 
clarifications regarding the evaluation of steam generator tube defects 
by separating the portion of the steam generator tube starting at the 
end of the tube up to the start of the tube-to-tube sheet weld from the 
remainder of the tube for the purposes of sample selection and repair 
when defects are found in this section of a steam generator tube.
    Date of issuance: October 20, 1994.
    Effective date: October 20, 1994.
    Amendment Nos.: 189 and 181.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: September 19, 1994. (59 
FR 47962).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 20, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company

Docket No. 50-346

    Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio.
    Date of application for amendment: April 5, 1994.
    Brief description of amendment: The amendment increases the 
surveillance test interval for the turbine-driven auxiliary feedwater 
pump and motor-driven feedwater pump from 31 days to 92 days; clarifies 
a requirement for a dedicated individual to be stationed at manual 
valves during surveillance testing because of the availability of the 
motor-driven feedwater system; addresses miscellaneous editorial 
corrections, and revises TS 3/4.7.1.2 and TS 3/4.1.7 and their 
associated bases.
    Date of issuance: October 18, 1994.
    Effective date: October 18, 1994.
    Amendment No.: 193.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27068).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 18, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

Union Electric Company

Docket No. 50-483

    Callaway Plant, Unit 1, Callaway County, Missouri.
    Date of application for amendment: February 10, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specification Table 2.2-1, ``Reactor Trip System Instrumentation Trip 
Setpoints,'' to correct Total Allowance values. The associated Bases 
section clarifies the relationship between the power supply and 
undervoltage relays.
    Date of issuance: October 27, 1994.
    Effective date: Date of issuance to be implemented within 30 days.
    Amendment No.: 93.
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14897).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 27, 1994.
    No Significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

Vermont Yankee Nuclear Power Corporation

Docket No. 50-271

    Vermont Yankee Nuclear Power Station, Vernon, Vermont.
    Date of application for amendment: December 6, 1993.
    Brief description of amendment: The proposed change removes the 
requirement to perform jet pump integrity and operability surveillances 
in the idle loop during operation with one recirculation loop.
    Date of issuance: October 26, 1994.
    Effective date: October 26, 1994.
    Amendment No.: 141.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29637).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 26, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301.

Virginia Electric and Power Company

Docket Nos. 50-280 and 50-281

    Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
    Date of application for amendments: October 19, 1993.
    Brief description of amendments: These amendments will add 
operability requirements, action statements, and surveillance 
requirements for the recirculation spray heat exchanger service water 
outlet radiation monitors. Also, surveillance requirements for several 
post-accident instruments are being reinstated.
    Date of issuance: October 27, 1994.
    Effective date: October 27, 1994.
    Amendment Nos.: 193 and 193.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67864).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 27, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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 December 9, 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-
[phone] (in Missouri 1-[phone]). 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).

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 20, 1994.
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Section 15.3.1.G, ``Operational Limitations,'' to 
reduce the reactor coolant system raw measured total flow rate and 
operating pressure, modify TS Section 15.2.3.1.B to increase the 
required reduction in the delta-T trip setpoint, and modify TS Figure 
15.2.1-1 to reflect new reactor core safety limits, all for Unit 2 
only. The applicable bases are also revised.
    Date of issuance: October 28, 1994.
    Effective date: October 28, 1994.
    Amendment Nos.: 156 and 160.
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
October 28, 1994.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Ernest L. Blake, Jr., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    Acting NRC Project Director: Cynthia A. Carpenter.

    Dated at Rockville, Maryland, this 2nd day of November, 1994.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 94-27613 Filed 11-8-94; 8:45 am]
BILLING CODE 7590-01-P