[Federal Register Volume 59, Number 71 (Wednesday, April 13, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-8780]


[[Page Unknown]]

[Federal Register: April 13, 1994]


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UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice

 

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 21, 1994, through April 1, 1994. The 
last biweekly notice was published on March 30, 1994 (59 FR 14884).

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 P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, 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 May 13, 1994, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC 20555 and at the local 
public document room for the particular facility involved. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope
of the amendment under consideration. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
room for the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: February 18, 1994
    Description of amendment requests: The proposed amendment would 
modify Technical Specifications (TS) 5.3.1, Fuel Assemblies, and TS 
5.6.1, Criticality. In addition, the proposed Amendment would add a new 
Technical Specification 3/4.9.13, Boron Concentration-Storage Pool, and 
its associated BASES. This proposed amendment is requested to allow 
credit to be taken for burnup of spent fuel assemblies in establishing 
storage locations within the PVNGS spent fuel pools.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    Standard 1--Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Radiological consequences of the fuel handling accident are not 
impacted by the formation of new storage regions since the fuel 
assembly design is unchanged. However, even though the probability of 
occurrence of a fuel misplacement error has increased slightly, the 
consequences are markedly reduced by the crediting of 2150 ppm of 
soluble boron in the spent fuel storage pool. The increase is also not 
significant because of the types of administrative controls being put 
into place in Regions 2 and 3. Furthermore, a fuel assembly 
misplacement error is not considered an accident, as defined in the 
UFSAR.
    Standard 2--Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    This amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. No 
changes are being made to the fuel assemblies or the storage racks, and 
controls will be employed to control the placement of assemblies in 
Regions 2 and 3. As such, there is no possibility of a new or different 
kind of accident being created. The existing design basis covers all 
possible accident scenarios in the spent fuel storage pool.
    Standard 3--Involve a significant reduction in a margin of safety.
    This amendment request will not involve a significant reduction in 
a margin of safety. There is no reduction in the margin of safety since 
a keff less than or equal to 0.95 is met under all analyzed 
conditions using conservative assumptions which do not credit the 
soluble boron in the spent fuel storage pool except under some accident 
conditions, as allowed by NRC guidelines. The original mechanical 
analyses are unchanged for thermal and seismic/structural 
considerations, as these analyses were originally performed for a fully 
loaded spent fuel storage pool.
    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: Theodore R. Quay

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

    Date of application for amendment request: March 11, 1994
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3/4.7.D, ``Primary Containment Isolation 
Valves.'' The proposed amendments will add check valves installed in 
the reference leg instrumentation line. The valves have been installed 
as part of the modifications required to meet NRC Bulletin (IEB) 93-03.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposed license amendment adds the reference leg backfill 
check valves to the Technical Specifications. As such, the proposed 
amendment does not change the probability nor does it change the 
consequences of any previously evaluated accident for Dresden and 
Quad Cities Stations.
    The proposed modifications (and proposed Technical Specification 
amendments) add reference leg backfill instrument lines and check 
valves to the reactor vessel level instrumentation. The proposed 
modifications will eliminate the phenomenon described in IEB 93-03 
(dissolved gases in the [Reactor Vessel Instrumentation System] 
RVLIS piping may produce uncertainties in the level instrumentation 
during RPV depressurization) by providing degassed Control Rod Drive 
(CRD) water to the RVLIS reference leg piping. The proposed design 
ensures that a continuous column of water, free of non-condensible 
gases is maintained in the RVLIS reference leg piping. As such, the 
proposed modifications do not affect any accident precursors or 
initiators. Therefore the proposed modifications for the reference 
leg backfill instrument lines do not increase the probability of any 
previously evaluated accidents for Dresden Station and Quad Cities 
Station.
    The proposed plant modifications for the reference leg backfill 
check valves will not increase the radiological consequences of any 
previously evaluated accident. The radiological impact from a 
reference leg backfill instrument line break is bounded by Dresden's 
and Quad Cities' Instrument Line Break analysis (UFSAR Section 
15.6.2). Therefore, the proposed plant changes will not increase the 
consequences of any previously evaluated accident.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The proposed modification connects the non-safety-related CRD 
system to each safety-related division of RPV instrumentation and 
Feedwater Level Control System. The backfill check valves will 
eliminate the potential for reference leg leakage if CRD piping 
integrity is lost. These check valves are classified as safety-
related and will be maintained and controlled such that overall 
plant safety is maintained. The addition of the reference leg 
backfill check valves to the Technical Specifications does not 
create the possibility of a new or different kind of accident for 
Dresden Station or Quad Cities Station.
    (3) Involve a significant reduction in the margin of safety 
because:
    Primary containment integrity is not compromised by the addition 
of a pair of check valves that provide isolation for the reference 
leg backfill lines. These valves have been demonstrated to meet the 
intent of the criteria specified in General Design Criterion (GDC) 
55. The maintenance and control applied toward all the reference leg 
backfill check valves ensures that overall plant safety is 
maintained. Therefore, the addition of the reference leg backfill 
valves to Technical Specification 3.7.D.1 and 3.7.D.2 does not 
reduce the margin of safety for Dresden Station or Quad Cities 
Station.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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

    Date of application for amendment request: March 26, 1993
    Description of amendment request: The proposed amendments would 
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 would 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-0213, Revision 4).
    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:

    Commonwealth Edison has evaluated this proposed amendment and 
determined that it involves no significant hazards considerations. 
According to 10 CFR 50.92(c), a proposed amendment to an operating 
license involves no significant hazards considerations if 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 proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    The proposed change to delete the requirement to close the 
suction valve of the idle loop during SLO potentially affects two 
transient or accident analysis previously evaluated. The first is 
the Loss-of-Coolant Accident (LOCA) which has been analyzed for the 
full range of break sizes, from a small rupture, where the makeup 
flow is greater than the coolant loss rate, to the largest, a highly 
improbable circumferential recirculation line break. The design 
basis LOCA at Dresden and Quad Cities is the double-ended guillotine 
break in a recirculation line. LPCI is one of the Emergency Core 
Cooling Systems that would be initiated to flood the core following 
the Design Basis LOCA accident.
    The LOCA analysis for Quad Cities, Units 1 and 2, takes no 
credit for closure of the recirculation suction valve to properly 
direct LPCI flow into the lower plenum of the reactor. Instead, the 
LPCI loop selection logic is relied upon to automatically close the 
recirculation discharge valve of the selected intact loop. For 
Dresden, Units 2 and 3, LPCI is not credited to inject because the 
limiting failure is the LPCI injection valve. The LOCA ECCS analyses 
previously performed for SLO remain applicable and the severity of a 
postulated LOCA event has not increased. The proposed changes do not 
physically change the plant in any manner that would increase the 
probability of a LOCA.
    The second transient considered is the inadvertent startup of an 
idle recirculation pump in an unisolated loop. This event is 
precluded, however, when the loop is unisolated because the 
discharge valve must be closed for the pump to start. To further 
decrease the probability of the occurrence of this transient, 
Section 3.6.H.3.e is added to require the pump to be electrically 
prohibited from starting. In addition, leaving the loop unisolated 
results in an increase in the temperature of the water in the loop, 
and a correspondingly lower reactivity insertion should the 
transient occur. For these reasons, neither the probability nor the 
consequences of an inadvertent idle pump start have increased.
    The additional requirements introduced in Section 3.6.H.5.a-b 
and 4.6.H.5 to monitor temperatures between the two loops and the 
reactor coolant do not cause an increase in the probability or 
consequences of an accident because they limit stresses in the 
vessel and primary piping system to acceptable levels.
    The procedures that are currently followed at Dresden, Units 2 
and 3, regarding inadvertent entrance into a region of instability, 
defined as Region A, B, and C in Reference (e), are more 
conservative than those recommended by the NRC Bulletin, and do not 
allow operation in the stability regions defined in the Dresden 
Technical Specifications. Removing Sections 3.6.H.3.b-c and Section 
4.6.H.3 only removes outdated material from the Dresden Technical 
Specifications and does not increase the probability or consequences 
of an accident previously evaluated.
    The removal of Section 3.6.H.4 allowing operation without forced 
circulation below 25% of rated power at Dresden, Units 2 and 3, will 
not increase the probability or consequences of an accident 
previously evaluated. This change is conservative because it will 
prohibit operation in a condition susceptible to instabilities. This 
section also is not included in the Standard Technical 
Specifications. In the same manner, Section 2.1.A.4 is removed from 
the Quad Cities Technical Specifications.
    The removal of Section 3.6.H.3.a from the Dresden Technical 
Specifications will not increase the probability or consequences of 
an accident, because a one-pump run-up transient is bounded by the 
two-pump run-up transient.
    The change in initiation time for SLO requirements for Quad 
Cities from 12 hours to 24 hours does not represent a significant 
change, and still allows adequate time to implement the 
requirements. Therefore, no increase in the probability or 
consequences of an accident will be caused by this change.
    For the reasons stated above, no increase in the probability or 
consequences of an accident previously evaluated is introduced by 
the proposed changes.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because:
    The proposed change to eliminate the requirement to close the 
recirculation suction valve of the idle recirculation loop during 
SLO only removes unnecessary conservatism which is not required to 
ensure proper LPCI injection into the vessel during a LOCA. The LPCI 
loop selection logic already ensures that the intact loop's 
recirculation pump discharge valves will close when selected for 
LPCI injection. Since all ECCS functions will continue to perform as 
designed, no new accident scenarios are created. Also, by requiring 
the idle loop to be electrically prohibited from starting, the 
possibility of a new event is not created.
    Section 3.6.H.5.a-b and 4.6.H.5 provide for temperature 
monitoring prior to starting an idle pump, and monitoring to 
maintain acceptable primary system stress levels. Since the changes 
[do] not adversely affect the performance of any safety related 
systems, no new accident scenarios are created.
    The change eliminating the requirements for actions when a 
region of instability is entered during SLO will not create a new or 
different type of accident because procedures are already in place 
that are consistent with NRC guidance in this area. These procedures 
are more conservative than the current Technical Specifications.
    The removal of Section 3.6.H.4 (Dresden) and Section 2.1.A.4 
(Quad Cities) does not create the possibility of a new or different 
kind of accident because the units will not be allowed to operate 
without forced circulation with these sections removed, and Section 
3.6.H.4 added. The possibility of accidents occurring from operating 
in this mode has been eliminated, and no new types of accidents are 
created.
    There is no possibility of a new type of accident being created 
from the removal of Section 3.6.H.3.a from the Dresden Technical 
Specifications. The analysis behind the reduced flow MCPR curves 
provides more thermal margin during SLO than two-loop operation, 
because the one-pump run-up transient is less severe than the two-
pump run-up transient.
    The speed requirement change (Section 3.6.H.3.d for Dresden) for 
the operating recirculation pump prior to idle loop startup is in a 
conservative direction, and no new types of accidents are created.
    The increase in allowed time to initiate SLO requirements for 
Quad Cities (Section 2.1.A.4) does not represent a significant 
change, and still allows adequate time to implement necessary 
requirements. No new types of accidents are created by this change.
    The proposed changes do not involve a significant reduction in a 
margin of safety because:
    The change to eliminate the requirements to close the 
recirculation suction valves of the idle loop during SLO maintains 
the assumptions of the LOCA analyses. The LPCI loop selection logic 
will automatically close the recirculation pump discharge valve of 
the unbroken loop to ensure proper LPCI injection. Therefore, the 
current MAPLHGR limits at Dresden and Quad Cities will continue to 
ensure that Appendix K criteria are satisfied.
    During normal dual loop operation, LPCI loop selection logic is 
relied upon to close the discharge valve of the unbroken loop 
following a LOCA. This function is performed during SLO, provided 
the discharge valve and the logic that automatically closes this 
valve upon the occurrence of a LOCA signal remain operable. Since 
the assumptions of the accident analysis are preserved by the 
proposed change, there is no reduction in any safety margin.
    The safeguards in place preventing the inadvertent start of an 
idle recirculation pump are more than adequate protection against 
this transient. Three concurrent failures are required for this 
transient to occur. The transient would also be less severe due to 
the warmer water in the loop. Therefore, no reduction in a margin of 
safety will occur with this change.
    The addition of Sections 3.6.H.5.a-b and 4.6.H.5 will not 
decrease margin to safety, since the temperature monitoring 
requirements will maintain acceptable stresses in the primary system 
during idle pump starts.
    For Dresden, the current procedures for entrance into a region 
of stability provide more margin to safety than the current 
Technical Specifications require, because operation in a stability 
region is not allowed.
    The elimination of Section 3.6.H.4 in the Dresden Technical 
Specifications and Section 2.1.A.4 in the Quad Cities Technical 
Specifications will not decrease a margin of safety because it 
prohibits operations in a potentially unstable region. This change 
is in a conservative direction.
    The elimination of Section 3.6.H.3.a in the Dresden Technical 
Specifications does not cause a decrease in margin to safety, 
because there is more thermal margin to SLO than two-loop operation.
    Changing the active loop speed requirement from 65% to 43% prior 
to idle loop startup for Dresden is in the conservative direction; 
therefore, margin to safety is increased.
    The increase in allowed time to initiate SLO requirements for 
Quad Cities still provides adequate time to implement these 
requirements, and is not a significant change. Margin to safety is 
not decreased by this change.
    Margin of safety does not, therefore, decrease due to the 
proposed Technical Specification amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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

    Date of application for amendments: March 11, 1994
    Description of amendment requests: The proposed amendments would 
provide surveillance requirements for new hydraulic snubbers, which 
will be installed on the Main Steam Lines (MSLs) during the current 
Unit 1 refueling outage. This outage began on March 13, 1994, and it is 
scheduled to end on July 3, 1994. These snubbers will also be installed 
on Unit 2 during the Unit 2 refuel outage (Q2R13) currently scheduled 
for the first quarter of 1995.
    The amendment request would also change the Snubber Visual 
Inspection Intervals and Corrective Actions in Technical Specifications 
Sections 3.6.1 and 4.6.1 to the format and content of the BWR 
Standardized Technical Specifications (STS), as revised by the 
provisions of Generic Letter (GL) 84-13 ``Technical Specification for 
Snubbers'', dated May 3, 1984 and GL 90-09 ``Alternative Requirements 
for Snubber Visual Inspection Intervals and Corrective Actions'', dated 
December 11, 1990.
    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:

    Commonwealth Edison Company has evaluated the proposed Technical 
Specification Amendment and determined that it does not represent a 
significant hazards consideration. Based on the criteria for 
defining a significant hazards consideration established in 10 CFR 
50.92, operation of Quad Cities Station Units 1 and 2 (Quad Cities) 
in accordance with the proposed amendment will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposed changes adopt the format and content of the BWR-
STS, as modified by the provisions of GL 84-13 and GL 90-09. As 
such, these proposed changes are administrative in nature and have 
no effect on the accident analyses or system operation.
    The proposed schedule for snubber visual inspection intervals 
described in GL 90-09 will maintain the same level of confidence as 
the existing schedule as documented in Generic Letter 90-09, 
Alternative Requirements for Snubber Visual Inspection Intervals and 
Corrective Actions, dated December 11, 1990. Also, the surveillance 
requirement and schedule for snubber functional testing remains the 
same providing a 95 percent confidence level that 90 to 100 percent 
of the snubbers operate within the specified limits. The proposed 
visual inspection schedule is separate from functional testing and 
adds to the confidence level that the installed snubbers will serve 
their design function and are being maintained operable. Accident 
analyses assume that snubbers are initially operable. Compliance 
with the Technical Specification Surveillance Requirements for 
functional testing in conjunction with the revised visual inspection 
schedule assures continued operability of the snubbers. Therefore, 
no initial assumptions are being changed and thus neither the 
probability nor consequences of any accidents previously evaluated 
are significantly increased.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The proposed changes adopt the format and content of the BWR-
STS, as modified by the provisions of GL 84-13 and GL 90-09. As 
such, these proposed changes are administrative in nature and have 
no effect on the accident analyses or system operation.
    The proposed schedule for snubber visual inspection intervals 
will maintain the same level of confidence as the existing schedule 
as documented in Generic Letter 90-09, Alternative Requirements for 
Snubber Visual Inspection Intervals and Corrective Actions, dated 
December 11, 1990. Also, the surveillance requirement and schedule 
for snubber function testing remains the same providing a 95 percent 
confidence level that 90 to 100 percent of the snubbers operate 
within the specified limits. The proposed visual inspection schedule 
is separate from functional testing and adds to the confidence level 
that the installed snubbers will serve their design function and are 
being maintained operable. As a result, the supported piping, 
components, etc. will be maintained operable, so that supported 
safety systems will perform as designed. Therefore, the possibility 
of a new or different kind of accident is not created.
    (3) Involve a significant reduction in the margin of safety 
because:
    The proposed changes adopt the format and content of the BWR-
STS, as modified by the provisions of GL 84-13 and G[L] 90-09. As 
such, these proposed changes are administrative in nature and have 
no effect on the accident analyses or system operation. In addition, 
the proposed amendment maintains the same level of confidence as the 
current technical specification that snubbers are operable through 
the current snubber functional testing and the revised snubber 
visual inspection
schedule and the associated corrective action requirements. Therefore 
the proposed changes do not impact the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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

    Date of amendment request: February 25, 1994
    Description of amendment request: The proposed amendment will add a 
new Technical Specification 3/4.7.12, ``Ultimate Heat Sink.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed changes do not involve an SHC [significant hazards 
consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The ultimate heat sink (Connecticut River) provides the cooling 
water necessary to ensure the removal of the normal heat loads and 
normal cooldown loads of the plant and to mitigate the effects of 
accidents at the plant within acceptable limits. By placing a 
technical specification limit on the maximum temperature of the 
ultimate heat sink for plant operation, CYAPCO will assure that 
sufficient heat removal capacity is available. The intake structure 
draws water from the ultimate heat sink for circulation by the 
service water system and circulating water system. By adding this 
new requirement to the technical specifications, CYAPCO will ensure 
that the design basis, as stated in the Final Safety Analysis 
Report, for the ultimate heat sink is not violated.
    The change does not affect any initiating event. Thus, the 
change does not affect the probability of occurrence of any design 
basis accidents previously evaluated.
    There are no adverse impacts on the design basis accidents due 
to the addition of the ultimate heat sink temperature limitation. 
This administrative change has no effect on the consequences of the 
previously evaluated accidents.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Currently, the Haddam Neck Plant controls the ultimate heat sink 
temperature limit to less than 90 deg.F for Mode 1, 2, 3, and 4 via 
a plant procedure. The proposed change institutes a technical 
specification in place of this administrative control.
    As such, the administrative change is consistent with the 
current plant practice and has no effect on plant operation. Since 
there are no changes in the way the plant is operated, there is no 
possibility of an accident of a different type than previously 
evaluated due to the change.
    3. Involve a reduction in a margin of safety.
    The proposed change does not impact the physical protective 
boundaries, nor does it affect the performance of safety systems. 
There is no degradation in operability and surveillance requirements 
for the ultimate heat sink. Therefore, there will be no adverse 
impact on the margin of safety as defined in the basis for any 
technical specifications.

    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: Russell Library, 123 Broad 
Street, Middletown, Connecticut 06457.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
    NRC Project Director: John F. Stolz

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: January 28, 1994
    Description of amendment request: This amendment request is an 
additional followup to the amendment request of May 29, 1992, published 
in the Federal Register on July 8, 1992, (57 FR 30242) which changed 
the Technical Specifications Section 1.0, Definitions, to accommodate a 
24-month fuel cycle and which proposed the extension of the test 
intervals for specific surveillance tests. This amendment proposes 
extending the surveillance intervals to 24 months for leak testing 
containment isolation valves. The changes requested by the licensee are 
in accordance with Generic Letter 91-04, ``Changes in Technical 
Specification Intervals to Accommodate a 24-Month Fuel Cycle.'' In 
addition, the request corrects an administrative error.
    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. A significant increase in the probability or consequences of 
an accident previously evaluated will not occur.
    It is proposed that the interval between leakage tests of the 
containment isolation valves listed in the Technical Specifications 
be revised from 24 months to 24 months (+25%), consistent with an 
exemption request to 10 CFR [Part] 50 Appendix J for type C tests 
which requests an identical extension in the interval between tests.
    The proposed changes do not involve a significant increase in 
the probability or consequences of a previously analyzed accident. 
These changes propose extending the surveillance intervals for 
containment leakage testing. The changes do not involve any physical 
changes to the plant or alter the way equipment functions. Other 
system testing (e.g., on-line tests) provides assurance of system 
operability. An evaluation of past equipment performance provides 
additional assurance that the longer surveillance intervals will not 
degrade system performance. The 25% increase in the surveillance 
interval for type C leak rate testing is compensated for by a 
proportionate increase in the margin between specified leakage limit 
and the allowable leakage limit. Valves that are sealed with fluid 
are exempted from the 10 CFR [Part] 50, Appendix J leakage 
requirements. The Technical Specifications establish separate 
acceptance criteria for such cases base[d] on system design 
considerations. Additionally, in most cases, containment isolation 
valve redundancy (two valves in series) provides additional 
assurance that actual leakage would be lower than the test results 
would indicate.
    2. The possibility of a new or different kind of accident from 
any accident previously evaluated has not been created.
    The proposed license amendment does not create the possibility 
of a new or different kind of accident. These changes propose 
extending the surveillance intervals for containment leakage 
testing. The changes do not involve any physical changes to the 
plant or alter the way equipment functions. Other system testing 
(e.g., on-line tests) provides assurance of system operability. An 
evaluation of past equipment performance provides additional 
assurance that the longer surveillance intervals will not degrade 
system performance. The 25% increase in the surveillance interval 
for type C leak rate testing is compensated for by a proportionate 
increase in the margin between specified leakage limit and the 
allowable leakage limit. Also, containment isolation valve 
redundancy (two valves in series) provides additional assurance, in 
most cases, that leakage would be lower than the test results would 
indicate.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    The proposed change[s] does[do] not involve a significant 
reduction in the margin of safety. These changes propose extending 
the surveillance intervals for containment leakage testing. Other 
system testing (e.g., on-line tests) provides assurance of system 
operability. An evaluation of past equipment performance provides 
additional assurance that the longer surveillance intervals will not 
degrade system performance. The 25% increase in the surveillance 
interval for type C leak rate testing is compensated for by a 
proportionate increase in the margin between specified leakage limit 
and the allowable leakage limit. Also, containment isolation valve 
redundancy (two valves in series), in most cases, provides 
additional assurance that leakage would be lower than the test 
results would indicate.

    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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: November 11, 1993
    Description of amendment request: The amendments would change the 
Technical Specification (TS) surveillance requirements for the 
emergency core cooling system (ECCS) subsystems. Specifically, the 
changes would revise the minimum developed head requirement for the 
centrifugal charging pumps (CCPs), the safety injection pumps (SIPs), 
and the residual heat removal pumps (RHRPs); revise the sum of the 
minimum injection flowrates for the CCPs, SIPs, and the RHRPs; and 
revise the total maximum pump flowrate (runout limit) for the CCPs and 
the SIPs.
    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:

    (Amendment would not) involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The TS [technical specification] changes proposed by this 
amendment request are not considered to be initiators of any Design 
Basis Accidents (DBA). During normal operation the SIPs and the 
RHRPs are in standby, they are not operating. In the event of an 
accident resulting in an Engineered Safeguard (ES) actuation, the 
pumps would start to provide flow to the reactor vessel. The minor 
changes proposed for these pumps (SIPs and RHRPs) would not cause 
any accidents or events that have been previously evaluated.
    During normal operation, a CCP is operating. The proposed minor 
changes provided by this submittal only impact the performance of 
these pumps in response to an ES actuation. The proposed changes do 
not affect, in any way, how these pumps are operated during normal 
operation. As such, the minor changes proposed for the CCPs would 
not cause any accidents or events that have been previously 
evaluated. Accordingly, the proposed TS changes would not increase 
the probability of an accident that has been previously evaluated.
    The purpose of the ECCS subsystem is to ensure sufficient flow 
is provided to the core in the event of a LOCA [loss of coolant 
accident], that is to mitigate the consequences of a LOCA. A LOCA 
analysis was performed to determine the impact of the proposed TS 
changes. The analysis was performed in accordance with the NRC 
approved LOCA methodology for McGuire Nuclear Station. The results 
of the analysis demonstrate that the acceptance criteria of 10 CFR 
50.46 are still satisfied. Further, the purpose of the proposed TS 
changes are to prevent runout of the ECCS subsystem pumps during the 
injection and recirculation phases of a LOCA. Accordingly, the 
proposed TS changes would not increase the consequences of an 
accident that has been previously evaluated.
    (Amendment would not) create the possibility of a new or 
different kind of accident from any kind of accident previously 
evaluated.
    The proposed TS changes would not require any modifications to 
any structures, systems or components at McGuire Nuclear Station. 
Some minor changes to certain testing procedures for the ECCS 
subsystem pumps would be necessary. These minor changes would only 
involve specific values identified within the procedure and would 
not result in any changes on how the test would be performed. No 
other changes to procedures on how the station is operated or 
maintained would occur. Accordingly, the proposed TS change would 
not create a new or different kind of accident than what has been 
previously evaluated.
    (Amendment would not) involve a significant reduction in a 
margin of safety.
    The results of the analysis that was performed to determine the 
impact of the proposed TS changes would have in mitigating a LOCA 
indicate that the acceptance criteria of 10 CFR 50.46 are still 
satisfied. The analysis that was performed demonstrate that the Peak 
Clad Temperature (PCT) would remain below 2200 deg.F. The proposed 
changes ensure that the ECCS subsystem pumps will be operated within 
the limits specified by the manufacturer. Accordingly, the proposed 
TS changes would not significantly reduce any margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: David B. Matthews, Director

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

    Date of amendment request: February 25, 1994
    Description of amendment request: The amendments would add four 
instruments to the Technical Specification (TS) Tables 3.3-10 and 4.3-7 
as part of the accident monitoring instrumentation, and delete five 
instruments from the TS Tables that are not part of the accident 
monitoring instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    RESPONSE: No
    This proposed change does not involve any significant increase 
in the probability or consequences of any accident previously 
evaluated because no changes in the types, categories, hardwares and 
setpoints of the instruments involved were made; only the 
designation of which instruments should be listed in the T/S 
[technical specification] Tables and labeled as PAM [Post-Accident 
Monitoring] in the control room is changed through this proposed 
change.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    RESPONSE: No
    This proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because no changes in the types, categories, hardwares and setpoints 
of the instruments involved were made; only the designation of which 
instruments should be listed in the T/S Tables and labeled as PAM in 
the control room is changed through this proposed change.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    RESPONSE: No
    This proposed change does not involve a significant reduction in 
a margin of safety because no changes in the types, categories, 
hardwares and setpoints of the instruments involved were made; only 
the designation of which instruments should be listed in the T/S 
Tables and labeled as PAM in the control room is changed through 
this proposed change.

    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: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: David B. Matthews, Director

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: March 3, 1994
    Description of amendment request: This amendment removes 
restrictions from the Arkansas Nuclear One, Unit No. 1 (ANO-1) 
technical specifications (TSs) that prohibit use of the auxiliary 
building crane to move spent fuel shipping casks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The ANO procedures, load paths, crane equipment certification, 
operator training and other related heavy load handling topics were 
evaluated as part of the control of heavy loads issue and
found acceptable. Spent fuel cask handling is discussed in Section 
9.6.2.6 of the ANO-1 SAR [safety analysis report], which shows that the 
cask will never travel over spent fuel. ANO-1 SAR Section 9.6.2.6 
further evaluates the unlikely event of a cask drop accident and shows 
that the consequences are acceptable. Deletion of TS 3.8.15 to allow 
handling of a spent fuel shipping cask by the auxiliary building crane 
will have no actual impact on the cask drop or any other previously 
analyzed accident and therefore, does not involve a significant 
increase in the probability or consequences of any accident previously 
evaluated.
    Criterion 2--Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed amendment will allow handling of a spent fuel 
shipping cask by the auxiliary building crane where [sic] previously 
prohibited pending NRC evaluation of the spent fuel cask drop 
accident and crane design. The cask handling methods and cask drop 
accident are discussed and evaluated in ANO-1 SAR Section 9.6.2.6. 
Additionally, the NRC performed an independent evaluation of the 
radiological consequences of a cask drop accident, as documented in 
the ANO-1 SER [safety evaluation report] dated June 6, 1973. The 
evaluation of the unlikely event of a cask drop accident included 
assessment of equipment failures and has shown the consequences to 
be within acceptable bounds. Since no new accident scenarios can be 
identified related to the proposed amendment request, this change is 
bounded by the analysis described in the SAR and does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Criterion 3--Does Not Involve a Significant Reduction in the 
Margin of Safety.
    Although allowing use of the Auxiliary Building crane where 
[sic] previously prohibited by the TS could increase the possibility 
of a cask drop accident, the margin of safety is preserved in that 
the acceptable consequences of the cask drop accident evaluation in 
SAR Section 9.6.2.6 are not affected by this change. The proposed 
amendment request will not adversely affect the adequacy and 
conservatism of the cask drip accident evaluation. Therefore, the 
cask handling issue at ANO-1 continues to exhibit an acceptable 
margin of safety and does not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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: February 22, 1994
    Description of amendment request: The proposed amendments will 
relocate the instrument response time limits for the Reactor Protective 
System and Engineered Safety Features Actuation System from the 
Technical Specifications (TS) to the Updated Safety Analysis Report for 
both units. The proposed changes are line-item TS improvements and 
conform to the guidance given in Enclosures 1 and 2 of NRC Generic 
Letter 93-08.
    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:

    Pursuant to 10 CFR 50.92, a determination may be made that a 
proposed license amendment involves no significant hazards 
consideration if 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. Each standard is discussed as 
follows:
    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 amendments for St. Lucie Units 1 and 2 simply 
relocate tables of response time limits for instrumentation of the 
Reactor Protective System (RPS) and Engineered Safety Features 
Actuation System (ESFAS) from the Technical Specifications (TS) to 
the Updated Safety Analysis Report (UFSAR). The proposed amendments 
conform to the guidance given in Enclosures 1 and 2 of USNRC Generic 
Letter 93-08 (GL 93-08). Neither the response time limits nor the 
surveillance requirements for performing response time testing will 
be altered by this submittal. The overall RPS and ESFAS system 
functional capabilities will not be changed and assurance that 
actions of the protective and engineered safety features systems are 
completed within the time limits assumed in the accident analyses is 
unaffected by the proposed TS changes. 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 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 measurement of instrumentation response times at the 
frequencies specified in the TS provides assurance that actions 
associated with the protective and engineered safety features 
systems are accomplished within the time limits assumed in the St. 
Lucie Units 1 and 2 accident analyses. The response time limits, and 
the measurement frequencies remain unchanged by the proposed 
amendments. The proposed changes do not alter the basis for any 
other Technical Specification that is 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: February 28, 1994.
    Description of amendment request: The proposed amendments will 
delete the minimum frequency criteria prescribed for quality assurance 
audits from Administrative Controls sections 6.5.2.8 and 6.8.4 of the 
Technical Specifications (TS). Audit periodicity will thereby be 
controlled by the program described in the Florida Power and Light 
Company (FPL) Topical Quality Assurance Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10 CFR 50.92, a determination may be made that a 
proposed license amendment involves no significant hazards 
consideration if 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. Each standard is discussed as follows:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed amendment relocates administrative control criteria 
for minimum audit frequencies from the facility Technical 
Specifications to the FPL Quality Assurance (QA) Program. The QA 
Program is described in the FPL Topical Quality Assurance Report 
pursuant to 10CFR50, Appendix B. The change does not alter the bases 
upon which assurance is provided that safety-related activities are 
performed correctly nor does it involve the conditions and assumptions 
utilized in the analyses of plant transients and accidents. 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 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 amendment does not alter the bases for assurance that 
safety-related activities are performed correctly or that compliance 
with the required Limiting Conditions for Operation will be achieved. 
The change does not alter the basis for any Technical Specification 
that is 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.

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

    Date of amendment request: March 1, 1994.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.2.4, ``Quadrant Power Tilt 
Ratio,'' to add an exception to the requirements of TS 3.0.4. 
Specifically, to add ACTION statement ``d. The provisions of 
Specification 3.0.4 are not applicable.''
    Basis for proposed no significant hazards consideration 
determination: In 1992, the Vogtle TSs were amended in accordance with 
the recommendations of Generic Letter 87-09 to revise the wording of TS 
3.0.4 and to delete from TS 3.2.4 the statement that the provisions of 
TS 3.0.4 are not applicable. With the revised wording of TS 3.0.4, the 
statement of the non-applicability of the provisions of TS 3.0.4 was 
redundant for many individual specifications, and its deletion caused 
no change in ACTION requirements. However, in the case of TS 3.2.4, the 
deletion had the unintended effect of prohibiting power escalation 
above 50% rated thermal power (RTP) whenever the quadrant power tilt 
ratio (QPTR) exceeds 1.0.2. This unnecessarily delays power escalation. 
The proposed amendment would correct this error and restore the 
originally intended meaning of TS 3.2.4. The intent of TS 3.2.4 is to 
permit the escalation of reactor power above 50% RTP for limited times 
and
under specified conditions when the QPTR is greater than 1.02.
    With the original requirements of TS 3.2.4 restored, plant 
operation and power escalation during startup would be the same as 
previously approved. Therefore, the proposed change (1) does not 
involve a significant increase in the probability or consequences or an 
accident previously evaluated, (2) does not create the possibility of a 
new or different kind of accident than previously evaluated, and (3) 
does not involve a significant reduction in the margin of safety.
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. The proposed change to the Technical Specifications does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because it does not allow operation 
in a condition that is not already allowed by the Technical 
Specifications.
    2. The proposed change to the Technical Specifications does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated because it will not allow 
operation under conditions different from those already allowed by 
the technical specifications.
    3. The proposed addition to the Technical Specifications does 
not involve a significant reduction in the margin of safety because 
the action requirements will continue to be met in the same manner 
as currently required by the Technical Specifications.

    Accordingly, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308.
    NRC Project Director: David B. Matthews, Director.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of amendment request: February 7, 1994.
    Description of amendment request: The purpose of the request is to 
require the TMI-1 annual radioactive effluent release report for the 
previous calendar year be required to be submitted prior to May 1 of 
each year. Changing the TMI-1 due date to prior to May 1 can enable the 
licensee to combine the reports for TMI-1 and TMI-2 into a single 
report with a common due date.
    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. Submission of the annual radioactive effluent release report 
on or before April 30 each year in accordance with the change, as 
compared to the current requirement of March 1, does not involve a 
significant increase in the probability of occurrence or the 
consequences of an accident previously evaluated. The date on which 
the report is due has no impact on plant operations or effluents. It 
does not change the control of plant activities or the monitoring of 
plant effluents.
    2. Operation of TMI-1 in accordance with the proposed amendment 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed 
amendment has no impact on plant operations, plant effluents or the 
control of plant operations or effluents. Therefore, there is no 
potential to create a new or different kind of accident.
    3. Operation of TMI-1 in accordance with the proposed amendment 
does not involve a significant reduction in a margin of safety. 
There is no specified margin of safety in regards to the due date 
for the annual radioactive effluent release report.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of amendment request: March 2, 1994.
    Description of amendment request: The purposes of the request are 
to change the plant Technical Specifications (TS) to modify Operational 
Safety Instrumentation requirements to specify completion times which 
allows for performance of maintenance or surveillance within a 
reasonable time and to be consistent with the allowable outage time for 
other safety-related equipment when only one train is affected.
    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 of occurrence or the consequence of an accident 
previously evaluated.
    The proposed amendment permits time to restore instrumentation 
channels to operable status which is consistent with existing times 
allowed for outage of other safety-related equipment affecting one 
train. With regard to the 1 hour timeclock, this time is sufficient 
to perform the required action necessary to restore minimum required 
conditions. Allowing 6 hours to reduce reactor power in an orderly 
manner without challenging plant systems is reasonable, based on 
operating experience. Thus, the proposed amendment maintains an 
adequate degree of equipment availability without requiring 
unnecessary initiation of a plant shutdown for partial equipment 
outages.
    Therefore, it can be concluded that the proposed changes do 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 affects the [Reactor Protection System] 
RPS and the [Engineered Safeguards Actuation System] ESAS by 
providing timeclocks to perform corrective actions. During this 
timeclock period the safety function of the RPS can still be 
completed by the remaining minimum required channels. If an accident 
occurred while one ESAS train was inoperable due to faulty pressure 
switches or a faulty manual actuation channel, the redundant train 
would complete the safety function. The proposed time allowed for 
the pressure switches in one train or the faulty manual actuation 
channel to be out of service is bounded by the allowable time for 
other safety-related equipment such that only one train is affected. 
The proposed 8 hour timeclock associated with the [Reactor Building] 
RB purge radiation monitor, RMA-9, provides adequate time to confirm 
a problem exists and perform minor troubleshooting. The containment 
isolation function for the RB purge valves would still be maintained 
by a redundant 4 psig ESAS signal and a redundant reactor trip 
containment isolation signal.
    Therefore, the proposed amendment does not create the 
possibility of a new or different accident.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    As noted in the Bases for Specification 3.5.1, every reasonable 
effort will be made to maintain all safety instrumentation in 
operation. If RPS or ESAS instrumentation is found to be inoperable 
or require maintenance to assure reliability, the proposed amendment 
will allow the performance of maintenance and surveillance in a 
reasonable time period. The change does not result in a significant 
reduction in a margin of safety for the RPS because the automatic 
functions and various alternative manual trip methods are still 
available. Also, this change does not result in a significant 
reduction in a margin of safety for the ESAS because at least one 
train of safety features is required for continued operation within 
the specified timeclocks with automatic and manual trip functions.
    Thus, operation of the facility in accordance with the proposed 
amendment 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: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of amendment request: March 11, 1994.
    Description of amendment request: The purposes of the request are 
to (1) change the plant Technical Specifications (TS) to specify an 
allowable outage time for the Emergency Feedwater (EFW) Pumps during 
surveillance activities and (2) change the requirement to test 
redundant components for operability to a requirement to ensure 
operability based on verification of completion of appropriate 
surveillance activities. Basis for proposed no significant hazards 
consideration determination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated.
    The proposed amendment specifies an allowable outage time for 
testing of the EFW pumps. Also, this proposed change reflects the 
current NRC staff position regarding the need for additional testing 
to assure OPERABILITY. The allowable outage time this change 
provides for EFW pump testing is acceptable because the operator 
action required to make the motor driven EFW pump OPERABLE is 
minimal and can be performed in a very short time by the Control 
Room Operator who is continuously present during the time by the 
motor-driven EFW pump is in the Pull-To-Lock position.
    The changes affecting OPERABILITY determinations of redundant 
train/components for reactor building isolation valves and the 
control room air treatment systems reflect the current NRC staff 
position. Verifying that the required periodic surveillance testing 
is current and there are no known reasons to suggest the redundant 
train/component is inoperable, provides adequate assurance of system 
OPERABILITY.
    Therefore, it can be concluded that the proposed changes do 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 specifies an allowable outage time and 
deletes unnecessary redundant equipment testing. These changes do 
not change system operational requirements or response to system 
transients. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment specifies an allowable outage time and 
replaces redundant equipment testing with verification that 
surveillance is current as an adequate means to ensure OPERABILITY. 
These changes do not involve any activities associated with the 
margin of safety envelope. Thus, operation of the facility in 
accordance with the proposed amendment 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: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: March 16, 1994.
    Description of amendment request: The licensee proposes to amend 
the South Texas Project technical specifications (TS) by modifying TS 
3.4.9.3, ``Reactor Coolant System--Overpressure Protection Systems,'' 
Figure 3.4-4, ``Nominal Maximum Allowable PORV Setpoint for the Cold 
Overpressure System,'' for the cold overpressure mitigation system 
(COMS) with a revised setpoint curve. The proposed amendment would 
account for the pressure losses of the reactor coolant flow through the 
reactor core with either two or four reactor coolant pumps (RCP) 
operating. It was determined that the original COMS setpoint curve 
neglected reactor coolant pressure losses due to flow through the 
reactor core with the RCPs operating. The resulting pressure at the 
reactor vessel downcomer at the elevation equivalent to the core mid-
plane was higher than the pressure at the sensing point located in the 
residual heat removal system suction line connected to the reactor 
coolant system (RCS) hot leg. The proposed amendment would lower the 
power operated relief valve (PORV) setpoint limit by a quantity equal 
to the pressure difference between the pressure at the reactor vessel 
downcomer at the elevation of the core mid-plane and the pressure at 
the location of the residual heat removal system pressure transmitters.
    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 a previously evaluated 
accident.
    The presently existing pressurizer Power Operated Relief Valves 
(PORVs) setpoints, provided by the Cold Overpressure Mitigation 
System Curve (Figure 3.4-4) of Technical Specification 3.4.9.3, are 
nonconservative in that they do not account for reactor coolant 
pressure losses due to flow through the reactor core with Reactor 
Coolant Pump (RCPs) in operation. When RCP operation is considered, 
the pressure at the reactor vessel downcomer, at an elevation 
equivalent to core midplane, is higher than the pressure sensing 
point located in the Residual Heat Removal System suction line 
connection to the Reactor Coolant System (RCS) hot leg. Houston 
Lighting & Power Company became aware of this condition and re-
analyzed the Cold Overpressurization Event for the South Texas 
Project. The re-analysis has resulted in modifications to Figure 
3.4-4 of Technical Specification 3.4.9.3. The re-analysis reduced 
the PORVs setpoint to account for the pressure losses and provides 
for a setpoint for two Reactor Coolant Pump operation and for four 
Reactor Coolant Pump operation.
    The proposed decrease in the PORVs setpoint reduces the pressure 
versus temperature limit for the RCS under start-up and shut-down 
operations. The decreased PORV setpoints for post-overpressure 
incidents will ensure that RCS pressure will be maintained within 
acceptable limits during low temperature water solid operation for 
both two and four pump operation.
    The proposed change is based on a re-analysis which accounts for 
reactor coolant pressure losses through the reactor core. Reflecting 
actual reactor coolant pressure losses and adjusting the PORV 
setpoint as necessary has no adverse effect on the probability or 
consequences of an accident previously evaluated. Therefore, the 
proposed changes not only do not involve a significant increase in 
the probability or consequences of an accident previously evaluated, 
but actually maintain the original design basis.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed decrease in the PORV setpoints ensures that 
staggered operation of the two PORVs are maintained, thus minimizing 
the potential for large pressure undershoots resulting from multiple 
valve operation which may compromise the Reactor Coolant Pump No. 1 
Seal integrity. It also restricts the total number of discharge 
ports at any given moment to that absolutely necessary for pressure 
control. In addition, operation of either PORV provides the required 
design basis relief capacity and the required redundancy necessary 
to meet single failure criteria.
    The proposed change is the result of a re-analysis of a 
previously evaluated 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 does not involve a significant reduction 
in the margin of safety.
    The proposed change corrects an error present in the original 
analysis by accounting for reactor coolant pressure losses through 
the reactor core. The revised COMS curves are the result of a re-
analysis of the original COMS analysis. The new analysis preserves 
the originally intended margin of safety. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document location: Wharton County Junior College, J.M. 
Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW, Washington, DC 20036.
    NRC Project Director: Suzanne C. Black.

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

    Date of amendment request: March 21, 1994.
    Description of amendment request: The licensee proposes to revise 
Technical Specifications 3.1.2.3 ``Reactivity Control Systems Charging 
Pumps--Shutdown'' and 3.1.2.1 ``Boration Systems Flow Paths--
Shutdown.'' The amendment would allow energizing of an inoperable 
centrifugal charging pump in preparation for switching of the 
centrifugal charging pumps, provided the pump discharge is isolated 
from the reactor coolant system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    (1) The proposed change does not involve a significant increase in 
the probability or consequences of a previously evaluated accident.
    The proposed change is to modify the note which permits energizing 
of an inoperable centrifugal charging pump for testing purposes, 
provided the pump discharge is isolated from the reactor coolant 
system, to include pump energization for switching purposes.
    The proposed changes can potentially impact two events during Modes 
4, 5, and 6: (1) Cold overpressurization of the reactor coolant system, 
and (2) boron dilution resulting in a return to criticality. The 
requirements of Specification 3.1.2.3 with regard to the cold 
overpressure mitigation system analysis would remain valid because the 
inoperable centrifugal charging pump would be isolated from the reactor 
coolant system. A return to criticality would be prevented because the 
action statement of Specification 3.1.2.1 would be entered if the boron 
injection flow path could not be restored following centrifugal 
charging pump switching. Therefore, allowing energization of an 
inoperable pump for switching would have an insignificant effect on the 
probability of an overpressurization and boron dilution accident.
    Energization of an inoperable pump is currently permitted for 
testing purposes provided the pump discharge is isolated from the 
reactor coolant system. It is operationally desirable to maintain flow 
to the reactor coolant pump seals during the centrifugal charging pump 
switching process. This proposed change will not only protect the 
reactor coolant system from overpressurization at low temperatures, but 
will also provide the capability of maintaining reactor coolant pump 
seal injection flow during the switching process.
    Therefore, there is no increase in the probability or consequences 
of a previously evaluated accident.
    (2) The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Required boron injection flow paths would be maintained in Modes 4, 
5, and 6 except during centrifugal charging pump switching. In the 
event the boron injection flow path could not be restored after 
centrifugal charging pump switching, the action statement of 
Specification 3.1.2.1 would be entered. The proposed changes would not 
affect the operability of safety-related equipment and reactor coolant 
pump seal injection flow could be maintained. The plant operators are 
knowledgeable of the potential situation being created by energizing 
two centrifugal charging pumps and will follow direct administrative 
controls to isolate the pumps from the reactor coolant system. 
Therefore, the possibility of a new or different kind of accident is 
not created.
    (3) The proposed change does not involve a significant reduction in 
the margin of safety.
    Cold overpressure mitigating system requirements in Specification 
3.1.2.3 would continue to be maintained as a result of the proposed 
change. Thus, 10 CFR 50 Appendix G limits will not be affected. 
Although the boron injection flow path required by Specification 
3.1.2.1 may briefly be compromised, there is no significant reduction 
in a margin of safety because core alterations would be halted and 
positive reactivity changes would not be made if the boron injection 
path could not be maintained after centrifugal charging pump switching. 
This action, coupled with the short time period required for 
centrifugal charging pump switching, would preclude a return to 
criticality event. Therefore, there is no significant reduction in a 
margin of safety.
    Based on this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendments involves no significant hazards 
consideration.
    Local Public Document Room Location: Wharton County Junior College, 
J.M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488.
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW, Washington, DC 20036.
    NRC Project Director: Suzanne C. Black.

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

    Date of amendment requests: February 16, 1994 (Reference LAR 94-
05).
    Description of amendment request: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 to revise TS 3/4.7.2.1, ``Steam Generator 
Pressure/Temperature Limitation,'' 3/4.7.7.1, ``Snubbers,'' 3/4.7.8.1, 
``Sealed Source Contamination,'' 3/4.7.11, ``Area Temperature 
Monitoring,'' and 3/4.7.13, ``Flood Protection,'' in accordance with 
the Commission's Final Policy Statement on TS Improvements for Nuclear 
Power Reactors. These TS would be relocated to plant administrative 
controls and the final safety analysis report by reference.
    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. Do the changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    These proposed changes simplify the TS, meet regulatory 
requirements for relocated TS, and implement the recommendations of 
the Commission's Final Policy Statement on TS Improvements. Future 
changes to these requirements will be controlled by 10 CFR 50.59. 
The proposed changes are administrative in nature and do not involve 
any modifications to any plant equipment or affect plant operation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    b. Do the changes create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by which any safety-related system performs 
its function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    c. Do the changes involve a significant reduction in a margin of 
safety?
    The proposed changes do not alter the basic regulatory 
requirements and do not affect any safety analyses. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University,
Robert E. Kennedy Library, Government Documents and Maps Department, 
San Luis Obispo, California 93407.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: Theodore R. Quay.

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

    Date of amendment requests: February 16, 1994 (Reference LAR 94-
04).
    Description of amendment request: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 to revise TS 4.2.2, ``Heat Flux Hot 
Channel Factor--FQ(z),'' and 6.9.1.8, ``Core Operating Limits 
Report,'' to implement the revised methodology for calculating the 
penalty to FQ(z). The specific TS changes proposed are as follows:
    (1) The 2 percent FQ(z) penalty listed in TS 4.2.2.2.e.1) 
would be deleted and the statement revised to indicate the use of an 
appropriate factor to be specified in the Core Operating Limits Report 
(COLR).
    (2) TS 6.9.1.8.b.1. would be changed to reference Revision 1 of 
WCAP 10216-P-A, ``Relaxation of Constant Axial Offset Control 
FQ(z) Surveillance Technical Specification,'' dated February 1994.
    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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The Heat Flux Hot Channel Factor, FQ(z), is not involved in 
the initiation of any accident. Verifying FQ(z) is below its 
limit ensures initial conditions for accident analyses are met. The 
proposed changes have been previously approved by the NRC and 
provide for application of a more conservative FQ(z) penalty 
which will ensure that possible FQ(z) margin decreases are 
adequately accounted for. Therefore, if the FQ(z) does exceed 
its limit, the appropriate actions in TS 3.2.1 and TS 3.2.2. will be 
taken and are adequate to ensure design basis accidents analyses 
assumptions are met.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    FQ(z) is not involved in the initiation of any accident. 
The FQ(z) surveillance provides assurance that the initial 
conditions for accident assumptions are met. FQ(z) is a 
measurement of a physical property and is not involved in the 
initiation of any accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    c. Does the change involve a significant reduction in a margin 
of safety?
    The FQ(z) surveillance ensures that certain core parameters 
are maintained consistent with supporting assumptions regarding the 
core for postulated accidents. The methodology used in Revision 1 to 
WCAP-10216-P-A adequately accounts for FQ(z) increases between 
monthly flux maps. Using the methodology of Revision 1 to WCAP-
10216-P-A results in a FQ(z) penalty which is more conservative 
than the current TS FQ(z) penalty of 2 percent. If the 
FQ(z) increases above the TS limit, appropriate actions in TS 
3.2.1 and TS 3.2.2. are adequate to ensure design basis accidents 
analyses assumptions are met.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: Theodore R. Quay

Portland General Electric Company, et al., Docket No. 50-344, Trojan 
Nuclear Plant, Columbia County, Oregon

    Date of amendment request: January 27, 1993, revised March 8, 1994.
    Description of amendment request: The proposed amendment, as 
revised, by Portland General Electric Company, PGE or the licensee, 
would change the Trojan Nuclear Plant (Trojan) Appendix A Technical 
Specifications to reflect the permanently defueled status of the 
facility. The permanent cessation of power generation at Trojan and the 
May 5, 1993 amendment to the license which granted the licensee a 
Possession Only License for the facility has rendered many of the 
existing provisions of the current Appendix A Technical Specifications 
inappropriate. PGE has developed Permanently Defueled Technical 
Specifications (PDTS) for Trojan using NUREG-1431, ``Standard Technical 
Specifications, Westinghouse Plants,'' as a basis for the PDTS scope 
and format.
    Based on a series of discussions between the PGE staff and the NRC 
on February 9, 1994 and February 28, 1994 the licensee has revised 
several requirements contained in the original June 27, 1993 amendment 
request. These revisions were forwarded to the NRC staff by letter 
dated March 8, 1994. The March 8, 1994 revision updates the June 27, 
1993 submittal deleting reference to sections that had been relocated 
out of the Technical Specifications by amendments granted since June 
1993. It also provided supplemental information concerning the deletion 
and/or relocation of certain existing Trojan Technical Specifications 
requested by the NRC staff. The March 8, 1994 submittal also clarified 
the long term organization at the site, established a line of 
succession for the operational command and control function, modified 
the review and audit functions performed by the Independent Review and 
Audit Committee assuring their independence of review, required 
independent review of certain programs and manuals, limited annual and 
quarterly doses and operability and usage of the effluent treatment 
systems to conform to Appendix I to 10 CFR Part 50, continued the 
existing requirement for surveillance of former structural 
modifications to the facility, and required the submission of an annual 
radioactive effluent release report in accordance with 10 CFR 50.36a.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.92(a), the licensee has 
provided an analysis of the issue of no significant hazards 
consideration. In accordance with the requirements of 10 CFR 50.92, 
Issuance of Amendment, this license amendment request, as revised, is 
judged to involve no significant hazards consideration based upon the 
following:

    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 licensee analysis shows that the worst case design basis 
accident for this plant, in its permanently shutdown defueled state, 
is a fire in the radioactive waste annex building. The licensee has 
also identified a second design basis accident scenario, a fuel 
handling accident in the vicinity of the Trojan spent fuel pool. 
Other Trojan Final Safety Analysis Report (FSAR) accident scenarios 
addressed in Chapter 15 are no longer applicable to Trojan in the 
permanently defueled mode. The proposed amendment, as revised, does 
not lessen any of the requirements associated with either the 
radioactive waste annex building or the spent fuel pool therefore 
the probability of either accident occurring is unchanged. The 
proposed amendment, as revised, does not change the consequences of 
the accident since it does not affect the magnitude, detection, or 
mitigation of either accident scenario. Additionally, the ability of 
the radioactive waste annex building and the spent fuel pool to 
withstand other applicable FSAR events, natural phenomena, and fires 
is either unchanged from the existing licensing basis or is improved 
during the permanently defueled condition.
    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.
    Maintaining the permanently defueled facility in accordance with 
the PDTS, as revised by the March 8, 1994 letter, does not create 
the possibility of a new or different kind of accident from any 
previously considered. Most of the existing plant systems and 
functions will not be operational in the permanently defueled 
condition since power operations are prohibited and all of the fuel 
at Trojan is stored in the spent fuel pool. However,
all structures, systems and components that are necessary for safe fuel 
handling and storage activities will be maintained operable during the 
permanently defueled condition. The proposed PDTS, as revised, provide 
operation and surveillance requirements and administrative controls 
which are sufficient to ensure that the required structures, systems 
and components will be maintained operable in the permanently defueled 
condition.
    3. Operation of the facility in accordance with the proposed 
amendment does not involve a significant reduction in a margin of 
safety.
    The proposed PDTS, as revised by the March 8, 1994 letter, are 
sufficient to ensure no reduction in a margin of safety, in part, 
because of the reduced range of design basis accidents against which 
the facility must be protected now that the facility is prohibited 
from power operations and is permanently defueled. Only a fire in 
the radioactive waste storage facility or a fuel handling accident 
are relevant during the permanently defueled condition. The margins 
of safety for both of these accidents will remain the same or 
improve by maintaining the facility in accordance with the proposed 
PDTS, as revised. None of the other Chapter 15 FSAR accidents are 
applicable since power operations are prohibited and the facility is 
permanently defueled. Additionally, the margins of safety for other 
applicable FSAR events, natural phenomena, and fires are either 
unchanged from the existing licensing basis or is improved during 
the permanently defueled condition.

    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, as revised by the March 8, 1994 supplement, involves 
no significant hazards consideration.
    Local Public Document Room location: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensee: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRC Project Director: Seymour H. Weiss

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: October 14, 1993
    Description of amendments request: The proposed changes would 
revise surveillance test intervals and allowed outage times for reactor 
trip system (RTS) and engineered safety feature actuation system 
(ESFAS) instrumentation. The proposed changes would also revise certain 
RTS/ESFAS functions, minimum channels operable, channel calibration, 
and channel functional test requirements to ensure they are in concert 
with the Westinghouse Standard Technical Specifications and WCAP-10271, 
``Evaluation of Surveillance Frequencies and Out-of-Service Times for 
Reactor Protection Instrumentation Systems.''
    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 to the RTS/ESFAS STIs [surveillance 
test intervals] and AOTs [allowed outage times], and Minimum 
Channels Operable, Channel Calibration and Channel Functional Test 
requirements will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The determination that the result of the proposed changes 
associated with STI and AOT are within all acceptable criteria 
[that] has been established in the SERs [Safety Evaluations] 
prepared for WCAP-10271; WCAP-10271 Supplement 1; WCAP-10271 
Supplement 2; and WCAP-10271 Supplement 2, Revision 1. 
Implementation of the proposed changes results in a slight increase 
in the Reactor Trip System yearly unavailability. This slight 
increase, which is primarily due to less frequent surveillance, 
results in a slight increase in Core Damage Frequency (CDF) and 
public health risk. The values determined by the WOG and presented 
in the WCAP for the increase in CDR 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. Increasing STIs and AOTs is not expected to 
affect the probability or consequences of previously evaluated 
accidents.
    The change associated with the Minimum Channel Operable 
requirement for the RTS Turbine Trip by Turbine Throttle Valve 
Closure provides additional operating flexibility based on the 
Westinghouse Standard Technical Specifications, Revision 5. The new 
action statement ensures that any inoperable channel is placed in 
trip, and the remaining operable channels fulfill the necessary 
reactor trip diversity function. The change associated with the 
ESFAS Minimum Channel Operable requirement for Containment 
Pressure--High-High assures that the Technical Specifications 
reflect the correct as-built design actuation logic while ensuring 
the function continues to meet the single failure criteria. The 
proposed change to the turbine trip reactor trip function Channel 
Calibration reflects the assumptions in WCAP-10271 and current 
Farley calibration practices. The proposed change to the safety 
injection ESF [engineered safety feature] input for the reactor trip 
Channel Functional Test is consistent with the assumptions in WCAP-
10271 and current Farley surveillance testing practices. The 
proposed changes to the ESF permissive interlocks Channel 
Calibration and Channel Functional Test requirements are in concert 
with the NRC SER for WCAP-10271 and Farley surveillance practices. 
The proposed change to the ESF manual initiation functions Channel 
Functional Test reflects the proper surveillance requirements for a 
Westinghouse Solid State Protection System (SSPS). The proposed 
changes to these RTS/ESFAS Channel Calibration and Channel 
Functional Test requirements are also consistent with Westinghouse 
Standard Technical Specifications.
    (2) The proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed changes to not involve hardware changes and 
do not result in a change in the manner in which the RTS/ESFAS 
provides plant protection or the manner in which surveillance 
testing is perform to demonstrate operability. Therefore, a new or 
different kind of accident will not occur as a result of these 
changes.
    (3) The proposed changes do not 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. Evaluations have been performed to assure that the plant 
setpoints properly account for these instrument uncertainties over 
the longer time interval. RTS diversity is still provided by the 
Turbine Throttle Valve closure logic circuits. Steam Line Isolation 
diversity continues to be provided by the ESFAS Containment 
Pressure--High-High. Changes to certain RTS/ESFAS Channel 
Calibration and Channel Functional Test surveillances clarify what 
tests are required and when the tests are performed. Implementation 
of the proposed changes is expected to result in an overall 
improvement in safety as noted below.
    a. Less frequent testing will potentially result in fewer 
inadvertent reactor trips and ESF component actuation.
    b. Longer allowed outage times provide for better assessments of 
problems and easier repairs, ultimately resulting in better 
equipment performance.
    c. Less frequent distraction of the plant operator and shift 
supervisor to attend to and support instrumentation testing will 
improve the effectiveness of the operating staff in monitoring and 
controlling plant operation.

    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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: James H. Miller, III, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: S. Singh Bajwa

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 29, 1992, as supplemented on 
February 22, 1994
    Description of amendment request: The proposed amendment would 
modify the following Technical Specifications and their associated 
bases to permit longer allowable outage times (AOT) and increase 
surveillance testing intervals from monthly to quarterly: 3/4.3.1 
``Reactor Protection System Instrumentation;'' 3/4.3.2 ``Isolation 
Actuation Instrumentation;'' 3/4.3.3 ``Emergency Core Cooling System 
Actuation Instrumentation;'' 3/4.3.4. ``Recirculation
Pump Trip Actuation Instrumentation ATWS Recirculation Pump Trip System 
Instrumentation;'' 3/4.3.5 ``Reactor Core Isolation Cooling System 
Actuation Instrumentation;'' 3/4.3.6 ``Control Rod Block 
Instrumentation;'' 3/4.3.9 ``Plant Systems Actuation Instrumentation;'' 
and 3/4.4.2 ``Safety Valves.''
    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.
    The proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated because 
the proposed changes do not involve a change to the plant design or 
operation. The changes simply involve the frequency at which testing of 
the instrumentation is performed and the AOT for instruments. There may 
be a small increase in average instrument failure frequency as a result 
of increasing the surveillance interval. However, the proposed changes 
will require that a check be made for the majority of instruments to 
assure that making them inoperable for surveillance testing or repair 
does not result in a loss of function. This added check assures that a 
loss of function has not occurred, or if it has, that the appropriate 
ACTION statement be entered promptly. Therefore, these proposed changes 
do not result in a significant increase in either the probability or 
consequences of any accident previously analyzed.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed changes do not result in any change in the plant 
design or operation. The changes are for increased AOT and decreased 
frequency of instrumentation testing. The instrumentation involved are 
those instruments which sense plant problems and/or accidents, and then 
initiate systems or alarms to respond to the plant problem/accident. 
The proposed changes do not modify any of the instruments, or the 
initiation logic formed by the instruments. Therefore, no new or 
different type of an accident has been created.
    The proposed changes do not involve a significant reduction in a 
margin of safety because the small increase in average instrument 
failure frequency is offset by safety benefits such as a reduction in 
the number of inadvertent test-induced scrams, a reduction in wear due 
to excessive equipment test cycling, and better optimization of plant 
personnel resources. In addition, the proposed changes will require 
that a check be made for the majority of instruments to assure that 
making them inoperable for surveillance testing or repair does not 
result in a loss of function. This added check assures that a loss of 
function has not occurred, or if it has, that the appropriate ACTION 
statement be entered promptly. Therefore, these proposed changes do not 
result in a significant reduction in a margin of safety.
    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, D.C. 20037.
    NRC Project Director: John N. Hannon.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: November 15, 1993
    Brief description of amendments: The proposed amendment would 
revise the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 
technical specifications to increase the maximum permitted power at 
which the post-refueling power ascension reactor coolant system (RCS) 
flow verification can be performed.
    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 a previously evaluated accident.
    The proposed change increases the power at which the 
verification of the Reactor Coolant System flow rate with a 
precision heat balance can be performed. The only potentially 
relevant concern for this change is the possibility of having 
insufficient flow to support the accident analyses at the higher 
(85%) power level. Power level and RCS flow are important parameters 
in determining the severity of an event but have no impact on the 
initiation of an event or accident. Thus, the change does not 
involve a significant increase in the probability of any previously 
analyzed accident.
    Although accidents tend to be more severe at higher initial 
power levels, the acceptance criteria of the applicable safety 
analyses continue to be met [even when the test is conducted at the 
higher power level]. Thus, the proposed change would not involve an 
increase in the consequences of any previously analyzed accident.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change merely increases the power at which the 
initial post-refueling startup verification of RCS flow with a 
precision heat balance may be performed. Since the new power level 
is within the normal operating range of the reactor, it does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed change increases the power which can be attained 
prior to verification of the Reactor Coolant System flow with a 
precision heat balance. The power value is increased from 75% RTP 
[rated thermal power] to 85% RTP. An analysis was performed which 
demonstrates that the safety analysis DNB [departure from nucleate 
boiling] limit will not be exceeded if the initial power is 85% or 
less, even with a significant reduction in flow. A flow reduction 
significantly different from the expected flow is highly unlikely, 
however, since RCS flow is verified by measurement of elbow tap 
differential pressure prior to operation in Mode 1. This flow 
measurement, although less accurate than the precision heat balance, 
is sufficient to assure adequate flow at 85% power. Adequate 
limitations on power level, F delta-H verification, and the power 
range neutron flux-high setpoint are imposed during post-refueling 
power ascension to ensure that if the RCS flow is not verified with 
a precision heat balance until 85% RTP, the results of the accident 
analyses would remain valid.
    The evaluation of the DNB-limited events, initiated from a power 
level of 85% RTP, considered the limitations imposed during the 
power ascension. Based on this evaluation, it is concluded that, 
even though 85% RTP is a more severe initial condition, the 
applicable event acceptance criteria would continue to be met; 
therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
    NRC Project Director: Suzanne C. Black

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: March 30, 1994
    Brief description of amendments: The proposed amendments would 
revise the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 
Technical Specifications to allow the use of an alternative method for 
verifying that the emergency diesel generator fuel oil meets 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of a previously evaluated accident.
    Either the current testing option, which remains valid for fuel 
oil with ASTM Color 5 or less, or the alternative testing method, 
provide the necessary assurance that the water and sediment quantity 
in the new fuel oil is acceptable. As the performance of the 
quantitative water and sediment test of ASTM-D1796-1968, maintains 
essentially the same attribute qualities of the new fuel oil, there 
should not be any undetected degradation in the Diesel Generator 
fuel oil supply.
    Therefore, since the fuel oil supply will be maintained at its 
present quality level, there should be no increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    As stated in (1) above, the proposed amendment only provides a 
quantitative method for the acceptability determination of new 
Diesel Generator fuel oil. The proposed testing will continue to 
verify the high quality and acceptability of the fuel oil supply. 
There should not be any possibility that a new or different kind of 
accident from those previously evaluated is created.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The only margin associated with this amendment is the margin 
between the acceptance limit on water and sediment in the fuel oil 
supply and the quantity of water and sediment that could impact 
Diesel Generator operation. The ``clear and bright'' testing per 
ASTM-D4176-1982 and the proposed quantitative testing per ASTM-
D1796-1968 are both written to detect and reject fuel oil containing 
water or sediment at essentially the same level. The margin of 
safety to Diesel Generator impairment is therefore not reduced. This 
amendment 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: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036.
    NRC Project Director: Suzanne C. Black.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: February 21, 1992
    Description of amendment request: The proposed amendment would 
revise Technical Specifications Appendix B, Environmental Protection 
Plan (Non-radiological), by removing Sections 2.3 and 4.3, ``Cultural 
Resources.'' Union Electric has developed and maintains a management 
plan for the protection of cultural resources on the Callaway Plant 
site. The amendment request summarizes the plan that provides the 
status and disposition of each portion of the current Appendix B 
sections related to cultural resources.
    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[s] do[es] not involve a significant hazards 
consideration because operation of Callaway Plant in accordance with 
these change[s] would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)).
    The proposed changes are administrative in nature and have no 
impact on safety-related structures, systems or components. 
Therefore, there is no impact on the probability of occurrence or 
the consequences of an accident or malfunction of equipment 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated (10 CFR 50.92(c)(2)).
    The proposed change[s] do[es] not affect any of the assumptions 
used in previous accident evaluations. All accidents continue to be 
bounded by previous analyses and deletion of satisfied Environmental 
Protection Plan requirements is administrative and will therefore 
not introduce the possibility of any new or different kind of 
accident.
    3. Involve a significant reduction in a margin of safety (10 CFR 
50.92(c)(3)).
    The proposed changes are administrative and do not affect the 
margin of safety as defined in the basis for any Technical 
Specification.

    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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: John N. Hannon.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: July 16, 1993
    Description of amendment request: The amendment would revise 
Technical Specifications 3/4.8.1.1 and 3/4.8.1.2, A. C. Sources, 
Operating and Shutdown. The proposed revision changes the minimum 
required storage volume of the Emergency Fuel Oil storage and day tanks 
from 85,300 gallons and 390 gallons to 80,400 gallons and 510 gallons. 
These changes are the result of inconsistencies found by Union 
Electric, in the calculations and T/S Bases for tank capacities, while 
performing a self-initiated Electrical Distribution System Functional 
Assessment (EDSFA). The Electrical Distribution System Functional 
Inspection (EDSFI) performed by the NRC re-examined this issue and it 
was determined that NUREG-1431 provided bases for the change to day 
tank level. This submittal is a T/S enhancement. Basis for proposed no 
significant hazards consideration determination: As required by 10 CFR 
50.91(a), the licensee has provided its analysis of the issue of no 
significant hazards consideration, which is presented below:

    The proposed changes to Technical Specifications 3/4.8.1.1 and 
3/4.8.1.2 do not involve a significant hazards consideration because 
operation of Callaway Plant with these changes would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The current minimum required volume of the day tanks as given in 
T/S 3.8.1.1b.1 and 3.8.1.2b.1 is based on fuel oil with a specific 
gravity value in the upper range of values allowed by T/S. 
Calculations made using fuel oil with a specific gravity of 39 
degrees API, which is the minimum allowed specific gravity, indicate 
a larger minimum required volume is needed in the day tanks. The 
increased minimum required volume provides additional conservatism 
for the day tanks to perform their intended safety function based on 
the possibility of using different specific gravity fuel oil. The 
current minimum required volume of the storage tanks as given in T/S 
3.8.1.1b.2 and 3.8.1.2b.2 was calculated by using an overly 
conservative Net Positive Suction Head (NPSH) available calculation. 
The original calculation assumed a static head was needed in order 
for the NPSH available to exceed the NPSH required and assure proper 
operation of the fuel oil transfer pumps. Union Electric 
calculations have determined that the NPSH required for the transfer 
pumps is maintained as long as the pump suction is submerged. Based 
on this calculation the minimum required volume of the storage tanks 
can be reduced and the diesel generators can still achieve their 
required seven days of operation at continuous rating plus an 
additional volume available to be used for testing.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    There is no new type of accident or malfunction being created 
and the method and manner of plant operation remains unchanged. The 
safety design bases in the FSAR have not been altered, the 
requirement for continuous operation remains unchanged.
    (3) Involve a significant reduction in a margin of safety.
    There are no plant design changes involved and no changes are 
being made to the safety limits or safety system settings that would 
adversely impact plant safety. The minimum required storage volumes 
of the day and storage tanks are being changed based on calculations 
using conservative data.
    Based on the above discussions, it has been determined that the 
requested Technical Specification revisions do not involve a 
significant increase in the probability or consequences of an 
accident or other adverse condition over previous evaluations; or 
create the possibility of a new or different kind of accident or 
condition over previous evaluations; or involve a significant 
reduction in a margin of safety. Therefore, the requested license 
amendment 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 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: February 25, 1994
    Description of amendment request: The proposed change would revise 
the Technical Specifications (TS) for the North Anna Power Station, 
Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed change would 
modify the surveillance frequency of the nozzles in the Quench Spray 
and the Recirculation Spray Systems at NA-1&2.
    The proposed changes to the surveillance requirements for the 
nozzles in the Quench Spray System and Recirculation Spray Systems are 
consistent with the intent of Generic Letter 93-05, ``Line-Item 
Technical Specifications Improvement to Reduce Surveillance 
Requirements for Testing During Power Operation,'' dated September 27, 
1993, which is to improve safety, decrease equipment degradation, and 
reduce unnecessary burden on personnel resources by reducing testing 
requirements that are marginal to safety.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Specifically, operation in accordance with the proposed 
Technical Specifications changes will not:
    1. Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated.
    The proposed reduced testing frequency of the Spray Systems 
nozzles does not change the way the systems are operated or the 
Spray Systems operability requirements. The proposed change to the 
surveillance frequency of safety equipment has no impact on the 
probability of an accident occurrence nor can it create a new or 
different type of accident. NUREG-1366, ``Improvements to Technical 
Specifications Surveillance Requirements,'' dated December 1992, 
concluded that the corrosion of stainless steel piping is negligible 
during the extended surveillance interval. Since the Spray Systems 
are maintained dry there is no additional mechanism that could cause 
blockage of the spray nozzles. Thus, the nozzles in the Spray 
Systems will remain operable during the ten year surveillance 
interval to mitigate the consequence of an accident previously 
evaluated. To date, no clogging or blockage of the nozzles in the 
Spray [System] during the five year surveillance tests [has] been 
observed at . . . North Anna. Testing of the Spray Systems nozzles 
at the proposed reduced frequency will not increase the probability 
of occurrence of a postulated accident or the consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed reduced frequency testing of the Spray Systems' 
nozzles does not change the way the Spray Systems are operated. The 
reduced frequency of testing of the spray nozzles does not change 
plant operation or system readiness. The reduced frequency testing 
of the Spray Systems' nozzles does not generate any new accident 
precursors. Therefore, the possibility of a new or different kind of 
accident previously evaluated is not created by the proposed changes 
in surveillance frequency of the Spray Systems nozzles.
    3. Involve a significant reduction in a margin of safety.
    Reduced testing of the Spray Systems' nozzles does not change 
the way the Systems are operated or the Spray Systems' operability 
requirement. NUREG-1366 concluded that the corrosion of stainless 
steel piping is negligible during the extended surveillance 
interval. Since the Spray Systems are maintained dry there is no 
additional mechanism that could cause blockage of the Spray Systems' 
nozzles. Thus, the proposed reduced testing frequency is adequate to 
ensure spray nozzle operability. The surveillance requirements do 
not affect the margin of safety in that the operability requirements 
of the Spray Systems remains unaltered. The existing safety analysis 
remains bounding. Therefore, no margins of safety are adversely 
affected.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Herbert N. Berkow.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: February 25, 1994
    Description of amendment request: The proposed changes would revise 
the surveillance frequency from 5 years to 10 years for the spray 
nozzles in the containment spray and recirculation spray systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Specifically, operation of Surry [Units 1 and 2] in accordance 
with the proposed Technical Specifications changes will not:
    1. Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated.
    The proposed reduced testing frequency of the Spray Systems' 
nozzles does not change the way the systems are operated or the 
Spray Systems' operability requirements. The proposed change to the 
surveillance frequency of safety equipment has no impact on the 
probability of an accident occurrence nor can it create a new or 
different type of accident. NUREG-1366, ``Improvements to Technical 
Specifications Surveillance Requirements,'' dated December 1992, 
concluded that the corrosion of stainless steel piping is negligible 
during the extended surveillance interval. Since the Spray Systems 
are maintained dry there is no additional mechanism that could cause 
blockage of the spray nozzles. Thus, the nozzles in the Spray 
Systems will remain operable during the 10 year surveillance 
interval to mitigate the consequence of an accident previously 
evaluated. To date, no clogging or blockage of the nozzles in the 
Spray Systems during the five year surveillance tests have been 
observed at Surry. Testing of the Spray Systems' nozzles at the 
proposed reduced frequency will not increase the probability of 
occurrence of a postulated accident or the consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed reduced frequency testing of the Spray Systems' 
nozzles does not change the way the Spray Systems are operated. The 
reduced frequency of testing of the spray nozzles does not change 
plant operation or system readiness. The reduced frequency testing 
of the Spray Systems' nozzles does not generate any new accident 
precursors. Therefore, the possibility of a new or different kind of 
accident previously evaluated is not created by the proposed changes 
in surveillance frequency of the Spray Systems nozzles.
    3. Involve a significant reduction in a margin of safety.
    Reduced testing of the Spray Systems' nozzles does not change 
the way the Systems' are operated or the Spray Systems' operability 
requirement. NUREG-1366 concluded that the corrosion of stainless 
steel piping is negligible during the extended surveillance 
interval. Since the Spray Systems are maintained dry there is no 
additional mechanism that could cause blockage of the Spray Systems' 
nozzles. Thus, the proposed reduced testing frequency is adequate to 
ensure spray nozzle operability. The surveillance requirements do 
not affect the margin of safety in that the operability requirements 
of the Spray Systems remains unaltered. The existing safety analysis 
remains bounding. Therefore, no margins of safety are adversely 
affected.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: February 17, 1994
    Description of amendment request: The amendment proposes to modify 
the Technical Specifications (TS) to reflect management and 
organizational changes at the Washington Public Power Supply System 
(the licensee) for operation of the WNP-2 facility. The proposed 
changes would (1) modify the reporting responsibility of the quality 
assurance organization from the Managing Director to the Assistant 
Managing Director, Operations (AMDO), and (2) modify the appointment 
authority for the Corporate Nuclear Safety Review Board (CNSRB) from 
the Managing Director to the AMDO. These changes are proposed to 
reflect the current designation of the AMDO as the licensee's 
designated official with corporate responsibility for overall plant 
nuclear safety, and as the direct report for the CNSRB.
    In addition, the proposed change would (1) delete the specific 
requirement for health physics/chemistry program procedures, (2) modify 
the titles of two positions on the Plant Operations Committee (POC) to 
reflect revised organizational titles, (3) modify the CNSRB quorum 
requirements from nine personnel to a minimum of nine personnel, and 
(4) delete the requirement that the CNSRB Executive Secretary be 
designated from the CNSRB membership.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    All of the proposed changes are administrative in nature, and do 
not involve any change in the design or operation of the plant. None of 
the changes affect any initiating events, nor do they affect plant 
response to postulated events already analyzed. The proposed changes do 
not, therefore, affect 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 changes do not affect the design, operation, 
maintenance, or testing of the plant. They do not, therefore, create 
the possibility for any 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 changes do not affect any accident analyses, and do 
not, therefore, affect any of the margins of safety affected by the 
design or operational limitations of the plant. The proposed changes do 
not, therefore, affect any margin of safety.
    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: 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

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

    Date of amendment request: February 23, 1994
    Description of amendment request: The proposed amendment revises 
Technical Specification 3.8.1.1, AC Sources Operating, and 3.8.1.2, AC 
Sources Shutdown, to increase the minimum volume of fuel oil required 
for the emergency diesel generator fuel oil day tanks. Several other 
changes have been proposed to correct editorial errors related to 
previously issued amendments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee 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 current minimum required volume of the day tanks as given in 
Technical Specifications 3.8.1.1 and 3.8.1.2 is based on fuel oil 
with a specific gravity value in the upper range of values allowed 
by the technical specifications. Calculations made using fuel oil 
with a specific gravity of 39 degrees API, which is the minimum 
allowed specific gravity, indicate a larger minimum required volume 
is needed in the day tanks. The increased minimum required volume 
provides additional conservatism for the day tanks to perform their 
intended safety function based on the possibility of using different 
specific gravity fuel oil. The proposed change will not prevent the 
Emergency Diesel Generator fuel oil system from performing its 
design function, nor require the system to be operated in a manner 
different than that for which it was designed. Therefore, the 
proposed change will not increase the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated. There 
is no new type of accident or malfunction being created and the 
method and manner of plant operation remains unchanged. The safety 
design bases in the Updated Safety Analysis Report (USAR) have not 
been altered, and current operating requirements of the Emergency 
Diesel Generators remain unchanged. 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 basis for the existing emergency fuel oil day tank level 
requirements is to ensure that sufficient fuel oil is available to 
meet the operational requirements specified in ANSI N195-1976. This 
proposed change to the minimum required storage volume of the day 
tanks is based on revised calculations performed in accordance with 
ANSI N195-1976 using conservative data. This proposed change will 
not change the operation of the plant. Thus, the proposed change 
will continue to ensure the Emergency Diesel Generator operating 
requirements. There are no changes being made to the safety limits 
or safety system settings that would adversely impact plant safety. 
Therefore, the proposed change will not cause a reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: Suzanne C. Black

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

    Date of amendment request: February 24, 1994
    Description of amendment request: The proposed amendment revises 
Technical Specification 3.9.4, Containment Building Penetrations, to 
allow use of temporary alternate closure methods for the emergency 
personnel escape lock and containment wall penetrations, during 
alterations of the core or movement of irradiated fuel within the 
containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee 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 of a previously evaluated accident 
is not increased because failure to maintain containment closure is 
not an initiating condition for fuel handling an accident. The use 
of temporary alternate closure methods for the emergency personnel 
escape lock and containment wall penetrations does not introduce any 
new potential accident initiating condition during refueling 
operation.
    The consequences of an accident previously evaluated is not 
increased because the use of
temporary alternate closure methods for the emergency personnel escape 
lock and containment wall penetrations will provide the assurance of 
containment closure during refueling activities. The ability of [the] 
containment to restrict the release of any fission product 
radioactivity to the environment remains unchanged.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The failure of the temporary alternate closure methods for the 
emergency personnel escape lock and containment wall penetrations 
during refueling will not result in a malfunction of any other plant 
equipment. The sole purpose of establishing containment closure for 
refueling is to restrict the release of any fission product 
radioactivity in the event of a fuel handling accident.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The temporary alternate closure methods for the emergency 
personnel escape lock and containment wall penetrations will provide 
the same assurance of containment closure during refueling for 
credible accident scenarios. The ability of containment to restrict 
the release of any fission product radioactivity to the environment, 
should a fuel handling accident occur, remains unchanged.

    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: Suzanne C. Black

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

    Date of amendment request: February 24, 1994
    Description of amendment request: The proposed amendment revises 
Technical Specification 4.7.1.2.1.a to require that the turbine-driven 
and motor-driven auxiliary feedwater pumps be tested at least quarterly 
on a staggered test basis instead of the currently required testing 
once per 31 days on a staggered test basis. In addition, the proposed 
changes would revise Technical Specification Bases 3/4.7.7, Emergency 
Exhaust System--Auxiliary Building, and 3/4.9.13, Emergency Exhaust 
System--Fuel Building, to eliminate the reference to the use of 
automatic control for the emergency exhaust system heaters.
    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 staff has reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's review is presented 
below.
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    This change only revises the surveillance requirement for the 
auxiliary feedwater pumps. The purpose of this surveillance requirement 
is to prove that the pumps are operable. The longer test interval 
should result in greater availability by reducing the rate of test 
induced failures which should offset any loss in reliability. The 
revised surveillance requirement does not affect the probability of 
accident initiation. The operability of the pumps is maintained and 
therefore the consequences of evaluated accidents is unaffected by the 
proposed change. The revised surveillance frequency is consistent with 
the guidance issued in Generic Letter 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation.''
    Changes to the emergency exhaust system technical specifications 
eliminate the words ``using automatic control'' associated with the 
humidity controlling heaters. These changes reflect the current method 
in which the fuel building emergency exhaust system heaters are 
controlled in that humidity sensors are bypassed to allow continuous 
operation of the heaters whenever the emergency exhaust system fans are 
operating. The proposed changes do not affect the probability of 
initiating an accident previously evaluated. The ability of the 
emergency exhaust systems to mitigate the consequences of an accident 
are likewise unaffected since the heaters remain available and 
operating to control humidity.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Verification of pump operability is still maintained with the 
change to the frequency of the surveillance requirement. No system 
configuration changes are being implemented in order to perform the 
surveillance testing and any potential accidents that may be associated 
with the surveillance testing were previously considered. The changing 
of the surveillance frequency does not introduce any additional failure 
modes for the auxiliary feedwater system.
    The proposed revisions to the emergency exhaust system technical 
specifications are limited to the automatic control of the humidity 
controlling heaters. The proposed change does not introduce any new 
potential challenges to fission product barriers or any new failure 
modes which might prevent the emergency exhaust systems from fulfilling 
their accident mitigating functions.
    3. The proposed change does not involve a significant reduction in 
the margin of safety.
    The inservice testing program will continue to ensure that 
auxiliary feedwater pumps operational readiness criteria are consistent 
with the requirements of ASME Section XI. System performance 
surveillances will continue to be conducted in accordance with the 
technical specifications. The proposed change does not significantly 
reduce the margin of safety because the availability and reliability of 
the auxiliary feedwater system are not significantly decreased and the 
heat removal requirements for the system are unchanged.
    The proposed changes to the emergency exhaust system technical 
specifications affect the requirements to maintain heater control in 
the automatic mode. The proposed changes do not decrease the exhaust 
systems' actual ability to control humidity or alter the other 
functional requirements for the system. Therefore, the proposed change 
does not significantly reduce any margin of safety related to the 
performance of the emergency exhaust systems.
    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: Suzanne C. Black

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

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

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: March 14, 1994
    Description of amendment request: The proposed change to the 
Millstone Unit 2
Technical Specifications (TS) would provide a one-time extension of the 
surveillance frequency from the required 18-month to the next refueling 
outage but no later than September 30, 1994, of the power operated 
valves in the service water system (TS 4.7.4.1.b) and in the boron 
injection flow path (TS 4.1.2.2.c). This would extend the surveillance 
for these valves approximately 5 months. Date of publication of 
individual notice in Federal Register: March 23, 1994 (59 FR 13751).
    Expiration date of individual notice: April 22, 1994
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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

    Date of application for amendment: April 15, 1992, as modified 
December 8, 1992, June 25, 1993, and February 2, 1994
    Brief description of amendment: This amendment revises the 
provisions in the Technical Specifications to incorporate Generic 
Letter 90-06, ``Resolution of Generic Issue 70, 'Power-Operated Relief 
Valve and Block Valve Reliability,' and Generic Issue 94, `Additional 
Low-Temperature Overpressure Protection for Light-Water Reactors,''' 
power-operated relief valve requirements for power operation, and to 
modify the primary coolant system overpressure protection specification 
venting requirements.
    Date of issuance: March 29, 1994
    Effective date: March 29, 1994, with full implementation within 60 
days
    Amendment No.: 160
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4937) The licensee's February 2, 1994, letter provided minor editorial 
changes which did not alter the staff's initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
March 29, 1994. No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendments: January 18, 1994, as 
supplemented March 8, 1994.
    Brief description of amendments: The proposed amendments remove the 
tables of containment penetration conductor overcurrent protective 
devices from the Technical Specifications (TS) in accordance with the 
guidance contained in Generic Letter 91-08, ``Removal of Component 
Lists from Technical Specifications.'' The tables will be relocated to 
Chapter 16 of the Catawba Final Safety Analysis Report (Selected 
Licensee Commitments Manual).
    Date of issuance: March 21, 1994
    Effective date: March 21, 1994
    Amendment Nos.: 114 and 108
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7688) The March 8, 1994, letter provided clarifying information that 
did not change the scope of the January 18, 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 March 21, 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

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

    Date of application for amendments: January 13, 1993, as 
supplemented January 28 and April 26, 1993
    Brief description of amendments: The amendments revise Technical 
Specifications (TS) 2.2.1, 3/4.1.2.5, 3/4.1.2.6, 3/4.3.3.12, 3/4.5.1, 
3/4.5.4, 3/4.9.2, their associated Bases, and TS 6.9.1.9 to relocate 
the values of certain cycle-dependent limits from the TS to the Core 
Operating Limits Report.
    Date of issuance: March 25, 1994
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 115 and 109
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46227). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 25, 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

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

    Date of application for amendments: January 10, 1994, as 
supplemented March 21, 1994
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Table 2.2-1 and TS 4.2.5 to allow a change in the 
method for measuring reactor coolant system flowrate from the 
calorimetric heat balance method to a method based on a one-time 
calibration of the reactor coolant system cold leg elbow differential 
pressure taps.
    Date of issuance: March 30, 1994
    Effective date: March 30, 1994
    Amendment Nos.: 116 and 110
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 1994 (59 FR 
3743). The March 21, 1994, letter provided clarifying
information that did not change the scope of the January 10, 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 March 30, 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

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

    Date of application for amendments: October 25, 1993, as 
supplemented December 3, 1993, and February 14, 1994.
    Brief description of amendments: The amendments reduce the required 
minimum measured reactor coolant system flow from 385,000 gallons per 
minute (gpm) to 382,000 gpm.
    Date of issuance: March 22, 1994
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 141 and 123
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67842) The December 3, 1993, letter provided information in response 
to a staff's November 19, 1993, request for additional information. The 
February 14, 1994, letter requested NRC staff approval that the flow-
reduction portion of the requested technical specification (TS) change 
be separated from the approval of the safety limit portion of the 
requested TS change. The resultant changes from the above described 
actions reduced the scope of the initial proposed no significant 
hazards consideration determination, but, otherwise, did not change the 
NRC staff's position that the three standards of 10 CFR 50.92(c) are 
satisfied.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 22, 1994. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

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 application for amendments: October 26, 1993
    Brief description of amendments: These amendments change Technical 
Specification 4.6.1.2.a, Containment Leakage Surveillance Requirements, 
to be consistent with the guidance of NUREG-1432, ``Standard Technical 
Specifications for Combustion Engineering Plants.''
    Date of Issuance: March 30, 1994
    Effective Date: March 30, 1994
    Amendment Nos.: 127 and 66
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67845) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 30, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: May 21, 1993, as supplemented 
January 25, 1994.
    Brief description of amendments: These amendments revise the 
containment design pressure from 59 psig to 55 psig.
    Date of issuance: March 30, 1994
    Effective date: March 30, 1994
    Amendment Nos. 160 and 154
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36434) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 30, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendment: July 28, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications by changing the reporting frequency of the Radioactive 
Material Release Report and the 10 CFR 50.59 reporting of facility 
changes, tests, and experiments. This is to reflect the new 
requirements of 10 CFR Part 50. The Offsite Dose Assessment Manual 
(ODAM) will remain part of the Radioactive Material Release Report and 
be submitted on an annual basis. The request for reporting of the 
safety and relief valve challenge information six months after each 
refueling outage is denied.
    Date of issuance: March 22, 1994
    Effective date: March 22, 1994
    Amendment No.: 196
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 15, 1993 (58 
FR 48384) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 22, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear 
Power Station, Unit 1, New London County, Connecticut

    Date of application for amendment: January 14, 1994
    Brief description of amendment: The amendment corrects an editorial 
error. Specifically, the amendment changes the reference in Limiting 
Condition for Operation (LCO) 3.4.D from ``3.3.A through C'' to 
``3.4.A, 3.4.B, and 3.4.C.''
    Date of issuance: March 24, 1994
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 72
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7693) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 24, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, Millstone 
Nuclear Power Station, Unit 1, New London County, Connecticut

    Date of application for amendment: November 22, 1993, as 
supplemented March 4, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications by clarifying the operability requirements relative to 
the design function of the scram discharge volume--water level high rod 
block. In addition, NNECO is adding a statement which defines 
operability and surveillance requirements for the rod block functions 
while the reactor mode selector switch is in the REFUEL or SHUTDOWN 
positions.
    Date of issuance: March 30, 1994
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 73
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67851). The March 4, 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 March 30, 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.

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

    Date of application for amendment: September 24, 1993
    Brief description of amendment: The amendment removes the listing 
of ``Enclosure Building Bypass Leakage Paths'' from the Technical 
Specifications, and makes a number of editorial changes.
    Date of issuance: March 24, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 89
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59752). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 24, 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.

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

    Date of amendment request: December 19, 1990, and June 1, 1992, as 
supplemented by letters dated February 1, 1993, and February 25, 1994.
    Brief description of amendment: The amendment changes the Technical 
Specification (TS) by revising the pressure-temperature limits in TS 
2.1.2 and would make the limits valid for 20 effective full-power years 
of operation. The amendment also modifies TS 2.1.1 to change the 
minimum requirements for starting a non-operating reactor coolant pump 
and modifies TS 2.3(3) to change the requirements for disabling high-
pressure safety injection pumps during scheduled heatup and cooldown 
operations. Lastly, the amendment modifies TS 2.1.6 to change the 
power-operated relief valve limiting conditions of operation and 
surveillance requirements. The amendment request was filed in response 
to Generic Letter 90-06.
    Date of issuance: March 23, 1994
    Effective date: March 23, 1994
    Amendment No.: 161
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 8, 1992 (57 FR 
30255). The additional information contained in the supplemental 
letters dated February 1, 1993 and February 25, 1994, was clarifying in 
nature and, thus, within the scope of the initial notice and did not 
affect the staff's proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 23, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of amendment request: September 17, 1993
    Brief description of amendment: The amendment made changes to the 
Technical Specification (TS) to revise the minimum requirement of fuel 
oil that must be in the Emergency Diesel Generator (EDG) fuel oil 
storage tank in TS 2.7(1).
    Date of issuance: March 29, 1994
    Effective date: 120-days from the date of issuance.
    Amendment No.: 162
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52991). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 29, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of application for amendment: February 17, 1994 (Reference LAR 
94-02)
    Brief description of amendment: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit 
Nos. 1 and 2 to revise TS 3/4.3.2, ``Engineered Safety Features 
Actuation System Instrumentation,'' as follows:
    (1) Table 3.3-3, functional unit 6.c.2), channels to trip, would be 
changed from 2/steam generator in one steam generator to 2/steam 
generator in any 2 steam generators, due to an administrative error.
    (2) Table 3.3-4 would be changed as follows:
    a. Functional Unit 4.e., Negative Steam Pressure Rate--High, trip 
setpoint and allowable value, would be changed from -100 psi/sec and 
-105.4 psi/sec to 100 psi and 105.4 psi, respectively.
    b. A note would be added stating that the time constants utilized 
in the rate-lag controller for Negative Steam Pressure Rate--High, are 
equal to 50 seconds.
    Date of issuance: April 1, 1994
    Effective date: 30 days after the date of issuance.
    Amendment Nos.: 92 and 91
    Facility Operating License Nos. DPR-80 and DPR-82: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1994 (59 FR 
9789). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 1, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendments: October 27, 1993
    Brief description of amendments: These amendments delete the 
requirement for the 1) Plant Manager or Superintendent-Operations, 2) 
Assistant Superintendent-Operations, and 3) Superintendent-Technical or 
Engineer-Systems to hold a Senior Reactor Operator (SRO) license, and 
add the requirement for the Senior Manager-Operations to hold an SRO 
license.
    Date of issuance: March 22, 1994
    Effective date: March 22, 1994
    Amendments Nos.: 185 and 190
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64613). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 22, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, Pennsylvania 
17105.

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: January 11, 1994
    Brief description of amendment: The amendment provides one-time 
relief from the requirement to perform Type C tests (local leak rate 
tests) at intervals of no greater than 2 years for the shutdown cooling 
isolation valves (10MOV-17 and 10MOV-18). This one-time only delay, 
until the next refueling outage currently scheduled to begin in 
November 1994, was requested for the performance of these leakage 
tests. The request was necessitated by the extended 1991-1993 refueling 
outage and the length of the current operating cycle.
    Date of issuance: March 18, 1994
    Effective date:  As of the date of issuance to be implemented 
within 30 days.
    Amendment No.: 208
    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 March 18, 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. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: May 18, 1993, as supplemented 
March 3, 1994
    Brief description of amendment: The amendment revised the Technical 
Specifications (Appendix A) and the Radiological Environmental 
Technical Specifications (Appendix B) to incorporate the changes listed 
below:
    (1) The frequency of city water connection to charging pumps and 
boric acid piping testing (specified in Appendix A Table 4.1-3) was 
changed to accommodate operation on a 24-month cycle.
    (2) The frequency of boric acid tank level instrument calibration 
(specified in Appendix A Table 4.1-1) was changed to accommodate 
operation on a 24-month cycle.
    (3) The frequency of boric acid makeup flow instrument calibration 
(specified in Appendix A Table 4.1-1) was changed to accommodate 
operation on a 24-month cycle.
    (4) The frequency of primary water storage tank level instrument 
calibration and functional testing (specified in Appendix B Table 3.1-
1) was changed to accommodate operation on a 24-month cycle.
    These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle,'' as applicable.
    Date of issuance: March 21, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 144
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34089) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 21, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: December 20, 1993
    Brief description of amendment: The Technical Specifications (TSs) 
amendment revised Section 3.3.D.1.a, and associated Bases in Sections 
3.3 and 4.4, to allow for the disconnection of those portions of the 
weld channel pressurization system that become inoperable and are not 
practicably accessible for repair. Additionally, the amendment revised 
TSs 3.3.B.3.b and 3.3.E.3.b, and the Bases of Section 3.3, to correct 
previous administrative errors.
    Date of issuance: March 31, 1994
    Effective date: As of the date of issuance to be implemented prior 
to exceeding cold shut down from the current outage.
    Amendment No.: 145
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2870) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 1994. No significant hazards 
consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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: November 24, 1993, as supplemented 
February 15, 1994.
    Brief description of amendments: The amendments change the TS to 
modify the requirements of TS 3.3.1 and 3.3.2. The proposed changes 
relocate Tables 3.3-2 and 3.3-5, which provide the response time limits 
for the reactor trip system and the engineered safety features 
actuation system instruments, from the TS to the updated final safety 
analysis report.
    Date of issuance: March 21, 1994
    Effective date: March 21, 1994
    Amendment Nos.: 105 and 98
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1994 (59 FR 
629) The February 15, 1994, submittal provided clarifying information 
that did not change the initial determination of no significant hazards 
consideration as published in the Federal Register. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated March 21, 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. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: December 7, 1993 (TS 93-16)
    Brief description of amendments: The amendments incorporate various 
changes to the Administrative Controls section of the technical 
specifications and the Sequoyah Operating License. These changes 
include overtime approval authority, titles for the Plant Operations 
Review Committee, Radiological Assessment Review Committee 
requirements, Offsite Dose Calculation Manual implementation, Quality 
Assurance Program procedures, condenser in-leakage monitoring 
requirements, and changes to position titles and references.
    Date of issuance: March 31, 1994
    Effective date: March 31, 1994
    Amendment Nos.: 178 and 169
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4948) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated March 31, 1994. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of application for amendment: February 8, 1994 TS 94-02
    Brief description of amendment: The amendment modifies Operating 
License Condition 2.C.(17) to provide a limited extension of the 
surveillance test intervals for certain specified instrumentation on 
Unit 2 to coincide with the Cycle 6 refueling outage that is scheduled 
to start in July 1994. The surveillance intervals that are affected 
will not exceed 28 months for 18-month surveillances and 46 months for 
the 3-year Containment fire hose hydrostatic surveillance test.
    Date of issuance: March 31, 1994
    Effective date: March 31, 1994
    Amendment No.: 170
    Facility Operating License Nos. DPR-79: Amendment revises the 
technical specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10015) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 31, 1994. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendment: November 13, 1992, as 
supplemented on July 15 and November 10, 1993.
    Brief description of amendment: The proposed amendment would revise 
Appendix A, TS 3/4.3.1, ``Reactor Protection System (RPS) 
Instrumentation'' and TS 3/4.3.2.3, ``Anticipatory Reactor Trip System 
(ARTS) Instrumentation'' to increase RPS and ARTS channel functional 
test surveillance test intervals and RPS allowed out of service times. 
These requests are made based on the NRC approved Babcock & Wilcox 
Topical Report, BAW-10167. Also, the addition of an action statement to 
permit continued operation for 48 hours with two RPS channels 
inoperable and to remove channel functional test surveillance 
requirements for source and intermediate range neutron flux 
instrumentation is requested. Finally, a revision to Table 4.3-1 to 
decrease the channel calibration surveillance test interval for the 
``High Flux Number of Reactor Coolant Pumps On'' trip is proposed.
    Date of issuance: March 28, 1994
    Effective date: March 28, 1994
    Amendment No. 185
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41516) The July 15 and November 10, 1993 letters, provided supplemental 
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 
March 28, 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.

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

    Date of application for amendment: August 4, 1993
    Brief description of amendment: The amendment revises the plant 
Technical Specifications (TSs) to modify the requirement for periodic 
surveillance of the emergency diesel generators (EDGs) to permit a slow 
start in place of the existing requirement to perform a monthly fast 
start. A fast start shall be performed every 6 months. The amendment 
also allows engine prelubrication and warmup when an EDG is started for 
surveillance testing.
    Date of issuance: March 22, 1994
    Effective date: March 22, 1994
    Amendment No.: 138
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62157) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 22, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301.

    Dated at Rockville, Maryland, this 6th day of April.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Acting Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation.
[FR Doc. 94-8780 Filed 4-12-94; 8:45 am]
BILLING CODE 7590-01-P