[Federal Register Volume 59, Number 128 (Wednesday, July 6, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-16174]


[[Page Unknown]]

[Federal Register: July 6, 1994]


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UNITED STATES NUCLEAR REGULATORY COMMISSION

 

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving 
No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from June 13, 1994, through June 23, 1994. The 
last biweekly notice was published on June 22, 1994 (59 FR 32226).

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, 11555 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 August 5, 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 John N. Hannon: 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.

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: June 2, 1994
    Description of amendment request: The proposed amendment would 
merge Toledo Edison Company into Cleveland Electric Illuminating 
Company. As described in the application, the company formed from the 
merger is intended to be renamed. Therefore, the licensee uses the 
nomenclature ``NEWCO'' as a temporary name of the combined operating 
company, and will provide the permanent name by supplemental letter. 
The amendment would (1) replace the Toledo Edison Company and Cleveland 
Electric Illuminating Company with ``NEWCO'' as a licensee, (2) 
designate ``NEWCO'' as the owner of the Perry Nuclear Power Plant, Unit 
1, and (3) make other administrative changes to the license as 
indicated in the amendment application. Centerior Service Company would 
be unaffected by the amendment and would remain a licensee.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to the Operating License are administrative 
and have no effect on any plant systems. All Limiting Conditions for 
Operation, Limiting Safety Systems Settings and Safety Limits 
specified in the Technical Specifications remain unchanged. This 
change meets one of the examples of a change not likely to involve a 
significant hazards consideration in that it is a purely 
administrative changes (48 FR 14864).
    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 to the Operating License are administrative 
and have no effect on any plant systems. All Limiting Conditions for 
Operation, Limiting Safety Systems Settings and Safety Limits 
specified in the Technical Specifications remain unchanged. This 
change meets one of the examples of a change not likely to involve a 
significant hazards consideration in that it is a purely 
administrative change (48 FR 14864).
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes to the Operating License are administrative 
and have no effect on any plant systems. All Limiting Conditions for 
Operation, Limiting Safety Systems Settings and Safety Limits 
specified in the Technical Specifications remain unchanged. This 
change meets one of the examples of a change not likely to involve a 
significant hazards consideration in that it is purely an 
administrative change (48 FR 14864).
    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: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John N. Hannon

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

    Date of amendment request: May 20, 1994
    Description of amendment request: The proposed amendment would 
permit the licensee to use an alternate repair criteria (ARC), 
designated as the F* criteria. Use of the F* criteria would 
allow tubes with otherwise pluggable indications, to remain in service 
as long as the indications are below the designated minimum distance of 
the F* criteria. The F* criteria defines a length of 1.7 
inches of undegraded expanded tube within the tubesheet as the minimum 
distance acceptable for implementing this ARC. Below the F* 
length, a circumferential tube defect can exist and the tube can remain 
in service. The proposed amendment will change the plugging limit 
definition and would exclude plugging steam generator tubes with 
indications that satisfy the F* criteria. The F* criteria 
maintains the structural integrity of the degraded tube as the primary 
pressure boundary and allows the tube to remain in service for heat 
transfer and core cooling.
    This alternate repair criteria qualification is documented in 
Babcock & Wilcox Nuclear Technologies (BWNT) Topical Report BAW-10196 P 
Revision 1, ``W-D4 F* Qualification Report'', which is included as 
part of the licensee's submittal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The supporting qualification report for the subject criteria 
demonstrates that the presence of the tubesheet will enhance the 
tube integrity in the region of the tube-to-tubesheet roll 
expansions by precluding tube deformation beyond its initial 
expanded outside diameter. The resistance to a tube rupture is 
strengthened by the presence of the tubesheet in that region. The 
results of hardrolling of the tube into the tubesheet provides a 
mechanical leak limiting seal between the tube and the tubesheet. A 
tube rupture cannot occur because the contact between the tube and 
the tubesheet does not permit sufficient movement of tube material.
    The type of degradation for which the F* criteria has been 
developed (cracking with a circumferential orientation) can 
theoretically lead to a postulated tube rupture event provided that 
the postulated through-wall circumferential crack exists near the 
top of the tubesheet. An evaluation including analysis and testing 
has been done to determine the resistive strength of the expanded 
tubes within the tubesheet. This evaluation provides the basis for 
the acceptance criteria for tube degradation subject to the F* 
criteria.
    The F* length of roll expansion is sufficient to preclude 
tube pullout from tube degradation located below the F* 
distance, regardless of the extent of the tube degradation. The 
Technical Specification leakage rate requirements and accident 
analysis assumptions remain unchanged in the unlikely event that 
significant leakage from this region does occur. The tube rupture 
and pullout is fully bounded by the existing steam generator tube 
rupture analysis included in the UFSAR. The leakage testing of the 
roll expanded tubes indicates that for tube expansion lengths 
approximately equal to the F* distance, any postulated primary 
to secondary leakage from F* tubes would be insignificant. The 
proposed alternate repair criteria does not adversely impact any 
other previously evaluated design basis accident.
    The leakage from an F* tube would be limited by the tube-
to-tubesheet interface since this leak would occur below the 
secondary face of the tubesheet. Qualification testing and previous 
experience indicate that normal and faulted leakage is well below 
Technical Specification and administrative limits creating no 
increase in the consequences associated with tube rupture type 
leakages. The UFSAR analyzed accident scenarios are still bounding 
since the normal and faulted leak rates are well within the normal 
operating limit of 150 gallons per day. This conclusion is 
consistent with previous F* programs approved and used at other 
operating plants.
    All of the design and operating characteristics of the steam 
generator and connected systems are preserved since the F* 
criteria utilizes the ``as rolled'' tube configuration that exists 
as part of the original steam generator design. The F* joint 
has been analyzed and tested for design, operating, and faulted 
condition loadings in accordance with Regulatory Guide 1.121 safety 
factors. The potential for tube rupture is not increased from the 
original submittal as demonstrated in the qualification analyses and 
testing completed in the BWNT report.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    B. The proposed changes do not create the possibility of a new 
or different type of accident from any accident previously 
evaluated.
    Implementation of the proposed F* criteria does not 
introduce any significant changes to the plant design basis. Use of 
the criteria does not provide a mechanism to initiate an accident 
outside of the region of the expanded portion of the tube. In the 
unlikely event the failed tube severed completely at a point below 
the F* region, the remaining F* joint would retain 
engagement in the tubesheet due to its length of expanded contact 
within the tubesheet bore. This engagement length would prevent any 
interaction of the severed tube with neighboring tubes. Any 
hypothetical accident as a result of any tube degradation in the 
expanded region of the tube would be bounded by the existing tube 
rupture accident analysis. Tube bundle structural integrity will be 
maintained. Tube bundle leak tightness will be maintained such that 
any postulated accident leakage from F* tubes will be 
negligible with regard to offsite doses.
    Therefore, there is not a potential for creating the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The use of the F* criteria has been demonstrated to 
maintain the integrity of the tube bundle commensurate with the 
requirements of Reg Guide 1.121 and the primary to secondary 
pressure boundary under normal and postulated accident conditions. 
Acceptable tube degradation for the F* criteria is any 
degradation indication in the tubesheet region, more than the 
F* distance from the secondary face of the tubesheet or the top 
of the last hardroll contact point which ever is further into the 
tubesheet. The safety factors used in the verification of the 
strength of the degraded tube are consistent with the safety factors 
in the ASME Boiler and Pressure Vessel Code and Reg Guide 1.121 used 
in steam generator design. The F* distance has been verified by 
various testing to be greater than the length of the roll expanded 
tube-to-tubesheet interface required to preclude both tube pullout 
and significant leakage during normal and postulated accident 
conditions. The protective boundaries of the steam generator 
continue to be maintained with the use of the F* criteria. A 
tube with the indication of degration previously requiring removal 
from service can be kept in service through the F* criteria. 
Since the joint is constrained within the tubesheet bore, there is 
no additional risk associated with the previously analyzed tube 
rupture event. The leak testing acceptance criteria are based on the 
primary to secondary leakage limit in the Technical Specifications 
and the leakage assumptions used in the UFSAR accident analyses.
    Implementation of the alternate repair criteria will decrease 
the number of tubes which must be taken out of service with tube 
plugs or repaired by sleeves. Both plugs and sleeves reduce the RCS 
flow margin; thus, implementation of the F* criteria will 
maintain the margin of flow that would otherwise be reduced in the 
event of increased plugging or sleeving.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the UFSAR or the Technical Specification 
Bases.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: For 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
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of amendment request: October 22, 1993
    Description of amendment request: The proposed amendment would 
modify the Reactor Trip System (RTS) and Engineered Safety Feature 
(ESF) instrumentation surveillance requirements to incorporate the 
applicable changes specified in NRC-approved WCAP-10271 and related 
supplements. Four specific changes were approved by the Nuclear 
Regulatory Commission for the RTS and Engineered Safety Feature 
Actuation System analog channels. These changes are limited to the 
specific Reactor Protection System (RPS) channels evaluated in the WCAP 
(including all supplements) and are subject to the conditions specified 
by the NRC.
    1. The surveillance or test frequency may be changed from monthly 
to quarterly.
    2. The time allowed for a channel to be inoperable or out of 
service in an untripped condition may be changed from 1 hour to 6 
hours.
    3. The time a channel in a functional group may be bypassed to 
perform testing may be increased from 2 to 4 hours. This bypass time 
applies to either an inoperable channel when testing is done in the 
tripped mode or to the channel in test when testing is done in the 
bypass mode. The Allowed Outage Time for maintenance of a channel is 12 
hours.
    4. Routine channel testing may be performed in the bypassed 
condition instead of the tripped condition.
    In addition, a number of editorial changes are made to improve 
clarity, and two-loop operating requirements are proposed to be 
deleted. Also, the surveillance test interval for RTS interlocks is 
proposed to be changed from monthly to once-per-refueling (about 18 
months). Although not part of WCAP-10271, this was previously approved 
generically by NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The determination that the results of the proposed changes are 
within all acceptable criteria was established in the SER(s) [Safety 
Evaluation Report(s)] prepared for WCAP-10271, WCAP-10271 Supplement 
1, WCAP-10271 Supplement 2 and WCAP-10271 Supplement 2, Revision 1 
issued by letters dated February 21, 1985, February 22, 1989 and 
April 30, 1990. Implementation of the proposed changes is expected 
to result in an acceptable increase in total Reactor Protection 
System yearly unavailability. This increase, which is primarily due 
to less frequent surveillance, results in an increase of similar 
magnitude in the probability of an Anticipated Transient Without 
Scram (ATWS) and in the probability of core melt resulting from an 
ATWS and also results in a small increase in Core Damage Frequency 
(CDF) due to Engineered Safety Features Actuation System 
unavailability.
    Implementation of the proposed changes is expected to result in 
a significant reduction in the probability of core melt from 
inadvertent reactor trips. This is a result of a reduction in the 
number of inadvertent reactor trips (0.5 fewer inadvertent reactor 
trips per unit per year) occurring during testing of RPS 
instrumentation. This reduction is primarily attributable to testing 
in bypass and less frequent surveillance.
    The reduction in [***] core melt frequency is sufficiently large 
to counter the increase in ATWS core melt probability resulting in 
an overall reduction in total core melt probability.
    The values determined by the WOG [Westinghouse Owners Group] and 
presented in the WCAP for the increase in CDF were verified by 
Brookhaven National Laboratory (BNL) as part of an audit and 
sensitivity analyses for the NRC Staff. Based on the small value of 
the increase compared to the range of uncertainty in the CDF, the 
increase is considered acceptable.
    Changes to Surveillance Test Frequencies for the Reactor Trip 
System Interlocks do not represent a significant reduction in 
testing. The currently specified test interval for interlock 
channels allows the surveillance requirement to be satisfied by 
verifying that the permissive logic is in its required state using 
the annunciator status light. The surveillance, as currently 
required, only verifies the status of the permissive logic and does 
not address verification of channel setpoint or operability. The 
setpoint verification and channel operability are verified after a 
refueling shutdown. The definition of the channel check includes 
comparison of the channel status with other channels for the same 
parameter. The requirement to routinely verify permissive status is 
a different consideration than the availability of trip or actuation 
channels which are required to change state on the occurrence of an 
event and for which the function availability is more dependent on 
the surveillance interval. The change in surveillance requirement to 
at least once every 18 months does not therefore represent a 
significant change in channel surveillance and does not involve a 
significant increase in unavailability of the Reactor Protection 
System.
    The proposed changes do not result in an increase in the 
severity or consequences of an accident previously evaluated.
    Implementation of the proposed changes affects the probability 
of failure of the RPS but does not alter the manner in which 
protection is afforded nor the manner in which limiting criteria are 
established.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not result in a change in the manner in 
which the Reactor Protection System provides plant protection. No 
change is being made which alters the functioning of the Reactor 
Protection System (other than in a test mode). Rather the likelihood 
or probability of the Reactor Protection System functioning properly 
is affected as described above. Therefore,the proposed changes do 
not create the possibility of a new or different kind of accident 
nor involve a reduction in a margin of safety as defined in the 
Safety Analysis Report.
    The proposed changes do not involve hardware changes except 
those necessary to implement testing in bypass. Some existing 
instrumentation is designed to be tested in bypass and current 
technical specifications allow testing in bypass. Testing in bypass 
is also recognized by IEEE [Institute of Electrical and Electronics 
Engineers] Standards. Therefore, testing in bypass has been 
previously approved and implementation of the proposed change for 
testing in bypass does not create the possibility of a new or 
different kind of accident from any previously evaluated. 
Furthermore, since the other proposed changes do not alter the 
functioning of the RPS, the possibility of a new or different kind 
of accident from any previously evaluated has not been created.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system setpoints or limiting conditions for 
operation are determined. The impact of reduced testing other than 
as addressed above is to allow a longer time interval over which 
instrument uncertainties (e.g., drift) may act. Experience has shown 
that the initial uncertainty assumptions are valid for reduced 
testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety by:
    a. Less frequent testing will result in fewer inadvertent 
reactor trips and fewer actuations of Engineered Safety Feature 
Actuation System components.
    b. Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation. This is due to less 
frequent distraction of the operator and shift supervisor to attend 
to instrumentation testing.
    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: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg, 
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., 
Washington, DC 20037.
    NRC Project Director: Walter R. Butler

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of amendment request: April 19, 1994
    Description of amendment request: The proposed amendment would 
modify Specifications 3.4.9.3 and 3.4.11 to incorporate power operated 
relief valve (PORV) Technical Specification (TS) changes in accordance 
with the guidance in Generic Letter 90-06 as implemented in NUREG-1431 
Improved Standard Technical Specifications (ISTS), with some exceptions 
and modifications to reflect plant specific design features. Certain 
other TS sections would also be modified to address related TSs.
    The proposed changes involve the details of (a) limiting conditions 
of operation, and (b) surveillance testing for equipment needed to 
protect the reactor vessel from overpressure conditions. This equipment 
includes PORVs and their associated block valves, charging pumps, 
reactor coolant system (RCS) vent, accumulators, and the overpressure 
protection system. Numerous administrative changes are also proposed, 
such as renumbering sections, spelling out mathematical symbols, 
changes in nomenclature for consistency, and relocating sentences and 
paragraphs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes consolidate the power operated relief valve 
requirements into Specifications 3.4.9.3 and 3.4.11 which generally 
adopt the new Improved Standard Technical Specifications of NUREG-
1431 to address the concerns identified in Generic Letter 90-06 
except for those changes required to reflect plant specific design 
features. These changes are proposed to enhance safety and improve 
the reliability of the PORVs and block valves. Since the proposed 
changes augment or preserve the requirements contained in the 
current technical specifications, we have concluded that these 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated in the UFSAR 
[Updated Final Safety Analysis Report].
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve any physical changes to the 
PORVs or their setpoints. These changes do not delete any function 
previously provided by the PORVs nor has the probability of 
inadvertent opening been increased. Accordingly, no new failure 
modes have been defined for any plant system or component important 
to safety nor has any new limiting single failure been identified as 
a result of these changes. Therefore, these changes will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated in the UFSAR.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes have been incorporated to enhance safety 
and improve the reliability of the PORVs and block valves to ensure 
their availability when called upon to perform their function. These 
changes do not affect the manner by which the facility is operated 
or involve a change to equipment or features which affect the 
operational characteristics of the facility. Therefore, operation of 
the facility in accordance with the proposed amendment would not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis. The staff notes 
that a significant effort has been made by the NRC and by industry over 
the last several years to improve and tighten the TS requirements for 
overpressure protection systems. These efforts are documented in 
Generic Letter 90-06 and NUREG-1431. The changes proposed by the 
licensee appear to result in TSs that are significantly more 
comprehensive and restrictive than those now existing for Beaver Valley 
Units 1 and 2, and, therefore, should help to reduce the probability of 
an accident. The staff disagrees with the licensee's claim that the 
changes do not affect the manner by which the facility is operated 
(consideration number 3 above). However, the staff believes that the 
proposed TS changes require more restrictive operation (such as more 
careful control of the number of charging pumps which can inject into 
the RCS) and do not involve a significant reduction in a margin of 
safety. Based on the NRC staff's 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: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg, 
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., 
Washington, DC 20037.
    NRC Project Director: Walter R. Butler

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of amendment request: June 9, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification section 4.8.1.1.2 to replace the current 
qualitative examination of new diesel generator fuel oil for water/
sediment and particulate contamination with a quantitative examination 
for the same properties.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
since the diesel generator availability and reliability is not being 
changed. The quantitative acceptance criteria for new fuel oil is 
not being changed. Diesel generator performance will therefore not 
be changed due to the proposed revision to SR [surveillance 
requirement] 4.8.1[.1].2.d.1.d. The diesel generator will continue 
to provide sufficient electrical power to ESF [engineered safety 
feature] systems. The ESF systems will continue to function, as 
assumed in the safety analyses, to ensure that the fuel, reactor 
coolant system, and containment
    design limits are not exceeded.
    Therefore, this changes will not increase the probability or 
consequences of an accident previously evaluated due to the 
continued availability and reliability of the emergency diesel power 
source.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not alter the method of operating the 
plant. This change will continue to ensure that the addition of new 
fuel oil complies with accepted standards regarding fuel oil 
quality. Since design requirements continue to be met and the 
integrity of the reactor coolant system pressure boundary is not 
challenged, no new failure mode has been created. As a result, an 
accident which is different than any already evaluated in the 
Updated Final Safety Analysis Report will not be created due to this 
change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety is not reduced because the emergency diesel 
generators will continue to provide sufficient capacity, capability, 
redundancy, and reliability to ensure availability of necessary 
power to ESF systems. The ESF systems will continue to function, as 
assumed in the safety analyses, to ensure that the fuel, reactor 
coolant systems, and containment design limits are not exceeded. The 
replacement of the clear and bright qualitative examination with the 
proposed quantitative test to determine the actual water/sediment 
and particulates will ensure that new fuel oil meets the required 
limits for these properties prior to addition to the storage tank, 
therefore assuring that the quality of the stored fuel is unaffected 
by the addition of new fuel.
    Therefore, this proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg, 
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., 
Washington, DC 20037.
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: February 16, 1994
    Description of amendment request: The proposed amendment would 
delete the Appendix B Section 4.2.2 requirement to perform infrared 
aerial photography every other year.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change will delete from Facility Operating License 
No. NPF-73 the Appendix B Section 4.2.2 requirement to perform 
infrared aerial photography every other year. The acceptance limit 
which forms the licensing basis for this technical specification is 
related to environmental impact and has no impact on the margin of 
safety, accident analysis, or other design basis impacting the 
margin of safety. No increase in adverse environmental impact has 
been identified over that previously identified in the Final 
Environmental Statement - Operating License Stage, environmental 
impact appraisals, or in any decisions of the Atomic Safety and 
Licensing Board. The Final Environmental Statement concluded, based 
on a model of combined drift from Units 1 and 2, that no adverse 
impacts to sensitive species of natural vegetation or to sensitive 
species of crops were expected. The staff also examined infrared 
aerial photographs taken from 1975 through 1983 and found no injury 
to vegetation from cooling tower drift in the vicinity of Unit 1.
    Continued terrestrial monitoring was performed for Beaver Valley 
Unit 2 by infrared aerial photography in 1986, 1988, 1990, and 1992. 
The results as provided in the Annual Environmental Reports Non-
Radiological concluded, ``Based on interpretation of the infrared 
photographs and field verification, there is no evidence to suggest 
that the BVPS [Beaver Valley Power Station] cooling towers are 
causing vegetation stress.''
    Based on the compilation of the infrared aerial photography 
performed for both BV-1 and BV-2, deletion of this terrestrial 
monitoring requirement will have no impact on the environment or the 
operation of the plant. Therefore, the proposed change will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The Appendix B Section 4.2.2 requirement to perform infrared 
aerial photography reflects a commitment described in Final 
Environmental Statement Section 5.14.1. Therein it is stated that 
the preoperational monitoring studies for BV-2 are based primarily 
on the BV-1 operational monitoring programs. ``Results of these 
studies have shown that there were no BV-1 operational impacts on 
flora, thus, the only terrestrial monitoring planned for BV-2 is 
continued infrared aerial photography every other year. The 
photographs will be compared with preoperational photographs of the 
BV-2 area, nd any signs of injury as a result of salt drift and 
other sources will be checked. The details of this terrestrial 
monitoring program will be specified in the Environmental Protection 
Plan that will be included in Appendix B of the operating license.'' 
The subject of this concern is the impact of salt and water drift on 
area vegetation including sensitive agricultural crops. From the 
standpoint of soil salinization (the effects of the accumulation of 
salts in the soil), described in the Environmental Report-Operating 
License Stage Section 5.3.3, no appreciable impact resulting from 
operation of the natural draft cooling towers is anticipated. This 
is based on the average rate of precipitation of 36.2 inches 
annually which greatly reduces the potential for accumulation of 
salt in the soil. The terrestrial monitoring program has been 
performed in accordance with the Environmental Protection Plan and 
has provided additional verification that operation of both cooling 
towers has not produced any evidence of vegetation stress. The 
proposed change does not introduce any new mode of plant operation 
or require any physical modification to the plant, therefore, this 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Infrared aerial photography surveillance does not affect safety 
systems and/or systems important to safety. Terrestrial monitoring 
is not used in any accident analysis and does not provide a basis 
for evaluating the radiological consequences of an accident. 
Deleting the requirement to perform infrared aerial photography will 
not result in any environmental impact from operation of the cooling 
tower and will not affect the operation of the cooling tower. The 
operating history of both the BV-1 and the BV-2 cooling towers has 
demonstrated that there is no evidence of vegetation stress in 
accordance with the results obtained from the infrared aerial 
photography and other associated methods of environmental 
monitoring. Therefore, based on the above, the proposed change will 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg, 
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., 
Washington, DC 20037.
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: May 23, 1994
    Description of amendment request: The amendment will revise 
Technical Specification (TS) 3.5.2 for Emergency Core Cooling Systems 
(ECCS) by removing the option that allows High Pressure Safety 
Injection (HPSI) Pump 1C to be used as an alternative to the preferred 
pump for subsystem operability. HPSI pump 1C is an installed spare 
which is not required to be maintained in an operable status, and this 
change is being requested to upgrade the ECCS operability requirements 
consistent with actual plant operating needs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed license amendment will remove the option of using 
High Pressure Safety Injection Pump 1C (HPSI-1C) to satisfy, in 
part, the Emergency Core Cooling System (ECCS) 
operabilityrequirements specified in Limiting Condition for 
Operation (LCO) 3.5.2. HPSI-1C is an installed spare pump that is 
not required to be operable unless it is being used in place of the 
preferred B-train ECCS high pressure pump. The required functional 
response of the ECCS or the required availability of the minimum 
equipment necessary to accomplish the ECCS safety function will not 
be changed by removing the spare pump option from the Technical 
Specifications.
    The calculated cooling performance of the St. Lucie Unit 1 ECCS 
during postulated accidents conforms to the criteria set forth in 10 
CFR 50.46 and the ability to achieve this required performance, 
including considerations of single-failure criteria, is independent 
from optional use of HPSI-1C. Therefore, operation of the facility 
in accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. Eliminating the 
option for the licensee to utilize HPSI-1C in place of the preferred 
B-train ECCS high pressure pump does not involve the addition of new 
or different types of equipment to the previously analyzed ECCS. 
Equipment important to safety will continue to perform their safety 
functions as previously analyzed and will not be affected by this 
proposed amendment. Therefore, operation of the facility in 
accordance with the proposed amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Removing the option to employ the installed spare HPSI pump 1C 
in lieu of the preferred B-train pump to determine ECCS operability 
only removes an operational flexibility that has rarely been used by 
the licensee. St. Lucie Unit 1 accident analyses do not take credit 
for an installed spare pump, the minimum complement of safety 
injection equipment required for safe operation of the facility and 
that is required by LCO 3.5.2 is not changed, and the results of 
plant accident and transient analyses are not influenced by this 
proposed amendment. The proposed change does not alter the bases for 
any Technical Specification related to the establishment of, or 
maintenance of, a nuclear safety margin. Therefore, operation of the 
facility in accordance with the proposed amendment would not involve 
a significant reduction in a margin of safety.
    Based on the discussion presented above and on the supporting 
Evaluation of Proposed TS Changes, FPL has concluded that this 
proposed license 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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: May 23, 1994
    Description of amendment request: The proposed amendments will 
relocate the seismic monitoring instrumentation Limiting Conditions for 
Operation, Surveillance Requirements, and the associated tables 
contained in Technical Specifications 3.3.3.3, 4.3.3.3.1 and 4.3.3.3.2 
to the Updated Final Safety Analysis Report. The basis for this request 
is consistent with NUREG-1432, ``Standard Technical Specifications, 
Combustion Engineering Plants'' and with the ``Final Policy Statement 
on Technical Specifications Improvements for Nuclear Power Reactors, 
``published in the Federal Register (58 FR 39132) dated July 22, 1993.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes are administrative in nature in that the 
specifications for operation and surveillance of the Seismic 
Monitoring Instrumentation system will be relocated from Appendix A 
of the facility operating license to the Updated Final Safety 
Analysis Report for St. Lucie Unit 1 and Unit 2. Changes to the 
system will be controlled by 10 CFR 50.59 and the safety analysis 
report is required to be updated pursuant to 10 CFR 50.71(e). 
Relocation of these requirements to the UFSAR is consistent with the 
NRC ``Final Policy Statement on Technical Specifications 
Improvements for Nuclear Power Reactors'' published in the Federal 
Register (58 FR 39132) dated July 22, 1993.
    Seismic monitoring instrumentation is not an accident initiator 
nor a part of the success path(s) which function to mitigate 
accidents evaluated in the plant safety analyses. The proposed 
technical specification change does not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident, nor do the changes 
alter any assumptions or conditions in any of the plant accident 
analyses. Therefore, operation of the facility in accordance with 
the proposed amendment would not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment to relocate the existing Technical 
Specification requirements for Seismic Monitoring Instrumentation to 
the Updated Final Safety Analysis Report will not change the 
physical plant or the modes of plant operation defined in the 
Facility License. The change does not involve the addition or 
modification of equipment nor does it alter the design or operation 
of plant systems. Therefore, operation of the facility in accordance 
with the proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed changes are administrative in nature in that 
operating and surveillance requirements for the Seismic Monitoring 
Instrumentation system will be relocated from Appendix A of the 
facility license to the Updated Final Safety Analysis Report for St. 
Lucie Unit 1 and Unit 2. Seismic monitoring instruments are not used 
to actuate safety-related equipment, provide interlocks, or 
otherwise perform plant control functions. The instruments are used 
to record the magnitude of a seismic event, should it occur. 
Conditions evaluated in plant accident and transient analyses do not 
involve seismic instruments. The proposed changes do not alter the 
basis for any technical specification that is related to the 
establishment of, or the maintenance of, a nuclear safety margin. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Based on the above discussion and the supporting Evaluation of 
Technical Specification changes, FPL has determined that the 
proposed license 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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

Florida Power and Light Company, et al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: May 23, 1994
    Description of amendment request: The proposed amendment revises 
Technical Specifications Section 3/4.7.1.1, Turbine Cycle, Safety 
Valves, to delete a specific reference to the 1974 edition of the ASME 
Code and refer to testing in accordance with Technical Specification 
4.0.5, the In-Service Inspection and In-Service Testing Specification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1)Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the main steam code safety valves will continue to 
be tested in accordance with current NRC requirements as implemented 
through 10 CFR 50.55a. The NRC specifies the ASME code requirements 
for a facility through revisions to 10 CFR 50.55a and through the 
review and approval of the plant specific in-service testing plan 
for pumps and valves at the beginning of each in-service inspection 
interval.
    The probability or consequences of an accident are not increased 
because testing of the main steam safety valves is in accordance 
with the appropriate NRC requirements.
    (2) Use of the modified specification would not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The use of this modified specification can not create the 
possibility of a new or different kind of accident from any 
previously evaluated since there is no physical change to the 
facility or the set points for the main steam safety valves. The 
valves will be tested in accordance with current requirements. No 
new failure mode is introduced due to the change because no plant 
change is being made and main steam safety valve test methods are 
consistent with the endorsed edition of the ASME Code.
    (3) Use of the modified specification would not involve a 
significant reduction in a margin of safety.
    The existing technical specification references an outdated 
version of the ASME Code. This change corrects the reference to 
Specification 4.0.5 which ensures that in-service testing of ASME 
Code Class 1, 2, and 3 pumps and valves will be performed in 
accordance with a periodically updated version of Section XI of the 
ASME Boiler and Pressure Vessel Code and Addenda as required by 10 
CFR 50.55a.
    Safety valve setpoints or tolerances are not changed by this 
proposal. Therefore, the modified specification corrects the ASME 
Code reference and does not involve a significant reduction in a 
margin of safety.
    Based on the above, we have determined that the proposed 
amendment does not (1) involve a significant increase in the 
probability or consequences of an accident previously evaluated, (2) 
create the probability of a new or different kind of accident from 
any previously evaluated, or (3) involve a significant reduction in 
a margin of safety; and therefore does not involve a significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: March 27, 1992, as supplemented on 
January 6, and May 27, 1994.
    Description of amendment request: The proposed amendment would 
revise the limiting conditions for operation and surveillance 
requirements for primary containment integrity, secondary containment 
integrity and other systems and equipment of Technical Specifications 
Section 3.7.
    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 because the requested revisions do not affect 
the FSAR safety analyses involving these system.
    Definitions
    The revisions to Definition 15, ``Primary Containment 
Integrity'' and Definition 16, ``Secondary Containment Integrity'' 
agree with the corresponding definitions of the STS. These changes 
are administrative in nature in that they only clarify the 
requirements for containment integrity and the appropriate means of 
isolating penetrations. These changes do not affect the operation or 
function of the containment isolation systems and, therefore, do not 
result in a significant increase in the probability or consequences 
of an accident previously evaluated.
    Primary Containment Integrity
    The revision to TS section 3.7.A, ``Primary Containment 
Integrity'', only adds a specific requirement to restore primary 
containment integrity within 1 hour or commence a plant shutdown. 
These actions are consistent with the actions specified in STS for 
primary containment integrity. No changes to the primary containment 
boundary or the requirements for primary containment integrity have 
been proposed. Therefore, this change does not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Primary Containment Power Operated Isolation Valves
    The revisions to TS section 3.7.B, ``Primary Containment Power 
Operated Isolation Valves'' are editorial in nature in that the 
wording has only been changed to be consistent with the STS 
requirements for primary containment isolation valves. These changes 
do not affect the function of the valves, the requirements to 
isolate a penetration with an inoperable containment isolation valve 
or the actual methods of isolation. Penetrations are still required 
to be isolated within 4 hours in a manner that cannot be adversely 
affected by a single active failure. Therefore, these changes do not 
result in a significant increase in the probability or consequences 
of an accident previously evaluated.
    Drywell Average Air Temperature
    The addition of limits, actions, and surveillance requirements 
for drywell average air temperature are intended to ensure that the 
initial assumptions in the DAEC Primary Containment Response 
Analysis to a DBA remain valid. The temperature limit (135 deg.F) 
corresponds to the initial drywell average temperature assumed for 
this analysis in the UFSAR. The specified limits, actions and 
surveillance requirements are consistent with STS. The addition of 
this limit to the TS will not affect the actual operation or 
function of any equipment but will ensure that the containment 
analysis remains valid. Therefore, the addition of this limit will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    Pressure Suppression Chamber-Reactor Building Vacuum Breakers
    The changes to TS section 3/4.7.D only provide additional detail 
and operability requirements for the pressure suppression chamber-
reactor building vacuum breakers. These additional details are 
consistent with the requirements of STS. Specifying separate 
operability requirements for vacuum breakers inoperable for opening 
(but known to be closed), or open better reflects the dual functions 
of these valves (vacuum relief and containment isolation). The 
additional surveillance requirement will better ensure that the 
containment isolation function of these valves is maintained. The 
rewording of existing surveillances only clarifies current 
requirements. These changes do not affect the actual function, 
setpoints, or number of valves required to be operable and therefore 
do not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    Drywell-Pressure Suppression Chamber Vacuum Breakers
    The changes to TS section 3/4.7.E, only provide additional 
detail and operability requirements for the drywell-pressure 
suppression chamber vacuum breakers. These additional details are 
consistent with the requirements of STS. Specifying separate 
operability requirements for vacuum breakers inoperable for opening 
(but known to be closed) or open better reflects the dual functions 
of these valves. The additional requirement to verify that each 
vacuum breaker is closed at least once per week will better ensure 
that the isolation boundary between the drywell and torus is 
maintained. The elimination of the requirement to exercise all 
operable drywell-pressure suppression chamber vacuum breakers upon 
determination that a vacuum breaker is inoperable for opening will 
not affect the reliability of these vacuum breakers. The only valid 
reason to exercise the operable vacuum breakers is if a common mode 
failure is suspected. We have reviewed the maintenance history of 
these valves and have not identified any instance of common mode 
failures. Conditional testing of these changes will not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Main Steam Isolation Valve Leakage Control System (MSIV-LCS)
    The change to TS Section 3/4.7.F, ``MSIV-LCS'', deletes the 
unnecessary and potentially non-conservative conditional 
surveillance testing of the redundant MSIV-LCS subsystems. Although 
the proposed change will reduce the amount of testing of the MSIV-
LCS, reliability of these systems would not be decreased and the 
necessary assurance that the alternate systems/subsystems/components 
will operate when needed is provided by the ASME Section XI IST 
Program.
    The possibility of human error will decrease with reduced 
testing. Human error such as a misalignment of valves after the 
system is returned to its normal configuration following testing and 
the misdirection of the operators attention from monitoring and 
directing plant operations is less likely to occur if this testing 
is eliminated. Additionally, reducing the scope and frequency of 
surveillance testing will decrease the probability of equipment 
failure (due to testing) which could require plant shutdown. 
Therefore, this change will not increase the probability of 
occurrence or consequences of an accident previously evaluated.
    Suppression Pool Level and Temperature
    The changes to TS section 3/4.7.F, ``Suppression Pool Level and 
Temperature'', are intended to clarify these requirements and make 
them more consistent with STS. The revision to the applicability 
statement which deletes the requirement for suppression pool level 
and temperature to be within the specified limits during work which 
has the potential to drain the vessel is in accordance with STS. 
Suppression pool level and temperature limits ensure that the 
suppression pool has the capability of acting as a heat sink for 
design basis events but are not appropriate or applicable during the 
refueling or cold shutdown conditions. No changes have been made to 
the actual suppression pool temperature or level limits and 
therefore, the assumptions made in the accident and transient 
analyses remain valid. These limits are consistent with the STS. The 
revisions to the surveillance requirements are also intended to 
improve clarity and consistency with STS. The deletion of the 
requirement to monitor suppression pool water temperature every 5 
minutes during relief valve operation is appropriate in that plant 
operating and emergency operating procedures already specify what 
actions are to be taken when suppression pool average water 
temperature increases above 95 deg.F including initiation of 
suppression pool cooling. Monitoring pool temperature every 5 
minutes during these events is not necessary and is redundant to 
other actions. Therefore, these changes will not significantly 
increase the probability of occurrence of the consequences of an 
accident previously evaluated.
    Containment Atmosphere Dilution
    The revisions to the applicability of TS section 3.7.H, 
``Containment Atmosphere Dilution'', requiring the containment 
atmosphere dilution system to be operable only when the reactor is 
in power operation and the primary containment is required to be 
inerted will not significantly increase the probability or 
consequences of an accident previously evaluated because the CAD 
system can only function when the containment is inerted. The 
function of the CAD system is to inject nitrogen into the 
containment after a LOCA and ensure the containment remains inerted. 
Drywell inspections performed after plant startup and prior to plant 
shutdown require that the primary containment be de-inerted for 
personnel access. Therefore, CAD system operability is not required 
during these inspections. No changes to the actual function or 
purpose of the CAD system are proposed.
    Oxygen Concentration
    The changes to TS section 3/4.7.1, ``Oxygen Concentration'' are 
administrative in that they only clarify the requirement that both 
the suppression chamber and the drywell must have oxygen 
concentrations less than 4 [percent] by volume. The revisions to the 
surveillance requirements are consistent with STS. Decreasing the 
frequency of verification of oxygen concentration from twice per 
week to once per week is in accordance with STS and reflects the 
fact that during power operation, the containment is inerted and 
slightly pressurized such that air (oxygen) cannot leak into the 
containment. Therefore, these changes will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    Secondary Containment
    The deletion of the requirement to operate the SGTS immediately 
after a secondary containment violation is identified will not 
affect the reliability of the secondary containment in that 
containment integrity is normally fully restored immediately after a 
violation is identified. The testing of the SGTS involves insertion 
of a Group III containment isolation signal and is only appropriate 
if the restoration of secondary containment involves a temporary or 
new secondary containment boundary. These modifications to a 
secondary containment boundary, however, would require that the SGTS 
be operated as part of post modification testing. Deleting the 
requirement for the SGTS to be operated after minor secondary 
containment violations will reduce the possibility of human error 
(such as misalignment of valves after the system is returned to its 
normal configuration) due to reduced testing. Operation of the SGTS 
after a secondary containment violation is not required by STS.
    Revision of the definition of calm wind conditions will not 
affect the reliability or availability of the secondary containment 
or SGTS. An engineering evaluation on the effects of wind speed and 
direction on the ability of the SGTS to maintain 1/4'' vacuum in 
secondary containment has been performed. The results indicate that 
while wind effects can be seen on individual instruments, there is 
minimal effect on the average instrument readings with wind speeds 
up to 15 mph. A discussion of this evaluation has been added to the 
Bases of TS section 3.7. Therefore, these changes will not 
significantly increase the probability or the consequences of an 
accident previously evaluated.
    Secondary Containment Automatic Isolation Dampers
    The addition of operability requirements, actions and 
surveillance requirements for secondary containment isolation 
dampers better ensures the integrity and isolation capability of the 
secondary containment. The new specifications are consistent with 
the requirements of the STS. The actual function or operation of the 
secondary containment isolation valves/dampers will not be affected. 
The appropriate valves/dampers will be incorporated in plant 
procedures that are subject to the change control provisions of TS. 
Therefore, these changes will not increase the probability of 
occurrence or consequences of an accident previously evaluated in 
the TS.
    Standby Gas Treatment System
    The change to the output requirements of the inlet heaters for 
each train of the SGTS from 11 kw to 22 kw better ensures that these 
heaters (and the SGTS) can perform their design function. The 22 kw 
output requirement ensures that the inlet air humidity does not 
exceed the 70 [percent] humidity specified in the UFSAR. This change 
does not affect the actual operation of the heaters or the SGTS.
    The requirement to demonstrate the HEPA filter uniform air 
distribution after HEPA filter replacement or after structural 
maintenance on the filter system housing (rather than annually) will 
not decrease the reliability of the SGTS. The air flow test will be 
performed after work or modifications which have the ability to 
disrupt the system geometry or result in potential flow blockage.
    Revising the shutdown LCO requirement in the various 
specifications from requiring the plant to be in Cold Shutdown in 24 
hours to requiring Hot Shutdown in 12 hours and Cold Shutdown (or 
other condition not requiring equipment operability) in the 
following 24 hours is consistent with STS and the shutdown 
requirements in TS section 3.5. This new requirement will allow the 
reactor to be shutdown in a more controlled manner and will not 
result in a significant increase in the probability or consequences 
of an accident previously evaluated.
    The revisions to the Bases are administrative in that they only 
reflect the changes to the individual specifications described 
previously in this section or correct minor discrepancies. All 
changes are consistent with the applicable specifications.
    (2) The proposed amendment will not increase the possibility of 
a new or different kind of accident from any accident previously 
evaluated for the following reasons.
    As described in the above response to question 1, none 
of the proposed changes alters the design of the plant or equipment 
or the plant's transient response. The changes to the definitions 
and limiting conditions for operation applicable to TS section 3.7 
are consistent with STS and better ensure that equipment assumed to 
be operable in our accident analysis will be operable upon demand. 
The addition of limiting conditions for operation for drywell 
average temperature and secondary containment isolation valves will 
better ensure that the assumptions in our accident analysis remain 
valid.
    The changes to the surveillance requirements are consistent with 
the STS. Those systems required to mitigate accidents evaluated in 
the UFSAR will still be operable and available.
    The reduction in conditional surveillance testing of certain 
systems and equipment will reduce the probability of equipment 
failure as a result of excessive testing or due to human error.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety for the following reasons.
    The revisions to the limiting conditions for operation in 
Chapter 3.7 of the TS will not invalidate the original licensing 
basis assumptions and will not invalidate any assumptions or input 
parameters for any DAEC event analysis. These changes provide more 
specific guidance only and are in accordance with the STS. Extending 
the time period within which the DAEC must achieve Cold Shutdown 
conditions will permit increased operator attention and minimal 
distractions for operators during shutdown, thus minimizing the 
risks of unexpected operational transients.
    Additional surveillance testing for certain instrumentation and 
systems will provide additional assurance that these systems will be 
available when needed.
    Elimination of unnecessary or conditional surveillance testing 
will not reduce the minimum necessary equipment operability 
requirements or equipment reliability. Elimination of the redundant 
testing will reduce equipment failure due to excessive testing or 
human error.
    In summary, the proposed administrative changes do not change 
the probability or consequences of an accident previously evaluated, 
do not create the possibility of a new or different kind of accident 
and do not involve a reduction in the margin of safety.
    Therefore, the proposed license amendment is judged to involve 
no significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036
    NRC Project Director: John N. Hannon

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

    Date of amendment request: May 27, 1974
    Description of amendment request: The amendment would temporarily 
allow the Operations Manager to not have a senior reactor operator 
(SRO) license for Millstone 3, providing other conditions are met.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:...The proposed change does not 
involve an SHC [significant hazards consideration] because the change 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change affects only an administrative control, 
which was based on the existing industry guidance in ANSI N18.1-
1971, that recommended the Operations Manager hold a SRO license. 
The current guidance in Section 4.2.2 of ANSI/ANS-3.1-1987 
recommends, as one option, that the Operations Manager have held a 
license for a similar unit and the Operations Middle Manager hold a 
SRO license. The Operations Middle Manager position does not exist 
at NNECO [Northeast Nuclear Energy Company]. Therefore, the proposed 
change requests an exception to ANSI N18.1-1971 to allow use of 
ANSI/ANS-3.101987 in a limited circumstance. Specifically, the 
proposed revision to Technical Specification 6.3.1 would temporarily 
allow the Operations Manager to have held a SRO at a PWR 
[pressurized water reactor] other than Millstone Unit No. 3. The 
proposed revision would be in effect for the period ending three 
years after the Staff's approval for this request.
    The proposed exception to ANSI N18.1-1971 will allow at least 
one of the Operations Assistants (instead of a Operations Middle 
Manager) to hold, and continue to hold, a SRO license, if the 
Operations Manager does not hold a license. The proposed change 
includes the requirement if the Operations Manager does not hold a 
SRO license at Millstone Unit No. 3, he shall have held a license 
for a similar unit in accordance with Section 4.2.2 of ANSI/ANS-3.1-
1987. For those areas of knowledge that require a SRO license, at 
least one of the Operations Assistants hold a SRO license and 
provides technical guidance normally required by the Operations 
Manager.
    The proposed change does not alter the design of any system, 
structure, or component. It does not change the way any plant 
systems are operated. It does not reduce the knowledge, 
qualifications, or skills of any operator on watch, and does not 
affect the way the Operations Department is managed by the 
Operations Manager in maintaining the effective performance of his 
personnel and to ensure the plant is operated safely and in 
accordance with the requirements of the Operating License.
    The proposed change does not detract from the Operations 
Manager's ability to perform his primary responsibilities. In this 
case, by having previously held a SRO license for a similar unit, he 
will have gained the necessary training, skills, and experience to 
fully understand the operation of plant equipment and the watch 
requirements for operators.
    The proposed change does not weaken the supervisory chain that 
presently exists in the Operations Department. All Control Room 
operators will continue to be supervised by the licensed Shift 
Supervisor.
    In summary, the proposed change does not affect the ability of 
the Operations Manager to provide the plant oversight required of 
his position. In addition, it does not have any [e]ffect on the 
probability or consequences of any previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed change to Technical Specification 6.3.1 does not 
affect the design or function of any plant system, structure, or 
component. It does not affect, in any way, the performance of NRC 
licensed operators, nor does it change the way any plant equipment 
is operated. Operation of the plant in conformance with technical 
specifications and other license requirements will continue to be 
supervised by personnel who hold an NRC SRO license. The proposed 
change to Technical Specification 6.3.1 ensures that the Operations 
Manager will be a knowledgeable and qualified individual. The 
proposed change does not introduce any new failure modes.
    3. Involve a significant reduction in a margin of safety.
    The proposed change involves only an administrative control 
which is not related to the margin of safety as defined in the 
technical specifications. The proposed change does not reduce the 
level of knowledge or experience required of an individual who fills 
the Operations Manager position, nor does it affect the conservative 
manner in which the plant is operated. All Control Room operators 
will continue to be supervised by personnel who hold a SRO license.
    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, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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: May 31, 1994
    Description of amendment request: This amendment will change the 
frequency for monitoring the Spray Pond ground water level from once 
per month to once every six months in Technnical Specification 
Requirement 4.7.1.3.c for each Susquehanna unit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed Technical Specification change to extend the 
monthly surveillance interval for Spray Pond ground water level 
measurement to biannual does not affect the the probability or 
consequences of an accident previously evaluated. The safety 
analysis performed for this change concludes that the ground water 
level is stable, predictable, and significantly below the acceptance 
criteria established in the Technical Specifications. Thus, less 
frequent monitoring of the ground water level does not increase the 
probability that the Spray Pond will become inoperable due to rising 
ground water levels or the probability of any accident scenarios 
associated with Spray Pond inoperability. The Technical 
Specification change will not impact the function or the method of 
operation of plant systems, structures, or components. Thus, the 
consequences of a malfunction of equipment important to safety 
previously evaluated in the FSAR are not increased by the change.
    II. This proposal does not create the possibility of a new or 
different kind of accident or from any accident previously 
evaluated.
    The proposed Technical Specification change to extend the 
monthly surveillance interval for Spray Pond ground water level 
measurement to biannual does not create the possibility of a new or 
different kind of accident or from any accident previously 
evaluated. The proposed change dies not affect systems, structures, 
or components (SSCs) or the operation of the SSCs; and therefore 
does not create the possibility of a new or different kind of 
accident.
    III. This change does not involve a significant reduction in a 
margin of safety.
    The proposed Technical Specification change to extend the 
monthly surveillance frequency to biannual does not reduce the 
margin of safety. The ground water level in the vicinity of the 
Spray Pond has been proven to be stable and predictable through 
twelve years of monthly data collection. This data has shown the 
highest ground water levels (still considerably lower than the 
Technical Specification limit) to occur in the months of April and 
October. Therefore, surveillance of the ground water level at the 
observation sites during April and October will adequately monitor 
this aspect of the Spray Pond operability.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: Charles L. Miller

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: May 25, 1994
    Description of amendment request: Salem is in the process of 
upgrading the Radiation Monitoring System. This upgrade will involve a 
replacement on many of the existing radiation monitors. The proposed 
change modifies Tables associated with Technical Specifications 3/
4.3.3.1 Radiation Monitoring Instrumentation, 3/4.3.3.8 Radioactive 
Liquid Effluent Monitoring Instrumentation, and 3/4.3.3.9 Radioactive 
Gaseous Effluent Monitoring Instrumentation. The proposed change 
relocates the Salem specific radiation monitor numbers from the table 
to a cross reference in the Bases. No required radiation monitoring 
functions are being changed.
    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. [This proposal does not involve] a significant increase in 
the probability or consequences of an accident previously analyzed.
    The proposed change to relocate the Salem specific radiation 
monitor numbers to the Bases is administrative. There are no 
modifications or changes in operating conditions associated with the 
proposed changes. Providing a more accurate description of a 
referenced radiation monitor in the note is editorial. The proposed 
changes do not affect the probability of occurrence or the 
consequences of accidents identified in the UFSAR. No accident 
precursors are being generated by the proposed changes. Therefore, 
the proposed changes do not involve a significant increase in the 
probability or consequences of a previously analyzed accident.
    2. [This proposal does not create] the possibility of a new or 
different kind of accident.
    The proposed changes to relocate the Salem specific radiation 
monitor numbers to the Bases are administrative. The change to the 
note on Table 3.3-6 is editorial to provide a more accurate 
description of the required radiation monitor. There are no 
modifications or changes in operating conditions associated with the 
proposed changes. Therefore, the proposed changes will not increase 
the possibility of a new or different kind of accident from any 
accident previously identified.
    3. [These changes do not involve] a significant reduction in a 
margin of safety.
    The Technical Specification operability requirements for the 
radiation monitors are not being changed. Relocating the Salem 
specific radiation monitor numbers to the Bases will not change any 
requirements for the radiation monitors. The change to the note on 
Table 3.3-6 is editorial to provide a more accurate description of 
the required radiation monitor. Therefore, the changes to the 
surveillance frequencies 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: Charles L. Miller

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

    Date of amendment request: May 23, 1994, as supplemented June 15, 
1994
    Description of amendment request: The proposed amendment would 
revise the Ginna Station Technical Specification (TS) 3.1.4 regarding 
allowable primary coolant levels of specific activity. The limit for I-
131 dose equivalent of iodine activity in the reactor coolant would be 
increased from 0.2 to 1.0 micro Ci/gm. The limit for total specific 
activity of the reactor coolant would be increased from 84 to 100/E 
micro Ci/gm, where E is the average beta and gamma energies per 
disintegration in Mev. Both increased allowable levels are consistent 
with NUREG-1431 ``Standard Technical Specifications, Westinghouse 
Plants, September 1992,'' and NUREG-0800 ``Standard Review Plan for the 
Review of Safety Analysis Reports for Nuclear Power Plants'' (Section 
15.6.3).
    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. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed changes do not affect any accident initiators and 
therefore the probability of any accident is not increased. 
Consequences of the changes are analyzed and shown acceptable in the 
[Westinghouse LOFTTR2 Analysis of Potential Radiological 
Consequences Following a Steam Generator Tube Rupture at the R.E. 
Ginna Nuclear Power Plant] analysis, WCAP-11668, Section III.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes involve no physical modifications to the 
plant; therefore, no new accident can be postulated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety, as no margin of safety is reduced by the proposed changes, 
as shown in WCAP-11668.
    Based upon the above information, it has been determined that 
the proposed changes to the Ginna Station Technical Specifications 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated, does not create 
the possibility of a new or different kind of accident previously 
evaluated, and does not involve a significant reduction in a margin 
of safety. Therefore, it is concluded that the proposed changes meet 
the requirements of 10 CFR 50.92(c) and do not involve a significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Project Director: Walter R. Butler

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: June 6, 1994
    Description of amendment request: The proposed amendment would 
merge Toledo Edison Company into Cleveland Electric Illuminating 
Company. As described in the application, the company formed from the 
merger is intended to be renamed. Therefore, the licensee uses the 
nomenclature ``NEWCO'' as a temporary name of the combined operating 
company, and will provide the permanent name by supplemental letter. 
The amendment would (1) replace the Toledo Edison Company and Cleveland 
Electric Illuminating Company with ``NEWCO'' as a licensee, (2) 
designate ``NEWCO'' as the owner of the Davis-Besse Nuclear Power 
Station, Unit 1, and (3) make other associated changes to the license 
as indicated in the amendment application. Centerior Service Company 
would be unaffected by the amendment and would remain a licensee.
    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, indicating that the proposed 
changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators or 
assumptions are affected. The proposed changes are administrative 
and have no direct affect on any plant systems. All Limiting 
Conditions for Operation, Limiting Safety System Settings and Safety 
Limits specified in the Technical Specifications will remain 
unchanged.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected. The proposed changes do not alter the 
source term, containment isolation, or allowable radiological 
consequences. The proposed changes are administrative and have no 
direct effect on any plant systems.
    2a. Not create the possibility of a new kind of accident from 
any accident previously evaluated because no new accident initiators 
are created. The proposed changes are administrative and have no 
direct effect on any plant systems. The changes do not affect the 
reactor coolant system pressure boundary and do not affect any 
system functional requirements, plant maintenance, or operability 
requirements.
    2b. Not create the possibility of a different kind of accident 
from any accident previously evaluated because no different accident 
initiators are created. The proposed changes are administrative and 
have no direct effect on any plant systems. The changes do not 
affect the reactor coolant system pressure boundary and do not 
affect any system functional requirements, plant maintenance, or 
operability requirements.
    3. Not involve a significant reduction in the margin of safety 
because the proposed changes do not involve new or significant 
changes to the initial conditions contributing to accident severity 
or consequences. The proposed changes are administrative and have no 
direct affect on any plant systems.
    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 Project Director: John N. Hannon

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

    Date of amendment request: May 20, 1994
    Description of amendment request: The proposed amendment would 
remove Core Spray (CS) High Sparger Pressure Instrumentation from the 
Vermont Yankee Technical Specifications for Emergency Core Cooling 
System (ECCS) Actuation Instrumentation. In addition, an unrelated 
administrative change is also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change to remove the Core Spray High Sparger 
Pressure Instrumentation from the Technical Specifications for ECCS 
Actuation Instrumentation is consistent with NRC requirements 
concerning this instrumentation. This instrumentation is considered 
NNS [nonnuclear safety] and performs a local monitoring and alarm 
function only. In addition, the NRC has recently approved the 
removal of Core Spray Sparger Break Detection Instrumentation from 
the Technical Specifications of another BWR [boiling water reactor] 
with a similar situation.
    The CS Sparger Piping is inspected every refueling outage to 
verify its integrity. No cracks in the CS Sparger piping have been 
identified since the first inspection in 1980. CS Sparger Piping 
integrity is still assured. The instrumentation systems to be 
removed from the ECCS Actuation Instrumentation Technical 
Specifications do not perform any automatic control or trip 
function. In addition, this instrumentation does not provide 
information that is required to permit the control room operator to 
take manual actions that are required for safety systems to 
accomplish their safety functions for design basis accident events.
    The proposed change does not result in any system hardware 
modification, function change or new plant configuration. The 
requested change to ECCS Actuation Instrumentation does not impact 
any FSAR [Final Safety Analysis Report] safety analysis involving 
the ECCS or Protection Systems. These monitoring functions are not 
contributors to the initiation of accidents.
    The administrative changes to correct typographical errors on 
Tables 3.2.1 and 4.2.1 will have no affect on plant hardware, plant 
design, safety limit setting or plant system operation and 
therefore, do not modify or add any initiating parameters that would 
significantly increase the probability or consequences of any 
previously analyzed accident.
    Therefore, it is concluded that there is not a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. The function of the Core Spray High Sparger Pressure 
Instrumentation to be removed from the Technical Specifications is 
for local indication and alarm only. These functions are not 
necessary for operators to accomplish any safety functions.
    The proposed change does not involve any change in hardware, 
function, Technical Specification trip setpoints, plant operation, 
redundancy, protective function or design basis of the plant. There 
is no impact on any existing safety analysis or safety design 
limits. Core Spray High Sparger Pressure Instrumentation functions 
do not initiate nuclear system parameter variations which are 
considered potential initiating causes of threats to the fuel and 
the nuclear system process barrier.
    As discussed above, the proposed administrative change only 
corrects typographical errors concerning equipment identification 
numbers. This change doe not affect any equipment and it does not 
involve any potential initiating events that would create any new or 
different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change to remove the Core Spray High Sparger 
Pressure Instrumentation from the Technical Specifications for ECCS 
Actuation Instrumentation does not affect any existing safety 
margins. This equipment is NNS and performs a local indication and 
alarming function only. The original intent of this detection system 
was because the first BWR plants had only the CS System for long-
term core cooling. Later, plants like VY [Vermont Yankee] were 
provided with Low Pressure Coolant Injection (LPCI) Systems in 
addition to CS.
    Existing Technical Specification requirements for automatic trip 
functions are unaffected. Failure of the Core Spray High Sparger 
Pressure Instrumentation does not preclude the ability of the CS 
System to perform its safety function to mitigate the consequences 
of accidents or of any other safety system to accomplish its safety 
functions. Proper ECCS functioning post-accident is not relied upon 
by NNS alarming functions but by such systems as safety related 
reactor level indication.
    The CS Sparger Piping is inspected every refueling outage to 
verify its integrity. No cracks in the CS Sparger piping have been 
identified since the first inspection in 1980. The removal from the 
Technical Specifications has no affect on the bases of Protective 
Instrumentation which is to operate to initiate required system 
protective actions. The Core Spray High Sparger Pressure 
Instrumentation does not perform any safety function.
    As discussed above, the proposed administrative change which 
corrects typographical errors does not affect any equipment involved 
in potential initiating events or safety limits. [***].
    Based upon the above, it is concluded that the proposed changes 
do not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301
    Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray, 
One International Place, Boston, Massachusetts 02110-2624
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: March 18, 1992, modifications submitted 
June 25 and July 28, 1992, and December 6, 1993
    Description of amendment request: On December 6, 1993, the licensee 
submitted a significant modification to its original request of March 
18, 1992 (April 19, 1992 (57 FR 12349)). The modified request would 
amend the Technical Specifications (TS) to provide two surveillance 
tests to determine the operability of the catalyst beds in the 
containment atmospheric control (CAC) system. One test compares 
hydrogen content in the influent to the hydrogen content in effluent 
process streams to assure the catalyst is operating. The second test 
measures the temperature profile in the catalyst bed to ensure that 
sufficient catalyst remains available for the recombination process 
during postulated accident conditions. Since the original proposal has 
been significantly changed, the staff is issuing a new notice and 
proposed no significant hazards consideration determination.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve any changes in the design 
or operation of the hydrogen recombiners, which are accident 
mitigation systems. The proposed change revises surveillance 
requirements to ensure that existing equipment will perform as 
designed in response to postulated events. The proposed change does 
not, therefore, involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve any changes in the design 
or operation of existing equipment, and does not, therefore, create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the amendment involve a significant reduction in a 
margin of safety?
    The proposed change increases surveillance requirements for 
existing equipment to ensure that the equipment will perform as 
designed. With equipment performing as designed in response to 
postulated accidents, the proposed change does not affect any 
existing margins of safety.
    Based on the licensee's analysis and the staff's analysis, 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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: Nicholas S. Reynolds, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Theodore R. Quay

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

    Date of amendment request: March 31, 1994
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant Technical Specifications (TS) 
by incorporating operability and surveillance requirements for the 
recently installed Auxiliary Feedwater Pump Low Discharge Pressure Trip 
instrumentation. Proposed surveillance requirements would be added to 
Table TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations and Test 
of Instrument Channels.'' TS 3.4, ``Steam and Power Conversions 
System,'' would be revised to explicitly link operability of the 
associated Auxiliary Feedwater Pump Low Discharge Pressure Trip channel 
to operability of the associated auxiliary feedwater pump. In addition, 
minor format inconsistencies in TS 3.4.b.1.A and 3.4.b.1.B would be 
corrected.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (a) Table TS 4.1-1
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change defines the necessary surveillance 
requirements for the recently installed Auxiliary Feedwater Pump Low 
Discharge Pressure Trip channels. The intent of adding surveillance 
requirements to the TS's is to ensure the availability and 
reliability of the components. The proposed change is an additional 
restriction not presently included in the TS's. Therefore, it will 
not increase the probability or consequences of an accident 
previously evaluated in the USAR.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change adds surveillance requirements to the TS for 
the Auxiliary Feedwater Low Discharge Pressure Trip channels. It 
does not alter the plant configuration or overall plant performance. 
Therefore, it does not create the possibility of a new or different 
kind of accident.
    3. Involve a significant reduction in the margin of safety.
    This proposed revision is an additional requirement in the TS's 
to ensure the availability and reliability of the Auxiliary 
Feedwater Pump Low Discharge Pressure Trip channels. It does not 
alter the input or assumptions of the safety analysis, and is an 
enhancement from an overall safety standpoint. Therefore, it will 
not involve a reduction in the margin of safety.
    (b) TS 3.4
    The proposed changes were reviewed in accordance with the 
provision of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change defines the necessary operability 
requirements for the recently installed Auxiliary Feedwater Pump Low 
Discharge Pressure Trip channels. Installation of this protection 
was recommended and approved by the NRC prior to their installation. 
The proposed change requires that the reactor not be heated 
 350 deg.F unless both motor driven Auxiliary Feedwater 
Pumps and their associated low discharge pressure trip channels are 
operable. Also, the reactor shall not be heated  
350 deg.F unless the turbine driven auxiliary feedwater pump and its 
associated low discharge pressure trip channel are operable, or if 
not demonstrated operable prior to  350  deg.F, they 
shall be declared inoperable when 350 deg.F is exceeded. 
Furthermore, when the reactor is  350 deg.F, an auxiliary 
feedwater pump low discharge pressure trip channel may be inoperable 
for a period not to exceed 4 hours. If this time is exceeded, the 
associated auxiliary feedwater pump shall be declared inoperable and 
the appropriate limiting condition for operation of TS 3.4.b.2 
entered. The intent of adding these operability requirements to the 
TS's is to ensure the availability of the components. The proposed 
change is an additional restriction not presently included in the 
TS's. Therefore, it will not increase the probability or 
consequences of an accident previously evaluated in the USAR.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change adds operability requirements to the TS for 
the Auxiliary Feedwater Pump Low Discharge Pressure Trip channels. 
It does not alter the plant configuration or overall plant 
performance. Therefore, it does not create the possibility of a new 
or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    This proposed revision is an additional requirement in the TS's 
to ensure the operability of the Auxiliary Feedwater Pump Low 
Discharge Pressure Trip channels. It does not alter the input or 
assumptions of the safety analysis, and is an enhancement from an 
overall safety standpoint. Therefore it will not involve a reduction 
in the margin of safety
    (c) Administrative changes to TS 3.4.b.1.A and TS 3.4.b.1.B
    The proposed changes were reviewed in accordance with the 
provision of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will 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 an accident previously evaluated, or
    3. Involve a significant reduction in the margin of safety.
    The proposed changes are administrative in nature and do not 
alter the intent of interpretation of the TS. Therefore, no 
significant hazards exist.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: John N. Hannon

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

    Date of amendment request: May 24, 1994
    Description of amendment request: The amendment request proposes to 
revise the Technical Specifications (TS) to implement the NRC's Final 
Policy Statement on Technical Specification Improvements for Nuclear 
Power Reactors (58 FR 39132). These improvements involve focusing the 
Technical Specifications on those requirements that are of controlling 
importance to operational safety by screening each TS in Section 3/4.1 
through 3/4.11 using the criteria provided in the policy statement. The 
purpose of the proposed amendment request is to relocate the 
specifications that do not meet any of the four policy statement 
criteria. The relocated specifications will be moved to Updated Final 
Safety Analysis (USAR) Chapter 16. Based on the screening, all or part 
of 38 technical specifications were identified as not meeting any of 
the criteria and, therefore, as candidates for relocation. The licensee 
has categorized the TS changes as (1) specifications relocated intact 
to USAR Chapter 16, (2) specifications relocated with portions retained 
in TS, (3) specifications relocated with programmatic requirements 
referenced in Section 6 of TS, (4) modifications to retained 
specifications to accomodate relocation of other specifications, and 
(5) new specification requirements incorporated into the TS. The last 
category is used to effect the retention of portions of relocated 
specifications and accomodate the policy statement recommendation to 
incorporate industry experience in the determination of TS content.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed Technical Specification changes involve relocating 
requirements that are not conditions or limitations on reactor 
operation necessary to obviate the possibility of an abnormal 
situation or event giving rise to an immediate threat to the public 
health and safety. The proposed changes were identified through the 
application of criteria designed to cull those requirements that are 
not important to operational safety from the Technical 
Specifications. In this process, selected provisions of the 
Technical Specifications identified for relocation were retained if 
necessary to support a Technical Specification that was to be 
retained. Thus, only specification requirements that have little or 
no operational safety significance are proposed for relocation. In 
addition, those requirements that would be relocated will be 
included in the Updated Final Safety Analysis Report (USAR) and, 
therefore, will be controlled and implemented as NRC commitments. In 
this manner, those requirements that have no operational safety 
significance but involve maintaining the plant in its as-designed 
state (for example, through surveillance programs) would be 
controlled.
    In addition, the criteria for identifying requirements to be 
retained in the Technical Specifications specifically call out, for 
retention, those structures, systems, or components that are 
required to mitigate accidents previously evaluated.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed changes involve relocating Technical Specification 
requirements to another licensee-controlled document. No changes or 
physical alterations of the plant are involved. Also, no changes to 
the operation of the plant or equipment are involved. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety.
    The proposed changes involve relocating Technical Specification 
requirements to the USAR. The requirements to be relocated were 
identified by applying the criteria endorsed in the Commission's 
Policy Statement. Thus, those specifications that would be relocated 
do not impose constraints on design and operation of the plant that 
are derived form the plant safety analysis report or from 
probabilistic safety assessment (PSA) information and do not belong 
in the Technical Specifications in accordance with 10 CFR 50.36 and 
the purpose of the Technical Specifications stated in the Policy 
Statement. Therefore, relocation of these requirements does not 
involve a significant reduction in the margin of safety.
    In addition, revisions to the USAR will be evaluated in 
accordance with the 10 CFR 50.59 process which considers the 
reduction in safety margin. Therefore, any future revisions to the 
provisions in the USAR will consider reductions in the margin of 
safety using the criteria for identifying an unreviewed safety 
question.
    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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Theodore T. Quay

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

    Date of amendment request: June 7, 1994
    Description of amendment request: The proposed amendment revises 
Technical Specification Table 2.2-1, Reactor Trip System 
Instrumentation Setpoints, to change the over-temperature-delta-
temperature (OTDT) axial flux difference (AFD) limits to reflect 
results of the Cycle 8 core maneuvering analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The probability of occurrence and the consequences of an 
accident evaluated previously in the Updated Safety Analysis Report 
(USAR) are not increased due to the proposed technical specification 
change. Operation at 3565 MWt does not affect any of the mechanisms 
postulated in the USAR to cause LOCA or non-LOCA design basis 
events. Analyses, evaluations and minimum DNBR [departure from 
nucleate boiling ratio] calculations confirm that the USAR 
conclusions remain valid for the proposed changes. On these bases it 
is concluded that the probability and consequences of the accidents 
previously evaluated in the USAR are not increased.
    2. The proposed change does not create the possiblity of a new 
or different kind of accident from any previously evaluated.
    There is no new type of accident or malfunction being created. 
The proposed change provides revised operating limits necessary to 
support Cycle 8, and does not change the method and manner of plant 
operation. The safety design bases in the USAR have not been 
altered. Thus, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed changes do not change the plant configuration in a 
way that introduces a new potential hazard to the plant and do not 
involve a significant reduction in the margin of safety. The 
analyses and evaluations discussed in the safety evaluation 
demonstrate that all applicable safety analysis acceptance criteria 
continue to be met for the proposed operating conditions. Items not 
specifically cited in this safety evaluation have been reviewed and 
have been found to be bounded by the evaluations performed for 
Reference 1 [Wolf Creek Generating Station Technical 
Specifications]. Therefore, it is concluded that the margin of 
safety, as described in the bases to any technical specification, is 
not reduced.
    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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Theodore. R. Quay

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.

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

    Date of amendment request: May 5, 1994, as supplemented June 16, 
1994.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications to increase Main Steam and 
Pressurizer Code Safety Valve Setpoint Tolerances.
    Date of publication of individual notice in Federal Register: June 
21, 1994 (59 FR 32029).
    Expiration date of individual notice: July 21, 1994
    Local Public Document Room location:
    Atkins Library, University of North Carolina, Charlotte (UNCC 
Station), North Carolina 28223.

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
rooms for the particular facilities involved.

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

    Date of application for amendments: September 29, 1992, as 
supplemented on October 22, 1993, and November 11, 1993.
    Brief description of amendments: The amendments revise the Site 
Boundary Map and the Low Population Zone Map.
    Date of issuance: June 22, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 190 and 167
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 28, 1992 (57 FR 
48813) The Commission prepared an Environmental Assessment and Finding 
of No Significant Impact which was published in the Federal Register on 
May 13, 1994 (59 FR 25129). The Commission's related evaluation of 
these amendments is contained in a Safety Evaluation dated June 22, 
1994.No significant hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station,Plymouth County, Massachusetts
    Date of application for amendment: February 11, 1993, as 
supplemented December 2, 1993, January 5, February 22, March 1, April 
15, and May 16, 1994.
    Brief description of amendment: This amendment increases the 
allowed fuel assembly storage cells from 2320 to 3859, changes the 
maximum loads allowed to travel over the spent fuel assemblies from 
1050 to 2000 lbs., and changes the limiting characteristics of 
assemblies to be stored in the spent fuel from a maximum KINIFITY 
less than or equal to 1.35 to a Maximum KINIFITY less than or 
equal to 1.32 and a maximum lattice average uranium enrichment of less 
than or equal to 4.6% by weight.
    Date of issuance: June 22, 1994
    Effective date: June 22, 1994
    Amendment No.: 155
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 30, 1993 (58 FR 
26171) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 22, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, IllinoisDocket 
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 
and 2, Rock Island County, IllinoisDate of application for 
amendments: March 26, 1993

    Brief description of amendments: The amendments revise Technical 
Specification 3/4.6 for Dresden and Quad Cities Stations to allow 
Single Loop Operation (SLO) with the recirculation loop suction and 
discharge valves open. The amendments also delete outdated and 
unnecessary portions of Technical Specification 3.6.H for Dresden Units 
2 and 3 and provide more consistency to the BWR Standard Technical 
Specifications (NUREG-0123, Revision 4).
    Date of issuance: June 16, 1994
    Effective date: June 16, 1994
    Amendment Nos.: 127, 121, 147, and 143
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17594) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 16, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: For Dresden, Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South CarolinaDate of 
application for amendments: March 24, 1994, as supplemented April 
11 and May 31, 1994

    Brief description of amendments: The amendments revise the 
Technical Specification (TS) to increase boron concentration for the 
spent fuel storage pool during Modes 1-3 operation and for the 
refueling canal during Mode 6 operation; include two reload related 
topical reports in TS 6.9.1.9; and correct errors in nomenclature and 
remove obsolete footnotes.
    Date of issuance: June 13, 1994
    Effective date: June 13, 1994
    Amendment Nos.: 120 and 114
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22006) The April 11 and May 31, 1994, letters provided clarifying and 
additional information that did not change the scope of the March 24, 
1994, application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated June 13, 1994.No 
significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
ConnecticutDate of application for amendment: December 17, 1993, as 
supplemented April 12, 1994.

    Brief description of amendment: The amendment changes the action 
statements for the limiting conditions for operation associated with 
the electrical power sources (Technical Specification 3.8.1.1).
    Date of issuance: June 14, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 177
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994(59 FR 
4943). The April 12, 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 June 14, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Power Authority of the State of New York, Docket No. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New YorkDate of 
application for amendment: December 28, 1993

    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 6.9(A)1.a. to permit startup reports for 
cycles subsequent to the initial fuel cycle to address only those 
startup tests that are actually performed. The amendment also revises 
TS Section 6.9(A) to clarify requirements for the submission of routine 
reports. These changes are consistent with the guidance provided in 
NUREG-1433, ``Standard Technical Specifications - General Electric 
Plants, BWR/4.''
    Date of issuance: June 16, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 212
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4945) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: December 29, 1993
    Brief description of amendment: The amendment revises Appendix B of 
the Technical Specifications (TSs), the Radiological Effluent TSs. 
Specifically, the amendment revises Appendix B Surveillance Requirement 
3.1.a. and Table 3.10-2 to provide surveillance requirements for data 
recorders associated with the gaseous effluent monitoring system. The 
amendment also makes an editorial change to Appendix B Limiting 
Condition for Operation 3.1.a. to improve consistency and clarity.
    Date of issuance: June 16, 1994Effective date: As of the date of 
issuance to be implemented within 30 days.
    Amendment No.: 213
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4946) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama.

    Date of amendments request: May 13, 1991, as supplemented October 
13, 1992.
    Brief description of amendments: The amendments modify the TS for 
the overpressure protection systems. The allowable outage time (AOT) 
for one inoperable residual heat removal (RHR) relief valve with one or 
more of the reactor coolant system cold leg temperatures less than or 
equal to 310 degrees Fahrenheit is being decreased from 7 days to 24 
hours for water-solid conditions. The required AOT for low temperature 
conditions, other than water-solid, will remain at 7 days with one RHR 
relief valve inoperable, provided the pressurizer level is less than or 
equal to 30 percent and a dedicated operator is assigned to monitor and 
control the reactor coolant system pressure.
    Date of issuance: June 16, 1994
    Effective date: June 16, 1994
    Amendment Nos.: 108 and 100
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: July 22, 1992 (57 FR 
32577) and February 17, 1993 (58 FR 8787)The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
June 16, 1994.No significant hazards consideration comments received: 
No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns 
Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama
    Date of application for amendment: April 1, 1992 (TS 302)
    Brief description of amendments: The amendments add requirements to 
the Browns Ferry Units 1 and 3 Technical Specifications to provide 
administrative controls for a post-accident sampling system, which were 
requested by Generic Letter 83-36, ``NUREG-0737 Technical 
Specifications.''
    Date of issuance: June 21, 1994
    Effective date: June 21, 1994
    Amendment Nos.:207 and 180
    Facility Operating License Nos. DPR-33 and DPR-68:
    Date of initial notice in Federal Register: May 27, 1992 (57 FR 
22269) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 21, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611

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: December 23, 1992, as 
supplemented on March 18, 1994.
    Brief description of amendment: This amendment revises TS 3/4 3.3.5 
for transfer switches used to meet 10 CFR Part 50, Appendix R (Fire 
Protection) requirements, and specifies a new special report 
requirement for TS 6.9.2.
    Date of issuance: June 14, 1994
    Effective date: June 14, 1994
    Amendment No. 187
    Facility Operating License No. NPF-3. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10016) The supplemental information submitted on March 18, 1994, did 
not change the initial proposed finding of no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated June 14, 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.

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

    Date of application for amendment: December 20, 1993, as amended 
March 25 and April 25, 1994
    Brief description of amendment: The amendment modifies the 
technical specifications (TS) to address new containment purge and vent 
valves to be installed in the 1994 refueling outage. The amendment 
changes the containment purge and vent valve TS as follows: (1) removes 
the requirement ensuring that valve position remains at less than or 
equal to 70 degrees, (2) changes the containment leak testing 
requirements for the metal-to-metal seated valves from 6 months to 2 
years since they have improved seat designs, and (3) makes 
administrative changes to delete an out-of-date note, to relocate an 
action statement requirement from the TS surveillance section to the TS 
action statement section, and to change a related TS reference to this 
surveillance section. Valve opening position does not need to be 
limited to less than or equal to 70 degrees. The resiliently-seated 
valves have a permanently installed mechanical stop to limit the open 
position to ensure adequate closure times. The metal-to-metal seated 
valves are designed to close from the 90-degree open position.
    Date of issuance: June 15, 1994
    Effective date: 15 days from the date of issuance
    Amendment No.:  124
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14901) The additional information contained in the March 25 and April 
25, 1994, letters was clarifying in nature, is within the scope of the 
initial notice, and did not affect the NRC staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated June 15, 1994.Public comments on proposed no significant hazards 
consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

    Date of application for amendment: February 8, 1994, as 
supplemented March 25, 1994
    Brief description of amendment: The amendment revises the WNP-2 
Technical Specifications. Specifically, the amendment increases the 
stroke time, as specified in Table 3.6.3-1, for reactor core isolation 
cooling (RCIC) valve RCIC-V-8 from 13 seconds to 26 seconds and deletes 
the Note (j) reference from RCIC-V-8 and RCIC-V-63. Note (j) indicates 
that the stroke time specified in the table reflects the requirement 
for containment isolation only.
    Date of issuance: June 17, 1994
    Effective date: June 17, 1994
    Amendment No.: 125
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24754) The additional information contained in the March 25, 1994, 
letter was clarifying in nature, was within the scope of the initial 
notice, and did not affect the NRC staff's proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
June 17, 1994.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352]
    Dated at Rockville, Maryland, this 28th day of June 1994.
    FOR THE NUCLEAR REGULATORY COMMISSION
Jack W. Roe,
Director, Division of Reactor Projects - III/IVOffice of Nuclear 
Reactor Regulation
[Doc 94-16174 Filed 7-5-94 8:45 am]
BILLING CODE 7590-01-F