[Federal Register Volume 60, Number 2 (Wednesday, January 4, 1995)]
[Notices]
[Pages 493-513]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-100104]



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[[Page 494]]

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 December 12, 1994, through December 21, 
1994. The last biweekly notice was published on December 21, 1994 (59 
FR 65809).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.

    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.

    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.

    By February 3, 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 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 [[Page 495]] 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, 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, 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: November 30, 1994
    Description of amendment requests: The proposed amendment would 
relocate Table 3.3-2, ``Reactor Protective Instrumentation Response 
Times,'' and Table 3.3-5, ``Engineered Safety Features Response 
Times,'' of Technical Specifications (TS) 3/4.3.1 and 3/4.3.2, 
respectively, to the Palo Verde Updated Final Safety Analysis Report 
(UFSAR) in accordance with the guidance provided in Generic Letter (GL) 
93-08. In addition, the proposed amendment would make administrative 
changes to two previous TS amendment requests to reflect the deletion 
of Tables 3.3-2 and 3.3-5. The amendment would also delete an obsolete 
footnote on page 3/4 3-17 of the Palo Verde Unit 2's TS.
    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 -- Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change relocates two tables of instrument response 
time limits from the TS to the UFSAR. The changes are in accordance 
with the guidance provided by the NRC in Generic Letter 93-08. The 
changes are administrative in nature and do not involve any 
modifications to plant equipment or affect plant operation. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Standard 2 - Does the proposed change create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed change relocates two tables of instrument response 
time limits from the TS to the UFSAR. The changes are in accordance 
with the guidance provided by the NRC in Generic Letter 93-08. The 
changes are administrative in nature, do not involve any 
modifications to plant equipment and cause no change in the method 
by which any safety-related system performs its function. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Standard 3 - Does the proposed change involve a significant 
reduction in a margin of safety?
    The proposed change relocates two tables of instrument response 
time limits from the TS to the UFSAR. The changes are in accordance 
with the guidance provided by the NRC in Generic Letter 93-08. The 
changes are administrative in nature, do not change or alter 
regulatory requirements and do not affect the safety analysis. Plant 
procedures contain response time testing acceptance criteria that 
reflect the reactor trip and ESFAS [engineered safety feature 
actuation system] response time limits in the tables being relocated 
from the TS into the UFSAR. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    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


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: December 7, 1994
    Description of amendment requests: The proposed amendment would 
change Table 4.3-1 of Technical Specification 3/4.3.1 to allow 
verification of the shape annealing matrix elements used in the Core 
Protection Calculators. This would provide the option to use generic 
shape annealing matrix elements in the Core Protection Calculators. 
Presently, cycle-specific shape annealing elements are determined 
during startup testing after each core reload. Use of a generic shape 
annealing matrix would eliminate approximately 2 to 3 hours of critical 
path work during startup after a refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis [[Page 496]] about the issue of no significant 
hazards consideration, which is presented below:
    Standard 1 -- Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed Technical Specification amendment does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated. The Technical Specification amendment 
provides the option to use generic shape annealing matrix elements 
in the Core Protection Calculators. The design basis of the Core 
Protection Calculators is to provide the DNBR [departure from 
nucleate boiling ratio] and linear heat rate trip functions for the 
Reactor Protection System so that the Specified Acceptable Fuel 
Design Limits on DNBR and fuel centerline melt are not exceeded 
during normal operation or Anticipated Operational Occurrences, and 
assist the Engineered Safety Features Actuation System in limiting 
the consequences of postulated accidents. The generic shape 
annealing matrix elements will be validated during startup testing 
and will meet the same acceptance criteria as the cycle specific 
shape annealing matrix elements. If the generic shape annealing 
matrix elements are not valid, cycle specific shape annealing matrix 
elements would be used in the Core Protection Calculators. This 
change will not affect the Core Protection Calculators capability to 
protect the plant by tripping the reactor, based on a conservative 
calculation of minimum DNBR and peak linear heat rate, to ensure 
that the Specified Acceptable Fuel Design Limits are not violated in 
the event of an Anticipated Operational Occurrence. Therefore, the 
generic shape annealing matrix elements will not affect the safety 
analysis, since there is no change to the design basis of the Core 
Protection Calculator System.
    Standard 2 -- Does the proposed change create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed Technical Specification amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated. Since the generic shape annealing 
matrix elements will still have to meet the same acceptance criteria 
as the cycle specific shape annealing matrix elements, the Core 
Protection Calculators will still generate axial power shapes that 
fall within the required uncertainties. The Core Protection 
Calculators will still trip the reactor, based on a conservative 
calculation of minimum DNBR and peak linear heat rate, to ensure 
that the Specified Acceptable Fuel Design Limits are not violated in 
the event of an Anticipated Operational Occurrence.
    Standard 3 -- Does the proposed change involve a significant 
reduction in a margin of safety?
    The proposed Technical Specification amendment will not involve 
a significant reduction in a margin of safety. There is no reduction 
in the margin of safety, since the generic shape annealing matrix 
elements will still have to meet the same acceptance criteria as the 
cycle specific shape annealing matrix elements. Therefore, this 
change will not affect the design basis of the Core Protection 
Calculators. The Core Protection Calculators will still provide a 
reactor trip based on a conservative calculation of minimum DNBR and 
peak linear heat rate.
    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


Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit No. 2, Maricopa County, 
Arizona

    Date of amendment request: November 30, 1994
    Description of amendment request: The proposed amendment would 
change the pressurizer code safety valve lift setting from 2500 psia to 
2475 psia. The lift setting is being changed to permit Unit 2 to 
operate with up to 1500 plugged tubes in each steam generator.
    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 -- Does the proposed change involve a significant 
increase in the probability or consequence of an accident previously 
evaluated?
    The proposed Technical Specification amendment does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated. Chapters 6 and 15 of the [Palo Verde 
Nuclear Generating Station] PVNGS [Updated Final Safety Analysis 
Report] UFSAR have been reviewed to address the impact of these 
changes (1500 plugged tubes and a pressurizer code safety valve 
nominal lift setpoint of 2475 psia) on accident consequences. For 
most of the events that were previously analyzed in the UFSAR, the 
proposed change does not have a significant affect or adversely 
impact the accident analysis. For RCS [reactor coolant system] 
pressure peaking events, Loss of Condenser Vacuum (LOCV) and 
Feedwater Line Breaks (FLB), a new analysis was performed to justify 
the acceptability of the changes.
    For the LOCV event (anticipated operational occurrence), the 
reanalysis determined that the peak RCS pressure, assuming 1500 
plugged tubes and a pressurizer code safety valve nominal lift 
setpoint of 2475 psia, is 2728 psia. The maximum reactor coolant 
system (RCS) pressure reached for this event as described in UFSAR 
Section 15.2.3 is 2742 psia. Therefore, this change is bounded by 
the reference cycle (UFSAR analysis) and remains below the 110% 
(2750 psia) design pressure limit.
    Several FLB scenarios are analyzed in support of PVNGS Unit 2 
operation. The scenario with the highest system pressures is the 
large FLB with a loss of alternating current (LOAC). For the large 
FLB with a LOAC event (limiting fault event), assuming 1500 plugged 
tubes and a pressurizer code safety valve nominal lift setpoint of 
2475 psia, is 2813 psia. The maximum RCS pressure reached for this 
event as described in UFSAR Section 15.2.8 is 2843 psia. The 
analysis shows that the RCS peak pressure for the large FLB with a 
LOAC (very low probability) event remains below the required value 
of 120% (3000 psia) of design pressure. Therefore, the analyses and 
reviews of the RCS pressure peaking events determined that the UFSAR 
design pressure limit is still bounding with this change. That is, 
the RCS design pressure limit will not be exceeded. Also, safety 
valves are accident mitigating devices and do not contribute to the 
probability of an event.
    Standard 2 -- Does the proposed change create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed Technical Specification amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated. The analyses and reviews show that 
the current licensing basis remains valid for this change (UFSAR 
design pressure limit is still bounding with this change). Safety 
valves are accident mitigating devices and do not contribute to the 
possibility of an accident. The pressurizer code safety valves are 
not manually or remotely operated, but are designed to automatically 
open to provide overpressure protection for pressure peaking events. 
The change in the pressurizer code safety valve setpoint to 2475 
psia does not significantly increase the probability of a 
pressurizer code safety valve opening, since the pressure is still 
well above the Technical Specification Table 2.2-1 reactor trip 
setpoint of 2383 psia for high pressurizer pressure.
    Standard 3 -- Does the proposed change involve a significant 
reduction in a margin of safety?
    The proposed Technical Specification amendment does not involve 
a significant reduction in a margin of safety. The analyses and 
reviews show that the limits in the licensing and design basis are 
still valid with this change. The analyses show that the RCS peak 
pressure remains below the 110% (2750 psia) design pressure limit 
for the LOCV event and remains below the required value of 120% 
(3000 psia) of design pressure RCS peak pressure for the large FLB 
with a LOAC (very low probability) event. The analyses 
[[Page 497]] and reviews of the RCS pressure peaking events 
determined that the UFSAR design pressure limit is still bounding 
with this change. Therefore, the proposed Technical Specification 
amendment maintains the margin of safety to the design pressure 
limit.
    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 request involves 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


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

    Date of amendments request: November 16, 1994Description of 
amendments request: The proposed revision to the Technical 
Specifications (TS) would change the Technical Specification 3/4.6.2 to 
remove the specific instrumentation requirements for monitoring of the 
suppression chamber average water temperature. Also, the proposed 
revision would change the TS Bases 3/4.6.2 to indicate the methods that 
are acceptable for determining suppression chamber average water 
temperature.Proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change maintains the same number of monitored 
locations from which an average suppression chamber water 
temperature can be derived, while making available additional valid 
RTD [resistance temperature detector] inputs from what was the 
redundant channel. No safety-related equipment, safety function or 
plant operation will be altered as a result of the proposed change. 
The SPTMS [suppression chamber temperature monitoring system] is 
neither an accident initiator nor does it provide any automatic 
accident mitigation function. The change does not affect the design, 
materials, or construction standards applicable to the suppression 
chamber average water temperature monitoring instrumentation.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The fundamental function and objective of the system is not 
affected by the proposed change. As stated above, no safety-related 
equipment, safety function or plant operations will be altered as a 
result of the proposed change. The change does not affect the 
design, materials, or construction standards applicable to the 
suppression chamber average water temperature instrumentation.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change allows the substitution of a qualified RTD 
already installed at a monitored location to insure the suppression 
chamber average water temperature remains valid. It does not involve 
any changes to the plant design or operation, therefore, no margins 
of safety, as defined by the plant's accident analyses, are 
impacted. Deletion of the defined instrument channels will not 
affect the ability to verify the suppression chamber ``average'' 
water temperature is being maintained below the maximum average 
temperatures required by the specification. This will insure the 
suppression chamber is Operable and able to perform its intended 
safety function.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: December 12, 1994
    Description of amendment request: The requested change would revise 
the containment spray (CS) nozzle surveillance interval from 5 to 10 
years.
    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 requested change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The requested change extends the surveillance interval for 
performance of qualitative flow testing of the CS nozzles. A 
revision to this surveillance interval can in no way increase the 
probability of any accident previously evaluated.
    Containment spray nozzle testing is not intended to track 
degradation of equipment by monitoring or trending performance. 
Rather, this surveillance constitutes a test of the passive design 
of the spray nozzles, i.e., it merely demonstrates whether the 
nozzles are or are not blocked or clogged. Based upon industry and 
plant-specific operating experience, a single failure rendering a 
significant number of nozzles inoperable as a result of blockage is 
considered highly unlikely. Since the reliability or functioning of 
the spray nozzles will not be affected by the revised surveillance 
interval, the consequences of any accident previously evaluated will 
not be increased. The requested change does not affect the physical 
design or operation of the plant, does not alter assumptions 
contained within the Updated Final Safety Analysis Report, and will 
not affect other Technical Specifications that preserve safety 
analysis assumptions. Therefore, operation of the facility in 
accordance with the requested change will not involve a significant 
increase in the consequences of any accident previously evaluated.
    2. The requested change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested change extends the surveillance interval for 
performance of qualitative flow testing of the CS nozzles. This 
change in the spray nozzle surveillance interval will not change or 
affect the physical plant or the modes of plant operation defined 
within the facility Operating License. This change does not involve 
the addition or modification of plant equipment, nor does it alter 
the design or operation of plant systems. Therefore, operation of 
the facility in accordance with the requested change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The requested change does not involve a significant reduction 
in the margin of safety.
    The requested change extends the surveillance for performance of 
qualitative flow testing of the CS nozzles. This revised 
surveillance interval will not change or otherwise influence the 
degree of operability assumed for the CS system within the plant 
safety analyses. As demonstrated by plant-specific and industry 
experience, an operational failure of the containment spray nozzles 
is considered highly unlikely. Since prior testing has demonstrated 
proper functioning of the CS spray nozzles, and operational single-
failures are considered highly unlikely, a reduction in testing 
frequency should not affect the ability of the CS system to mitigate 
the affects of a large loss-of-coolant or steam release accident. 
[[Page 498]] Therefore, operation of the facility in accordance with 
the requested change will not result in a significant reduction in 
the margin or safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

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

    Date of amendment request: October 24, 1994
    Description of amendment request: The proposed amendments would 
restructure the primary containment integrity and primary containment 
leakage technical specifications (TS) to reduce the repetition of those 
requirements contained in NRC regulations such as Appendix J to 10 CFR 
50. The amendments also support proposed exemptions from Appendix J 
requirements related to the scheduling of containment integrated leak 
rate tests (CILRT). In addition to the restructuring and scheduling 
changes, the proposed amendments incorporate (1) the relocation of the 
list of primary containment isolation valves in accordance with Generic 
Letter 91-08, ``Removal of Component Lists from Technical 
Specifications,'' and (2) a revision of the interval for functional 
testing of hydrogen recombiners from 6 months to 18 months in 
accordance with Generic Letter 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation.''
    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 of the 
following:
    a. The relocation of Technical Specification 3/4.6.1.2, Primary 
Containment Leakage, and Surveillance Requirements 4.6.1.1.a, 
4.6.4.3, and 4.6.6.1.d to specification 3/4.6.1.1, Primary 
Containment Integrity, as Surveillance Requirement 4.6.1.1.b 
continues to assure that Primary Containment leakage is maintained 
within the analyzed limit assumed for accident analysis by testing 
in accordance with 10 CFR part 50, Appendix J as modified by 
approved exemptions.
    The requirement to be less than 0.75 La for as-left Type A 
test and less than 0.60 La for Type B and C tests prior to 
first unit startup following testing performed in accordance with 10 
CFR part 50, Appendix J, as modified by approved exemptions, 
provides margin for degradation between tests and thus primary 
containment integrity is maintained during the time period between 
required leakage testing. The current Limiting Condition for 
Operation 3.6.1.2 in conjunction with Surveillance Requirements 
4.6.1.2 basically require the same leakage limits as proposed 
Surveillance Requirement 4.6.1.1.b. The Limiting Condition for 
Operation (LCO) is required to be less than 1.0 La and is 
applicable during a fuel cycle for the Type A test. The LCO for Type 
B and C combined leakage total is currently required to be less than 
0.60 La. The proposed Surveillance Requirement maintains the 
following:
    1.The current LCO for Overall Containment leakage (as determined 
by a Type A test) and for the Type B and C combined leakage during 
the cycle by requiring overall containment leakage to be less than 
1.0 La and Type B and C leakage total less than 0.60 La.
    2. The associated limits specified in the current Action 
Statements are maintained by verifying Overall Containment leakage 
to be less than 0.75 La and Type B and C leakage total less 
than 0.60 La prior to startup from an outage in which the 
applicable leakage testing is conducted.
    Therefore, there is no change to the consequences of an accident 
previously evaluated, because maintaining leakage within the 
analyzed limit assumed for accident analysis does not change either 
the onsite or offsite dose consequences resulting from an accident. 
In addition to this, containment leakage is not an accident 
initiator, so there is no effect on the probability of accident 
initiators. Thus there is no increase in the probability of an 
accident previously analyzed.
    b. Relocation of Technical Specification table of Primary 
Containment Isolation Valves, Table 3.6.3-1, to the LaSalle UFSAR is 
an administrative change to remove the component list of Primary 
Containment Isolation Valves, Table 3.6.3-1, from the Technical 
Specifications. The Limiting Condition for Operation (LCO), 3.6.3, 
is being revised to define which components the LCO applies to. The 
wording of the revised LCO encompasses all of the components listed 
in the current Technical Specification Table 3.6.3. Removal of this 
component list does not change the probability of any accident 
initiators or change any other relevant initial assumptions. Also, 
there is no change to the consequences of an accident previously 
evaluated, because removing this list from Technical Specifications 
does not change either the onsite or offsite dose consequences 
resulting from the event. The component list will be controlled by 
an Administrative Procedure and can only be changed by the 10 CFR 
50.59 change process with review and approval per the Onsite Review 
and Investigative Function. Therefore, there is no increase in 
either the probability or consequences of an accident previously 
evaluated.
    c. The change in the functional test interval for the Drywell 
and Suppression Chamber Hydrogen Recombiner systems from ``once per 
6 months'' to ``once per 18 months'' was determined by the NRC in 
NUREG 1366 and Generic Letter 93-05 to be acceptable by evaluation 
of the industry Licensing Event Reports (LERs) to assess the 
reliability of hydrogen recombiners. The conclusion was that the 
interval should be changed, because of the redundancy and apparent 
high reliability. A review of LaSalle LERs has shown only one LER 
that involved the operability of the hydrogen recombiner system and 
that was due to a Part 21 issue regarding circuit breaker 
environmental qualification. The breakers were replaced with 
qualified breakers. Therefore, the LaSalle Hydrogen Recombiner 
reliability is consistent with or better than that found by the NRC 
in determining this surveillance interval extension based on all 
LERs. Also, redundancy is the same as that assumed by the NRC; 
because, LaSalle has two hydrogen recombiner subsystems that are 
shared by Unit 1 and Unit 2. Both hydrogen recombiners subsystems 
are required to be Operable for either or both units in Operational 
Conditions 1 and 2. Based on LaSalle operating experience, the 
hydrogen recombiner subsystems are expected to continue to be 
demonstrated operable when the functional test is performed at an 18 
month frequency.
    Therefore, there is minimal or no change to the consequences of 
an accident previously evaluated, because at least one of the 
hydrogen recombiner subsystems is expected to be available to meet 
its design function to reduce the potential for hydrogen explosion 
or hydrogen burn in the primary containment. By preserving the 
integrity of the primary containment, there is no change to either 
the onsite or offsite dose consequences resulting from an accident. 
In addition to this, control of hydrogen concentration by use of a 
hydrogen recombiner subsystem is not an accident initiator, so there 
is no effect on the probability of accident initiators. Thus there 
is no significant increase in the probability of an accident 
previously analyzed.
    d. The first exemption request is from the requirements of 
paragraph III.A.6(b) of Appendix J to allow LaSalle County Station 
Unit Two to return to or resume a Type A test schedule of three 
times in ten years (40 plus or minus 10 months). Due to consecutive 
failures, 10 CFR 50 Appendix J requires that Type A tests be 
performed every refueling outage on Unit Two until two consecutive 
Type A tests are satisfactory. 10 CFR Part 50 has an exemption 
process and is specified in 10 CFR Part 50.12(a), which states:
    ``The Commission may, upon application by any interested person 
or upon its own [[Page 499]] initiative, grant exemptions from the 
requirements of the regulations of this part,...''
    The exemption process requires showing that the granting of the 
exemption is authorized by law, will not present an undue risk to 
the public health and safety, and is consistent with the common 
defense and security. Also, special circumstances are required to be 
present for the granting of an exemption. One of the special 
circumstances that would apply in this instance is 10 CFR part 
50.12(a)(2)(ii) which states:
    ``Application of the regulation in the particular circumstances 
would not serve the underlying purpose of the rule or is not 
necessary to achieve the underlying purpose of the rule''.
    This requires that it be shown that unacceptable containment 
leakage will be identified and corrected, by alternative methods. 
The alternative method is specifically Type B and C tests, which 
will identify any local penetration leakage. This is acceptable, 
because Type C test failures have been the cause for failures of as-
found Type A tests in the LaSalle Unit 2 first, third, and fourth 
refueling outages.
    Exceeding the allowable leakage rate during the performance of 
the Type A test is indicative of either a passive or a structural 
component that is leaking or that there is an inadequacy in the 
Local Leak Rate Test (Type B and C tests) program. When the failure 
of a Type A test is due to a passive or structural component, the 
only test for adequate repair would be the Type A test. For a Local 
Leak Rate Test program inadequacy, the Type A test would serve as a 
means of verification of the results of the test program. The Type A 
tests have not found new significant Type B or C tested local 
penetration leakage that has not been identified by Type B or C 
testing alone. Therefore, the LaSalle Local Leak Rate Test program 
is adequate to find and correct Type B and C containment penetration 
leakage.
    When it is determined that Type A tests failed as a direct 
result of as-found Type B and C minimum path leakage penalty 
additions and not due to a non Type B or C tested components or 
structures, then performance of the Type A test more frequently as 
required by 10 CFR Part 50, Appendix J, due only to Type B and C 
test failures is redundant to the performance of Type B and C tests. 
Therefore, Type B or C tested penetration leakage that can be 
determined by Type B or C tests is evaluated and corrected, as 
applicable, to maintain overall containment leakage within limits, 
without an additional Type A test.
    Primary Containment leakage which includes the minimum path 
Primary Containment Isolation Valve leakage is an assumption in any 
analyzed accident which could involve an offsite radioactive 
release. Because performance of Type B and C tests will find and 
allow correction/repair of leaking valves/penetrations, verification 
of as-found and as-left local leakage assures that Primary 
Containment leakage will be within the analyzed limit assumed for 
accident analysis.
    Therefore, for this one-time exemption for LaSalle Unit 2, there 
is little or no increase in the consequences of an accident 
previously evaluated involving the dose previously calculated either 
onsite or offsite at the site boundary due to any analyzed accident. 
In addition to this, containment leakage is not an accident 
initiator, so there is no effect on the probability of accident 
initiators. Thus there is no significant increase in the probability 
of an accident previously analyzed.
    e. The request for a partial exemption from paragraph III.D of 
Appendix J to 10 CFR 50 involves a deletion of the requirement to 
perform the third Type A test for each 10-year service period during 
the shutdown for the 10-year plant inservice inspections. There is 
no significant benefit in coupling these two surveillances (i.e., 
the Type A test and the 10-year ISI program). Each of the two 
surveillances is independent of the other and provides assurance of 
different plant characteristics. The Type A test assures the 
required leak-tightness for the reactor containment building be less 
than Appendix J acceptance criteria. This demonstrates compliance 
with the guidelines of 10 CFR Part 100 based on the assumptions used 
in the UFSAR which conform to NRC Safety Guide 4. The 10-year ISI 
program provides assurance of the integrity of the plant structures, 
systems, and components in compliance with 10 CFR 50.55(a). There is 
no safety-related concern necessitating their coupling to the same 
refueling outage. As a result, this change cannot increase the 
consequences (i.e., offsite dose) of any accident previously 
evaluated. Furthermore, since the decoupling of the test schedules 
has no affect on the test's effectiveness, decoupling their 
schedules will not increase the probability of an accident.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    a. Technical Specification 3/4.6.1.2, Primary Containment 
Leakage, and Surveillance Requirements 4.6.1.1.a, 4.6.4.3, and 
4.6.6.1.d are being relocated to specification 3.4.6.1.1, Primary 
Containment Integrity, as Surveillance Requirement 4.6.1.1.b. The 
proposed Surveillance Requirement 4.6.1.1.b assures that Primary 
Containment leakage is maintained within the analyzed limit assumed 
for accident analysis by testing in accordance with 10 CFR part 50, 
Appendix J as modified by approved exemptions. Primary containment 
leakage is an assumption in accident analyses, and is maintained by 
both the current specifications and the proposed specification. The 
leakage does not cause an accident and no new failure modes are 
created. Therefore this request for exemption does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    b. This is an administrative change to control the list of 
Primary Containment Isolation Valves outside the LaSalle Unit 1 and 
Unit 2 Technical Specifications. The administrative controls 
provided to control this component list assure that the design and 
operation of the plant will continue to be in accordance with the 
UFSAR, Facility License and the associated Technical Specifications. 
Therefore, the possibility of a new or different kind of accident 
from any previously evaluated is not created.
    c. The change in the functional test interval for the Drywell 
and Suppression Chamber Hydrogen Recombiner systems from ``once per 
6 months'' to ``once per 18 months'' is based on good equipment 
performance on a 6 month frequency. The expected outcome of the 18 
month surveillances, based on the low failure rate at a six month 
frequency, is to show the hydrogen recombiner subsystems Operable. 
This system is for mitigating the consequences of an accident that 
causes generation of hydrogen and oxygen in the primary containment. 
No new failure modes are created by this change in surveillance 
frequency. Therefore, the possibility of a new or different kind of 
accident from any previously evaluated is not created.
    d. The first exemption is from the requirements of paragraph 
III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to 
return to or resume a Type A test schedule of three times in ten 
years (40 plus or minus 10 months). Containment leakage testing, 
including both Type B and C testing and Type A testing as specified 
in the LaSalle County Station Safety Analysis Report were evaluated 
in Section 6.2.6 of Safety Evaluation Report, NUREG-0519, and found 
to be acceptable. Since Type B and C testing will find and verify 
correction of penetration leakage when Type B and C test as-found 
penalties are specifically what caused the failure of the as-found 
Type A tests, then Type B and C testing will provide adequate 
assurance of the continued integrity of the Primary Containment 
without increasing the frequency of Type A tests. As a result, the 
Primary Containment will continue [to] be maintained as designed and 
previously evaluated.
    Based on this, the requirement of two acceptable as-found Type A 
tests prior to returning to the Appendix J paragraph III.D frequency 
of three times in ten years (40 plus or minus 10 months) is not 
necessary to assure that the primary containment remains within the 
analyzed leakage limits. Containment leakage is an assumption for 
the dose consequences of accident analyses, and not an accident 
initiator. Also, no new failure modes are created by this exemption. 
Therefore this Amendment does not create the possibility of a new or 
different kind of accident.
    e. The request for a partial exemption from paragraph III.D of 
Appendix J to 10 CFR 50 involves a deletion of the requirement to 
perform the third Type A test for each 10-year service period during 
the shutdown for the 10-year plant inservice inspections. The 
proposed exemption does not involve any change to the plant design 
or operation. As discussed above, this change cannot increase the 
consequences of any accident previously evaluated. As a result, no 
new failure modes are created. Therefore, this proposed change 
cannot create the possibility of any new or different kind of 
accident from any accident previously evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    a. Technical Specification 3/4.6.1.2, Primary Containment 
Leakage, and Surveillance Requirements 4.6.1.1.a, 4.6.4.3, 
[[Page 500]] and 4.6.6.1.d are being relocated to specification 3/
4.6.1.1, Primary Containment Integrity, as proposed Surveillance 
Requirement 4.6.1.1.b. The proposed Surveillance Requirement 
4.6.1.1.b continues to assure that Primary Containment leakage is 
maintained within the analyzed limit assumed for accident analysis 
by testing in accordance with 10 CFR part 50, Appendix J as modified 
by approved exemptions.
    As stated in 1)a. above, the proposed Surveillance Requirement 
4.6.1.1.b maintains the acceptance criteria and limits for continued 
operation of the current specification for primary containment 
leakage. Therefore, the margin of safety is not reduced by this 
change. Also, the proposed addition of a definition for the maximum 
allowable primary containment leakage rate assures that the margin 
of safety is maintained.
    The leakage limits for MSIVs and hydrostatically tested valves 
are maintained by relocating the current surveillance requirements 
to specification 3/4.6.3, with the acceptance criteria of the 
current specification retained. Thus preserving the current margin 
of safety by maintaining the leakage rates as assumed in the 
accident analyses.
    b. The Limiting Condition for Operation for Technical 
Specification 3.6.3, Primary Containment Isolation Valves, is 
revised by this Technical Specification change to specifically 
define the components to which the LCO applies. Therefore, removal 
of Technical Specification Table 3.6.3-1, which lists the specific 
components to which the LCO applies does not change the scope or 
applicability of the specification. The component list will be 
controlled administratively with any changes to the list made in 
accordance with the 10 CFR 50.59 change process. Therefore, this is 
an administrative change only and there is no reduction in the 
margin of safety.
    c. The change in the functional test interval for the Drywell 
and Suppression Chamber Hydrogen Recombiner systems from ``once per 
6 months'' to ``once per 18 months'' is based on good equipment 
performance on a 6 month frequency. The expected outcome of the 18 
month surveillances, based on the low failure rate at a six month 
frequency, is to show the hydrogen recombiner subsystems Operable. 
The change in frequency has no affect on the hydrogen or oxygen 
generation assumptions or the recombination rate of the hydrogen 
recombiner subsystems. Therefore, the margin of safety is not 
reduced or changed by this surveillance interval change.
    d. The first exemption is from the requirements of paragraph 
III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to 
return to or resume a Type A test schedule of three times in ten 
years (40 plus or minus 10 months). The limit of total leakage 
determined from Type B and C tests will remain the same, providing a 
margin of 40 percent to the maximum allowable containment leakage 
rate (La) at the design basis accident pressure specified in 
proposed Technical Specification definition of La. This 40 
percent is as specified by 10 CFR Part 50, Appendix J. In addition 
to this, administrative guidelines have been set for each 
penetration/valve, so that any abnormal leakage will be corrected by 
adjustment or repair as needed. Any postponement of repairs is based 
on a technical evaluation and then only if the total Type B and Type 
C leakage is maintained at less than 0.60 La. Repairs will be 
required to restore the leakage rate to less than the administrative 
limit at the next refueling outage.
    This request for exemption is based the fact that Type B and C 
testing minimum path leakage rate penalties are the direct cause of 
the failure of as-found Type A tests. The leakage through Type B and 
C tested penetrations is best measured and corrected via a local 
leak test. Therefore, verification of an adequate margin of safety 
is assured by conducting Type B and C tests, and not another 
increased frequency Type A test.
    e. The request for a partial exemption from paragraph III.D of 
Appendix J to 10 CFR 50 involves a deletion of the requirement to 
perform the third Type A test for each 10-year service period during 
the shutdown for the 10-year plant inservice inspections. The 
proposed exemption does not change the acceptance criteria that must 
be met for inservice inspections, does not relax the condition of 
containment that must be met prior to plant restart, and does not 
change the requirements that must be met between plant refueling 
outages. Therefore, the proposed change does not result in 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: November 21, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to allow a one-time extension of 
the allowed outage time for an inoperable reserve source of offsite 
power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability of occurrence or consequences of an 
accident previously evaluated.
    The proposed changes will extend the allowed outage time for the 
Reserve source of off-site power, on a one time basis, to allow the 
installation of high speed protective relays on the unit system 
auxiliary transformers which will increase the level of protection 
from ground faults on the low voltage (secondary) side of the 
transformers. Operation of Zion, Units 1 and 2, in accordance with 
the proposed requirements will not affect the initiators or 
precursors of any accident previously evaluated. Operation in 
accordance with the proposed requirements will not increase the 
likelihood that a transient initiating event will occur because 
transients are initiated by equipment malfunction and/or 
catastrophic system failure. As a result, the probability of 
occurrence of accidents previously evaluated is not significantly 
increased.
    During the [system auxiliary transformer] SAT outage, power to 
the shut down unit will be provided by backfeeding off-site power 
through the unit main power transformers and the UAT to supply the 
unit non-essential 4-KV service buses. Emergency on-site power will 
be available to the shut down unit from at least one unit specific 
[emergency diesel generator] EDG when fuel is in the reactor core. 
This will ensure that at least one train of Residual Heat Removal 
(RHR) will have an emergency source of AC power at all times. RHR 
Train A is powered by ESF bus 149(249) which can be energized by the 
1B(2B) EDG during a loss of off-site power. RHR Train B is powered 
by bus 148(248) which can be energized by the 1A(2A) EDG. Because 
the 'O' EDG must be operable for the operating unit, it will also be 
available to energize the Division 7 ESF bus on the shut down unit. 
The 'O' EDG can supply buses 147 and 247 simultaneously if the need 
should arise during an emergency.
    Power to the operating unit (opposite unit) will be provided by 
the SAT and the UAT in the normal at-power configuration. Emergency 
on-site will be provided by the two unit specific EDGs (A and B) and 
the common 'O' EDG. In accordance with the proposed requirements, 
the Reserve source of off-site power will not be removed from 
service unless all three EDGs are operable and the normal source of 
off-site power is operable. Administrative controls will be in place 
to limit activities in the switchyard that could impact the 
reliability of the remaining source of off-site power to the unit.
    The Zion PRA was used to compare the impact of extending the 
action time versus the impact of manual reactor shutdown on core 
damage probability. The PRA result concluded that the risk of 
continuing to operate the operating unit for an additional 11 days 
with the shutdown unit's SAT out of service is not significantly 
greater than the risk of manually shutting down the operating unit 
at the expiration of the current 72 hour action statement and is not 
significant when compared to the total core damage probability in a 
year.
    The revised surveillance requirements will provide additional 
assurance that redundant sources of power are maintained operable 
while the reserve source of off-site power is 
[[Page 501]] unavailable. The ability to safely shut down the 
operating unit and mitigate the consequences of all accidents 
previously evaluated will be maintained. The reserve source of off-
site power is not relied upon in any design basis accident. 
Therefore, based on the previous discussion, the proposed changes do 
not involve a significant increase in consequences of any accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any previously analyzed.
    The proposed changes to the Technical Specifications do not 
involve the addition of any new or different types of safety-related 
equipment, nor does it involve the operation of equipment required 
for safe operation of the facility in a manner different from those 
addressed in the safety analysis. No safety related equipment or 
function will be altered as a result of the proposed changes. Also, 
the procedures governing normal plant operation and recovery from an 
accident are not changed by the proposed Technical Specification 
changes. The proposed changes will extend the allowed outage time 
for the Reserve source of off-site power, on a one-time basis, to 
allow the installation of high speed protective relays on the unit 
system auxiliary transformers which will increase the level of 
protection from ground faults on the low voltage (secondary) side of 
the transformers. The addition of the high speed relaying has been 
evaluated pursuant to 10 CFR 50.59, and no unreviewed safety 
questions were identified.
    Requirements will be modified to require additional assurance 
that the remaining off-site source of AC power and the on-site 
source of emergency (emergency diesel generators) are OPERABLE. 
Since no new failure modes or mechanisms are added by the proposed 
changes, the possibility of a new or different kind of accident is 
not created.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes will extend the allowed outage time for the 
reserve source of off-site power, on a one-time basis, to allow for 
installation of high speed protective relays on the unit system 
auxiliary transformers which will increase the level of protection 
from ground faults on the low voltage (secondary) side of the 
transformers.
    During the SAT outage, power to the operating unit (opposite 
unit) will be provided by the unit SAT and the UAT in the normal 
configuration. Emergency on-site power will be provided by the two 
unit specific EDGs (A and B) and the common 'O' diesel generator. 
Because the accident analyses take no credit for offsite power 
availability, this temporary degradation will not impact the 
analysis results.
    No safety system setpoints are changed by this proposal. There 
is no impact on any physical design margins, and no analytical 
results are affected by this change. The revised surveillance 
requirements will provide additional assurance that redundant 
sources of power are maintained operable while the Reserve source of 
off-site power is unavailable.
    Based on the above discussion, the ability to safely shut down 
the operating unit and mitigate the consequences of all accidents 
previously evaluated will be maintained. Therefore, the margin of 
safety is not significantly affected.
    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: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: October 5, 1994
    Description of amendment request: The proposed amendment would (1) 
revise primary coolant system (PCS) pressure-temperature (P-T) limits, 
power-operated relief valve (PORV) setting limits, and primary coolant 
pump starting limits to accommodate reactor vessel fluence for an 
additional 4 effective full power years (up to 2.192 x 1019nvt). 
The existing P-T limit curves are calculated for a fluence of 1.8 x 
1019 could be reached as early as March 1, 1995; (2) require the 
high pressure safety injection (HPSI) pumps to be ``rendered incapable 
of injection into the PCS'' when the PCS is below 300 deg.F, rather 
than the existing requirement to render both HPSI pumps ``inoperable'' 
when the PCS is below 260 deg.F. This change supports the assumption in 
the P-T limit analyses that HPSI injection would not occur below 
300 deg.F; and (3) establish a more restrictive limit on pressurizer 
heatup rate to achieve consistency between design assumptions and 
technical specification (TS) limits. The limit in the existing TS is 
less restrictive than used in design calculations. Neither the design 
heatup rate nor the TS heatup rate limit is achievable with installed 
equipment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The following evaluation supports the finding that operation of 
the facility in accordance with the proposed Technical 
Specifications would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The revision of the Primary Coolant Pump [PCP] starting limits, PCS 
P-T curves, and PORV setting limits would not cause any changes to the 
capability or operation of plant systems that would affect the 
probability of occurrence or consequences of an accident. These 
revisions simply update the existing requirements to account for 
additional reactor vessel fluence.
    The reduction of the allowable pressurizer heatup rate would 
have no effect on operation of the plant. The current limit is 
physically unobtainable with installed equipment. The proposed 
change better aligns the Technical Specification limits with the 
design analysis. The change in the pressurizer heatup rate limit 
will not increase the probability or consequences of an accident.
    Requiring the HPSI pumps to be operable when above 325 deg.F, 
rather than when above 300 deg.F does not affect the probability or 
consequences of any accident previously evaluated. Neither the 
existing 300 deg.F requirement nor the proposed 325 deg.F 
requirement has an analytical base. This requirement was recently 
changed from 325 deg.F to 300 deg.F simply for uniformity. With the 
revised P-T limit analysis requirement to assure that inadvertent 
HPSI injection will not occur below 300 deg.F, it is necessary to 
revert to the former limit of 325 deg.F to provide time to 
transition between these two contrasting HPSI pump requirements.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The revised specifications, PCP starting limits, PCS P-T limits, 
pressurizer heatup rate, PORV setting limits, and HPSI pump 
restrictions, all are directly related to, and intended to prevent, 
a previously analyzed event, failure of the Reactor Coolant Pressure 
Boundary. Revision of these limits would not create the possibility 
of a new or different kind of accident.
    3. Involve a significant reduction in a margin of safety.
    The revised PCP starting limits, PCS P-T limits, and PORV 
setting limits are calculated using a similar methodology as the 
limits which they replace. Therefore they provide the same margin of 
safety.
    The revised pressurizer heatup rate reduces the currently 
allowable limit which is in the direction of increased margin of 
safety. Since there is no equipment installed which would cause 
either the existing or the proposed limit to be reached, there will 
be no change on the operation of the plant equipment. Therefore 
reducing the limit on the pressurizer heatup rate will not involve a 
significant reduction in the margin of safety.
    Requiring the HPSI pumps to be operable when above 325 deg.F, 
rather than when above 300 deg.F does not involve a significant 
reduction in any margin of safety. Neither the existing 300 deg.F 
requirement nor the proposed [[Page 502]] 325 deg.F requirement has 
an analytical base. This requirement was recently changed from 
325 deg.F to 300 deg.F simply for uniformity. With the revised P-T 
limit analysis requirement to assure that inadvertent HPSI injection 
will not occur below 300 deg.F, it is necessary to revert to the 
former limit of 325 deg.F to provide time to transition between 
these two contrasting HPSI pump requirements.
    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: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: John N. Hannon

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

    Date of amendment request: November 8, 1994
    Description of amendment request: The proposed amendment revises 
technical specifications (TSs) associated with requirements for 
performing the containment integrated leak rate test (ILRT). The 
proposed change describes the ILRT test frequency by referencing the 
test frequency requirements included in 10 CFR Part 50, Appendix J. The 
existing specifications paraphrase the Appendix J requirements, but 
include differences that result in interpretation problems.
    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 proposed change revises Technical Specification 4.4.1.1.4 to 
reference the testing frequency requirements of 10 CFR 50, Appendix 
J, and to state that NRC approved exemptions to the applicable 
regulatory requirements are permitted. The current requirements of 
TS 4.4.1.1.4 paraphrase the requirements of Section III.D.1.(a) of 
Appendix J. The proposed administrative revision simply deletes the 
paraphrased language and directly references Appendix J. No new 
requirements are added, nor are any existing requirements deleted. 
An approved exemption to Section III.D.1.(a) of Appendix J would not 
necessarily affect the requirements of TS 4.4.1.1.4, unless the 
proposed clarification phrase permitting the use of approved 
exemptions is added. Any specific changes to the requirements of 
Section III.D.1(a) will require a submittal from Entergy Operations 
under 10CFR50.12 and subsequent review and approval by the NRC prior 
to implementation. The proposed change is stated generically to 
avoid the need for further TS changes if different exemptions are 
approved in the future.
    The proposed change, in itself, does not affect reactor 
operations or accident analysis and has no radiological 
consequences. The change provides clarification so that TS changes 
will not be necessary in the future to correspond to applicable NRC 
approved exemptions from the requirements of Appendix J. Therefore, 
this change 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 change provides clarification to a specification 
which paraphrases a codified requirement. Since the proposed 
amendment would not change the design, configuration or method of 
operation of the plant, it would 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.
    The proposed change is administrative and clarifies the 
relationship between the requirements of TS 4.4.1.1.4, Appendix J, 
and any approved exemptions to Appendix J. It does not, in itself, 
change a safety limit, an LCO, or a surveillance requirement on 
equipment required to operate the plant. The NRC will directly 
approve change proposed exemption to III.D.1.(a) of Appendix J prior 
to implementation. Therefore, this 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: 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

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, 
Appling County, Georgia

    Date of amendment request: December 2, 1994
    Description of amendment request: The proposed amendments would 
replace Appendix B, ``Environmental Technical Specifications'' with an 
Environmental Protection Plan (Nonradiological) and revise the 
Operating Licenses to reflect these changes. The proposed changes are 
administrative in nature, altering only the format and location of 
programmatic controls and procedural details relative to 
nonradiological environmental monitoring.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1) The proposed amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to the Environmental Technical 
Specifications (ETS) are administrative in nature, altering only the 
format and location of programmatic controls and procedural details 
relative to nonradiological environmental values. The proposed 
Environmental Protection Plan (EPP) (Nonradiological) contains the 
programmatic controls now residing in the ETS, with appropriate 
plant procedures serving as implementing documents. The proposed 
changes to the operating licenses are also administrative in nature 
and change the Appendix B reference from ETS to EPP. Compliance with 
applicable regulatory requirements will be maintained. In addition, 
the proposed changes do not alter the conditions or assumptions in 
any of the accident analyses. Therefore, these proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2) The proposed amendments do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

    The proposed changes to the ETS do not involve any change to the 
configuration or method of operation of any plant equipment. These 
proposed changes are administrative in nature and consist of 
replacing the ETS with an EPP. The proposed changes to the operating 
licenses are also administrative in nature and change the Appendix B 
reference from ETS to EPP. Accordingly, no new failure modes have 
been identified for any plant system or component important to 
safety nor has any new limiting single failure been identified as a 
result of the proposed changes. 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 amendments do not result in a significant 
reduction in the margin of safety.
    The proposed changes to the ETS relate primarily to matters 
involving recordkeeping, reporting, and administrative procedures or 
requirements. No significant change in the type or quantity of any 
effluent release will result from this action. These changes replace 
[[Page 503]] the ETS with an EPP. The proposed EPP contains the 
programmatic controls now residing in the ETS, with appropriate 
plant procedures serving as implementing documents to ensure 
compliance with applicable regulatory requirements. The proposed 
changes to the operating licenses are also administrative in nature 
and change the Appendix B reference from ETS to EPP. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: November 21, 1994
    Description of amendment request: The proposed amendment would 
eliminate the Main Steam Isolation Valve (MSIV) - Leakage Control 
System (LCS) including the primary containment isolation valves 
associated with the MSIV - LCS, along with increasing the allowable 
MSIV leakage rates.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    I. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to TS Section 3.6.1.2 do not involve a 
change to structures, components, or systems that would affect the 
probability of an accident previously evaluated. The TS limits for 
MSIVs are increased from 46 scf per hour for all four main steam 
lines to less than or equal to 100 scf per hour for any one MSIV and 
a combined maximum pathway leakage rate of less than or equal to 300 
scf per hour for all four main steam lines. The consequences of an 
accident are affected as discussed in this section.
    The proposed changes to TS Section 3.6.1.4 eliminate the Main 
Steam Isolation Valves (MSIVs) Leakage Control System (LCS) 
requirements from the TS. As described in Section 6.7 of the FSAR, 
the LCS is manually initiated in about 20 minutes following a design 
basis Loss of Coolant Accident (LOCA). Since the LCS is operated 
only after an accident has occurred, these proposed changes have no 
effect on the probability of an accident.
    Since MSIV leakage and operation of the LCS are included in the 
radiological analysis for the design basis LOCA as described in 
Section 15.6.5 of the FSAR, the proposed changes do not affect the 
precursors of other analyzed accidents. Analysis of the effects of 
the proposed changes do, however, result in acceptable radiological 
consequences for the design basis LOCA previously evaluated in 
Section 15.6.5 of the FSAR.
    SSES, Units 1 and 2 have an inherent MSIV leakage treatment 
capability as discussed below. We propose to use the drain lines 
associated with the main steam lines and main turbine condenser as 
an alternative to the guidance in Regulatory Guide 1.96, ``Design of 
Main Steam Isolation Valve Leakage Control System For Boiling Water 
Nuclear Power Plants'', Revision 0, May 1975, for MSIV leakage 
treatment. If approved, we will incorporate this alternate method in 
the appropriate operational procedures and Emergency Operating 
Procedures.
    The Boiling Water Reactor Owners' Group (BWROG) has evaluated 
the availability of main steam system piping and main condenser 
alternate pathways for processing MSIV leakage, and has determined 
that the probability of a near coincident LOCA and a seismic event 
is much smaller than for other plant safety risks. Accordingly, this 
alternate MSIV leakage treatment pathway is available during and 
after a LOCA. Nevertheless, the BWROG has also determined that main 
steam piping and main condenser design are extremely rugged, and the 
design requirements applied to SSES Unit 1 and Unit 2 main steam 
system piping and main condenser contain substantial margin, based 
on the original design requirements. Therefore, the alternate 
treatment method has been evaluated for its capability to mitigate 
the consequences of a LOCA, and has been evaluated to assure its 
availability considering a seismic event.
    In order to determine the capability of the main steam piping 
and main condenser alternate treatment pathway, the BWROG has 
reviewed earthquake experience data on the performance of non-
seismically designed piping and condensers during past earthquakes. 
The data is summarized in General Electric (GE) Report, ``BWROG 
Report for Increasing MSIV Leakage Rate Limits and Elimination of 
Leakage Control Systems,'' NEDC 31858P, Revision 2, submitted to the 
NRC by BWROG letter dated October 4, 1993. This study concluded that 
the possibility of a failure that could cause a loss of steam or 
condensate in Boiling Water Reactor (BWR) main steam piping or 
condensers in the event of a design basis (i.e., safe shutdown) 
earthquake is highly unlikely, and that such a failure would also be 
contrary to a large body of historical earthquake experience data, 
and thus unprecedented.
    A verification has been performed of the seismic adequacy of the 
Unit 1 and Unit 2 main steam piping and main condenser consistent 
with the guidelines discussed in Section 6.7 of NEDC-31858P, 
Revision 2, to provide reasonable assurance of the structural 
integrity of these components. An evaluation, including the walkdown 
report outliers, ``MSIV Leakage Alternate Treatment Method Seismic 
Evaluation,'' for Unit 1 and Unit 2, is attached. The results of the 
evaluation clearly demonstrate that the MSIV Leakage Alternate 
Treatment Method meets the intent of 10CFR100 Appendix A, with 
regards to seismic qualification. Except for the requirement to 
establish a proper flow path from the MSIVs to the condenser, the 
proposed method is passive and does not require any additional logic 
control and interlocks. The method proposed for MSIV leakage 
treatment is consistent with the philosophy of protection by 
multiple barriers used in containment design for limiting fission 
product release to the environment.
    A plant-specific radiological analysis has been performed in 
accordance with NEDC-31858P, Revision 2, to assess the effects of 
the proposed increase to the allowable MSIV leakage rate in terms of 
control room and off-site doses following a postulated design basis 
LOCA. This analysis utilizes the hold-up volumes of the main steam 
piping and condenser as an alternate method for treating the MSIV 
leakage. As discussed earlier, there is reasonable assurance that 
the main steam piping and condenser remain intact following a design 
basis earthquake. The radiological analysis uses standard 
conservative assumptions for the radiological source term consistent 
with Regulatory Guide (RG) 1.3, Assumptions Used for Evaluating the 
Potential Radiological Consequences of a Loss-Of-Coolant Accident 
for Boiling Water Reactor, Revision 2, dated April 1974.
    The analysis results demonstrate that dose contributions from 
the proposed MSIV leakage rate limit of 100 scfh per steam line, not 
to exceed a total of 300 scfh for all four main steam lines, and 
from the proposed deletion of the LCS, result in an insignificant 
increase to the LOCA doses previously evaluated against the 
regulatory limits for the off-site doses and control room doses 
contained in 10CFR100 and 10CFR50, Appendix A, General Design 
Criterion (GDC) 19, respectively. The off-site and control room 
doses resulting from a LOCA are discussed in Section 15.6.5 of the 
FSAR. The off-site and control room doses resulting from a LOCA 
associated with the proposed changes are the sum of LOCA doses 
evaluated in the power uprate revision to the design basis DBA-LOCA 
calculation (EC-RADN-1009) and the additional doses calculated using 
the alternate MSIV leakage treatment method. Enclosure 3 [of 
application dated November 21, 1994] summarizes the off-site and 
control room doses and compares the alternate treatment method doses 
to the original MSIV-LCS treatment method doses.
    The 30-day whole body doses at the Low Population Zone (LPZ) did 
not change and remained at .37 rem for the alternate treatment 
method. The 30-day control room whole body doses increased slightly 
from .38 [[Page 504]] rem to .76 rem for the alternate treatment 
method. The increase in control room dose is not significant since 
the revised doses are well below the regulatory limits, i.e., .76 
rem calculated versus the limit of 5 rem in the control room. The 
two-hour whole body dose at the Exclusion Area Boundary (EAB) 
decreased slightly from 2.47 rem to 2.217 rem.
    The 30-day thyroid dose at the LPZ increased from 30.4 rem for 
the MSIV-LCS treatment method to 41.74 rem for the alternate 
treatment method. This increase is not significant since the revised 
dose of 41.74 rem is well within the regulatory limit of 300 rem. 
The two-hour thyroid dose at the EAB decreased slightly from 127.8 
rem to 125.61 rem. The 30-day control room thyroid dose increased 
from 14.19 rem for the MSIV-LCS treatment method to 18.55 rem for 
the alternate treatment method. The increased control room thyroid 
dose is not significant since the revised dose remains well below 
the regulatory limit of 30 rem.
    The 30-day control room beta dose increased insignificantly from 
12 rem for the MSIV-LCS treatment method to 12.17 rem for the 
alternate treatment method, remaining a small fraction relative to 
the limit of 75 rem.
    In summary, the proposed changes discussed above do not result 
in a significant increase in the radiological consequences of a LOCA 
when the same assumptions and methods specified in the FSAR are 
used, recognizing that radiological consequences calculated in the 
FSAR and for these proposed changes are significantly higher than 
those using more realistic assumptions and methods. Nevertheless, 
the calculated off-site and control room doses resulting from a LOCA 
remain well below the regulatory limits.
    The proposed change to TS Table 3.6.3-1 deletes the LCS valves 
from the list of primary containment isolation valves. This proposed 
change is consistent with the proposed deletion of the LCS. The LCS 
lines that are connected to the main steam piping are welded and/or 
capped closed to assure primary containment integrity is maintained. 
The welding and post weld examination procedures will be in 
accordance with American Society of Mechanical Engineers (ASME) 
Code, Section III requirements. These welds and/or caps will be 
periodically tested as part of the Containment Integrated Leak Rate 
Test (CILRT). This proposed change does not involve an increase in 
the probability of equipment malfunction previously evaluated in the 
FSAR. In fact, this proposed change reduces the probability of 
equipment malfunction since, upon implementation of these proposed 
changes, the plant will be operated with less primary containment 
isolation valves subjected to postulated failure. This proposed 
change has no effect on the consequences of an accident since the 
LCS lines will be welded and/or cap closed, thus assuring that the 
containment integrity, isolation and leak test capability are not 
compromised.
    The proposed change to TS Table 3.8.4.2.1-1 deletes the LCS 
motor operated valves from the list of ``Motor Operated Valves 
Thermal Overload Protection - Continuous.'' The proposed change has 
no effect on the probability or consequences of an accident since 
the valves are eliminated and not performing a safety function.
    Therefore, as discussed above, the proposed changes do not 
involve a significant increase in the probability or consequences 
from any accident previously evaluated.
    II. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    As stated in Section I, the proposed changes do not involve a 
change to structures, components, or systems that would affect the 
probability of an accident previously evaluated , nor would these 
changes create any new or different kind of accident from any 
previously evaluated. The proposed changes will introduce and take 
credit for a new level of operational performance for existing plant 
systems and components to mitigate the consequences of the accident. 
The effect on this equipment has been evaluated and found to provide 
an acceptable level of reliability resulting in the required level 
of protection. This conclusion is based on the evaluation performed 
in NEDC 31858P, Revision 2, and the plant specific seismic 
evaluation provided in the Enclosure 2 [of application dated 
November 21, 1994], ``MSIV Leakage Alternate Treatment Method 
Seismic Evaluation.'' The Leakage Control System has been installed 
to direct any leakage past the MSIVs during the LOCA; acting after 
the accident has occurred. The resulting consequences of the 
evaluated accidents have been affected as discussed in Section I 
resulting in no significant increase in the probability or 
consequences of said accident. Therefore, reliance on different 
equipment than previously assumed to mitigate the consequences of an 
accident does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The BWROG evaluated MSIV performance and concluded that MSIV 
leakage rates up to 200 scfh per valve will not inhibit the 
capability and isolation performance of the MSIVs to effectively 
isolate the primary containment. Implementation of the proposed 
changes does not result in modifications which could adversely 
impact the operability of the MSIVs. The LOCA has been analyzed 
using the main steam piping and main condenser as a treatment method 
to process MSIV leakage at the proposed maximum rate of 100 scfh per 
main steam line, not to exceed 300 scfh total for all four main 
steam lines. Therefore, the proposed TS Section 3.6.1.2 change to 
increase the allowed MSIV leakage rate does not create any new or 
different kind of accident from any accident previously evaluated.
    The proposed TS Section 3.6.1.4 change to eliminate the LCS does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated because the removal of the 
LCS does not affect any of the remaining SSES Unit 1 and Unit 2 
systems, and the LOCA has been re-analyzed using the proposed 
alternate method to process MSIV leakage. The associated proposed 
change to delete the LCS isolation valves from TS Table 3.6.3-1 and 
Table 3.8.4.2.1-1 does not create the possibility of a new or 
different kind of accident. The affected main steam piping will be 
welded and/or capped closed to assure that the primary containment 
integrity, isolation, and leak testing capability are not 
compromised. The affected LCS motor operated valves will be 
eliminated so their thermal overloads will not need to be bypassed.
    Therefore, as discussed above, the proposed changes do not 
create the possibility for any new or different kind of accident 
from any accident previously evaluated.
    III. Involve a significant reduction in a margin of safety.
    The proposed change to TS Section 3.6.1.2 to increase the MSIV 
allowable leakage does not involve a significant reduction in the 
margin of safety. As discussed in the current Bases for TS Section 
3/4.6.1.2, the allowable leak rate limit specified for the MSIVs is 
used to quantify a maximum amount of leakage assumed to bypass 
primary containment in the LOCA radiological analysis. Accordingly, 
results of the re-analysis supporting these proposed changes are 
evaluated against the dose limits contained in 10CFR100 for the off-
site doses, and 10CFR50, Appendix A, GDC 19, for the control room 
doses. As discussed above, sufficient margin relative to the 
regulatory limits is maintained even when assumptions and methods 
(e.g., RG 1.3) that are considered highly conservative relative to 
more realistic assumptions and methods are used in the analysis.
    Results of the radiological analysis demonstrate that the 
proposed changes do not involve a significant reduction in the 
margin of safety. Whole body doses, in terms of margin of safety, 
are insignificantly reduced by .38 rem in the control room. The 
margin of safety remains constant for the LPZ whole body dose or 
actually increases by .253 rem for the EAB whole body dose. The 
margin of safety for thyroid dose category is reduced by 11.34 rem 
at the LPZ and 4.36 rem in the control room. The margin of safety is 
found to increase for the EAB thyroid dose by 2.19 rem. The margin 
of safety for beta dose is insignificantly reduced by .17 rem in the 
control room. The reductions in the margin of safety are not 
significant since the revised calculated doses are highly 
conservative yet remain well below the regulatory limits, and 
therefore, a substantial margin to the regulatory limits is 
maintained.
    The proposed change to eliminate the LCS from TS Section 3.6.1.4 
does not reduce the margin of safety, in fact, the overall margin of 
safety is increased. The function of the LCS for MSIV leakage 
treatment will be replaced by alternate main steam drain lines and 
condenser equipment. This treatment method is effective in reducing 
the dose consequences of MSIV leakage over an expanded operating 
range compared to the capability of the LCS and will, thereby, 
resolve the safety concern that the LCS will not function at MSIV 
leakage rates higher than the LCS design capacity. Except for the 
requirement to establish a proper flow path from the MSIVs to the 
condenser, the proposed method is passive and does not require any 
new logic control and interlocks. This proposed method is consistent 
with the [[Page 505]] philosophy of protection by multiple barriers 
used in containment design for limiting fission product release to 
the environment. Furthermore, as previously identified, based on the 
evaluations discussed in NEDC-31858P, Revision 2, and the seismic 
evaluation provided in the Enclosure 2 [of application dated 
November 21, 1994] report, ``MSIV Leakage Alternate Treatment Method 
Seismic Evaluation,'' the design of the MSIV leakage alternate drain 
pathway, meets the intent of the 10CFR100, Appendix A requirement 
for seismic qualification. Therefore, the proposed method is highly 
reliable and effective for MSIV leakage treatment.
    The revised calculated LOCA doses remain within the regulatory 
limits for the off-site and the control room. Therefore, the 
proposed method maintains a margin of safety for mitigating the 
radiological consequences of MSIV leakage for the proposed TS 
leakage rate limit of 100 scfh per main steam line, not to exceed a 
total of 300 scfh for all four main steam lines.
    The proposed change to delete LCS isolation valves from TS Table 
3.6.3-1 and Table 3.8.4.2.1-1 does not reduce the margin of safety. 
Welded and/or capped closure of the LCS lines assures that the 
primary containment integrity and leak testing capability are not 
compromised. These welds and/or caps will be periodically leak 
tested as part of the CILRT. The LCS motor operated valves will be 
eliminated so their thermal overloads will not need to be bypassed. 
Therefore, the proposed deletion of the LCS isolation valves does 
not involve a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: November 18, 1994
    Description of amendment request: The proposed change would revise 
the Reactivity Control System Technical Specification Limiting 
Conditions for Operation for boration flow paths and charging pumps by 
reducing the number of operable charging pumps required for boron 
addition in Mode 4 from two to one.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident or malfunction of equipment important to 
safety previously evaluated.
    The Emergency Core Cooling System (ECCS) requirements assume 
that only one charging pump will be available below 350 deg.F 
without single failure considerations on the bases of the stable 
reactivity condition of the reactor and limited core cooling 
requirements. Therefore, the Mode 4 Applicability has been deleted 
from LCOs 3.2.1.2 and 3.2.1.4, and was added to LCOs 3.2.1.1 and 
3.2.1.3 consistent with the requirements of LCO 3.5.3.
    The current Bases for the Unit 2 Technical Specification for 
boration system flow paths via the charging pumps supports the use 
of a similar LCO for Salem Unit 1.
    The limitation for a maximum of one centrifugal charging pump to 
be operable when the RCS temperature is less than or equal to 
312 deg.F has been added to LCO 3.1.2.3 for clarity and is 
consistent with the Cold Overpressure Protection (POPS) analysis and 
the requirements of Technical Specification 3.5.3.
    The requirements for Boric Acid Transfer Pump operability are 
adequately addressed in Technical Specifications 3.1.2.1 and 3.1.2.2 
which specify the boron injection flow paths to be operable and the 
components required to perform this function. This includes the 
availability of the transfer pumps to meet this Technical 
Specification requirement.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated in the UFSAR.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    As discussed in response to Question 1 above, the proposed 
amendment to the number of charging pumps required to be operable in 
Mode 4 is consistent with the current Technical Specification 
requirements for the ECCS LCO and the POPS. The current bases for 
the Unit 2 Technical Specification for boration system flow paths 
via the charging pumps supports the use of a similar LCO for Salem 
Unit 1. The requirements for Boric Acid Transfer Pump operability 
for Unit 1 are adequately addressed in Technical Specifications 
3.1.2.1 and 3.1.2.2 which specify the boron injection flow paths to 
be operable and the components required to be available to perform 
this function including the transfer pumps. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety. The proposed amendment to the number of charging pumps 
required to be operable in Mode 4 will not result in any changes to 
the assumptions or conditions for the current ECCS analysis and POPS 
analysis. The current bases for the Unit 2 Technical Specification 
for boration system flow paths via the charging pumps supports the 
use of a similar LCO for Salem Unit 1 (i.e., the Bases are 
essentially the same). The requirements for Boric Acid Transfer Pump 
operability for Unit 1 are adequately addressed in Technical 
Specifications 3.1.2.1 and 3.1.2.2 which specify the boron injection 
flow paths to be operable and the components required to be 
available to perform this function including the transfer pumps. 
Therefore, 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: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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: December 19, 1994Description of 
amendments request: The proposed change to Table 3.7-3 of the Technical 
Specifications includes the revision to the main steam safety valve 
(MSSV) setpoint tolerance from plus or minus 1 percent to plus or minus 
3 percent and modifies the bases to 3/4.7.1.1 to increase the relieving 
capacity of the MSSVs to at least 12,984,660 pounds per hour which 
corresponds to approximately 112 percent of total secondary steam flow 
at 100 percent rated thermal power. In addition, modifications to Table 
3.7-1 are proposed to reduce the allowable power range neutron flux 
high setpoints for multiple inoperable steam generator safety valves. 
The proposed amendment includes an editorial correction to Bases 3/
4.7.1.2 to indicate required auxiliary feedwater flow at ``1133 psia'' 
rather than ``1133 psig.''
    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 license amendment does not involve a significant 
increase in the [[Page 506]] probability or consequences of an 
accident previously evaluated.
    These proposed changes to the Farley Technical Specifications do 
not result in a condition where the design, material and 
construction standards of the MSSVs that were applicable prior to 
the proposed change are altered. The valves will continue to 
function as designed. All applicable safety analyses have been 
reviewed, evaluated or reanalyzed and all applicable safety criteria 
continue to be met. No accident sequences are altered because of the 
proposed amendment. The radiological consequences for the Steam 
Generator Tube Rupture were reanalyzed and 10 CFR 100 criteria 
continue to be met. All other FSAR radiological analyses remain 
bounding. Analyses have been performed to justify the proposed high 
nuclear flux setpoint changes. All acceptance criteria for these 
analyses continue to be met. Therefore, the proposed amendment does 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different accident from any accident 
previously evaluated.
    The MSSVs continue to have the required pressure relieving 
capacity to ensure that system design pressure remains below 110% of 
shell design pressure. The proposed changes are not accident 
initiators nor do they create any new accident scenarios or any new 
limiting single failures. The ability of the MSSVs to respond to an 
accident condition is not impaired by the proposed changes. The 
proposed high nuclear flux setpoints for multiple valves out of 
service ensure all applicable safety criteria for accident analyses 
are met. No new accident scenarios are created by these proposed 
changes. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety.
    Acceptance criteria for accident analysis continue to be met. 
Radiological consequences for the affected Chapter 15 analysis 
remain within 10 CFR 100 acceptance criteria. No safety limits or 
safety system setpoint requires modification due to the proposed 
changes. The current secondary side over-pressure limit of 100% of 
steam generator shell design pressure is not violated. Analysis for 
the high nuclear flux setpoints have verified that there is no 
reduction in margin for the events analyzed. Therefore, there is not 
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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: William H. Bateman

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

    Date of amendment request: September 9, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.8.1 and its associated Bases to 
improve emergency diesel generator reliability and availability. 
Several surveillance requirements would be revised or eliminated, and 
guidance provided in Regulatory Guide 1.9, Revision 3, and Generic 
Letter 93-05 would be incorporated.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

The proposed changes do not involve a significant hazards 
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.
    These proposed changes do not involve a change in the 
operational limits or physical design of the emergency power system. 
Emergency diesel generator operability and reliability will continue 
to be assured while minimizing the number of required emergency 
diesel generator starts. Also, emergency diesel generator 
reliability will be enhanced by minimizing service test conditions 
which can lead to premature failures.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    These proposed changes do not involve a change in the 
operational limits or physical design of the emergency power system. 
The performance capability of the emergency diesel generator will 
not be affected. Emergency diesel generator reliability and 
availability will be improved by the implementation of the proposed 
changes. There is no actual impact on accident analysis.
    3. Involve a Significant Reduction in the Margin of Safety.
    These proposed changes do not involve a change in the 
operational limits or physical design of the emergency power system. 
The performance capability of the emergency diesel generator will 
not be affected. Emergency diesel generator reliability and 
availability will be improved by the implementation of the proposed 
changes. No margin of safety is reduced.
    Based on the above discussions, it has been determined that the 
requested technical specification revision does 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. 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: Leif J. Norrholm

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

    Date of amendment request: September 9, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.8.2.1 and 3.8.2.2, 125-volt D.C. 
busses for battery bank and chargers and provides for the installation 
of swing chargers during the next refueling outage. Technical 
Specifications 3.8.3.1 and 3.8.3.2 would be revised to address the 120-
volt A.C. Vital Busses.
    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 the Technical Specifications do not 
involve a significant hazards consideration because operation of 
Callaway Plant in accordance with these changes would not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    These proposed Technical Specification changes do not involve 
any hardware changes nor do they affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters or accident mitigation capabilities. There will be no 
increase in the consequences of any accident or equipment 
malfunction. [[Page 507]] 
    2) Create the possibility for accident or malfunction of 
equipment of a different type than previously evaluated in the FSAR.
    The proposed Technical Specification changes do not involve any 
design changes nor are there any changes to the method by which any 
safety-related plant system performs its safety function. The normal 
manner of plant operation is unaffected. No new accident scenarios, 
transient precursors, failure mechanisms, or limiting single 
failures are introduced as a result of these changes.
    Involve a significant reduction in the margin of safety.
    There will be no affect [SIC] on the manner in which safety 
limits or limiting safety system settings are determined, nor will 
there be any effect in those plant systems necessary to assure the 
accomplishment of protection functions. There will be no impact on 
DNBR limits, FQ, F-delta-H, LOCA PCT, peak local power density 
or any other margin of safety.
    Based on the information presented above, the proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated, create the 
possibility of a new or different kind of accident from any 
previously evaluated, or involve a significant reduction in a margin 
of safety. Therefore, it is concluded that the proposed changes meet 
the requirements of 10 CFR 50.92(c) and does [SIC] 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: Leif J. Norrholm

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

    Date of amendment request: August 27, 1993
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to be consistent with recent 
revisions to 10 CFR Part 20 and 10 CFR 50.36a. Administrative changes 
are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The changes as proposed consist of revisions to the 
Technical Specifications to meet new regulatory requirements as 
contained in 10CFR20 and 10CFR50.36a, and other related changes of 
an administrative nature. There is no change in the types and 
amounts of effluents released, nor will there be any increase in 
individual or cumulative occupational radiation exposures. None of 
the changes proposed will affect any plant hardware, plant design, 
safety limit settings, or plant system operation, and therefore do 
not modify or add any initiating parameters that would significantly 
increase the probability or consequences of any previously analyzed 
accident.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The changes as proposed do not physically alter the plant 
nor do they change the operation of the plant.
    3. The proposed change does not involve a significant reduction 
in the margin of safety. The changes will not increase the amount or 
types of effluents that may be released offsite, nor do they 
significantly increase individual or cumulative occupational 
radiation exposures. These changes will not alter any of the 
requirements or responsibilities for protection of the public and/or 
employees against radiation hazards.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301
    Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray, 
One International Place, Boston, Massachusetts 02110-2624
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: March 31, 1994
    Description of amendment request: The proposed amendment would 
modify the requirements for avoidance and protection from thermal 
hydraulic instabilities to be consistent with the Boiling Water Reactor 
(BWR) Owner's Group long-term solution Option 1-D described in the 
Licensing Topical Report, ``BWR Owner's Group Long-Term Stability 
Solutions Licensing Methodology, NEDO-31960 June 1991'' and NEDO-31960, 
Supplement 1, dated March 1992. NEDO-31960 and NEDO-31960, Supplement 
1, were accepted by the NRC staff in a letter to L.A. England (BWR 
Owner's Group) dated July 12, 1993.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The implementation of BWR Owner's Group long 
term stability solution Option 1-D at Vermont Yankee does not modify 
the assumptions contained in the existing accident analysis. The use 
of an exclusion region and the operator actions required to avoid 
and minimize operation inside the region do not increase the 
possibility of an accident. Conditions of operation outside of the 
exclusion region are within the analytical envelope of the existing 
safety analysis. The operator action requirement to exit the 
exclusion region upon entry minimizes the possibility of an 
oscillation occurring. The actions to drive control rods and/or to 
increase recirculation flow to exit the region are maneuvers within 
the envelope of normal plant evolutions. The flow biased scram has 
been analyzed and will provide automatic fuel protection in the 
event of an instability. Thus, each proposed operating requirement 
provides defense in depth for protection from an instability event 
while maintaining the existing assumptions of the accident analysis.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from an accident previously 
evaluated. As stated in 1), the proposed operating requirements 
either mandate operation within the envelope of existing plant 
operating conditions of force specific operating maneuvers within 
those carried out in normal operation. Since operation of the plant 
with all of the proposed requirements are within the existing 
operating basis, an unanalyzed accident will not be created through 
implementation of the proposed change.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety. Each of the proposed requirements 
for plant thermal hydraulic stability provides a means for fuel 
protection. The combination of avoiding possible unstable conditions 
and the automatic flow biased reactor scram provides an in depth 
means for fuel protection. Therefore, the individual or combination 
of means to avoid and suppress an instability supplements the margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. 
[[Page 508]] 
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301
    Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray, 
One International Place, Boston, Massachusetts 02110-2624
    NRC Project Director: Walter R. Butler

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: November 29, 1994
    Description of amendment request: Virginia Electric and Power 
Company plans to insert fuel assemblies containing fuel rods, guide 
thimble tubes, instrumentation tubes, and mid-span grids fabricated 
with Westinghouse Electric Corporation's (Westinghouse's) advanced 
zirconium alloy material, ZIRLO, into the Surry Units 1 and 2 reactors, 
beginning with Cycle 14 at each unit. In the current fuel design, these 
components are fabricated from Zircaloy-4.
    Because the Technical Specifications define the fuel rod cladding 
material as Zircaloy-4, implementation of this material change requires 
changes to the Technical Specifications. Technical Specification 
5.3.A.1 is being modified to allow the use of either Zircaloy-4 or 
ZIRLO fuel rod cladding, and an additional reference for the 
calculation of the heat flux hot channel factor for loss-of-coolant-
accident evaluations of fuel with ZIRLO cladding is being defined in 
Technical Specification 6.2. The use of the ZIRLO fabricated guide 
thimble tubes, instrumentation tubes, and mid-span grids does not 
require changes to the Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Specifically, operation of Surry Power Station in accordance 
with the Technical Specifications changes will not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated. The Surry fuel 
assemblies containing fuel rods, guide thimble tubes, 
instrumentation tubes and mid-span grids fabricated with ZIRLO alloy 
meet the same fuel assembly and fuel rod design bases as the current 
fuel assemblies fabricated with Zircaloy-4 components. In addition, 
the 10 CFR 50.46 criteria will be applied to the fuel rods, guide 
thimble tubes, instrumentation tubes and mid-span grids fabricated 
with ZIRLO alloy. The use of these fuel assemblies will not result 
in a change to the Surry Units 1 and 2 reload design and safety 
analysis limits. The ZIRLO alloy is similar in chemical composition 
to Zircaloy-4, and also has physical and mechanical properties 
similar to those of Zircaloy-4. Thus the cladding integrity is 
maintained and the structural integrity of the fuel assembly is not 
affected. The ZIRLO clad fuel rods improve corrosion resistance and 
dimensional stability. Since the dose predictions in the safety 
analyses are not sensitive to the fuel rod cladding material changes 
as specified in this report, the radiological consequences of 
accidents previously evaluated in the safety analyses remain valid. 
Therefore, neither the probability of occurrence nor the 
consequences of any accident previously evaluated is significantly 
increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously identified, since the Surry Units 1 and 
2 fuel assemblies containing fuel rods, guide thimble tubes, 
instrumentation tubes and mid-span grids fabricated with ZIRLO alloy 
will satisfy the same design bases used for previous fuel regions 
containing Zircaloy-4 components. Since the original design criteria 
are being met, the fuel rods, guide thimble tubes, instrumentation 
tubes and mid-span grids fabricated with ZIRLO alloy will not be 
initiators for any new accident. Applicable design and performance 
criteria will continue to be met and no single failure mechanisms 
have been created. In addition, the use of these fuel assemblies 
does not involve any alteration to plant equipment or procedures 
which would introduce any new or unique operational modes or 
accident precursors. Therefore, the possibility for a new or 
different kind of accident from any accident previously evaluated is 
not created.
    3. Involve a significant reduction in a margin of safety. The 
Surry Units 1 and 2 fuel assemblies containing fuel rods, guide 
thimble tubes, instrumentation tubes and mid-span grids fabricated 
with ZIRLO alloy do not change the Surry Units 1 and 2 reload design 
and safety analysis limits. The use of fuel assemblies containing 
fuel rods, guide thimble tubes, instrumentation tubes and mid-span 
grids fabricated with ZIRLO alloy will take into consideration the 
normal core operating conditions allowed in the Technical 
Specifications. For each cycle reload core these fuel assemblies 
will be specifically evaluated using approved reload design methods 
and approved fuel rod design models and methods. This will include 
consideration of the core physics analysis peaking factors and core 
average linear heat rate effects. Analyses or evaluations will be 
performed each cycle to confirm that the 10 CFR 50.46 criteria will 
be met for the use of fuel with fuel rods, guide thimble tubes, 
instrumentation tubes and mid-span grids fabricated with ZIRLO 
alloy. Therefore, the margin of safety as defined in the Bases to 
the Surry Units 1 and 2 Technical Specifications is not 
significantly reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Mohan C. Thadani, Acting

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

    Date of amendment request: December 2, 1994
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
3.2 by eliminating the requirements for the charging pumps, high 
concentration boric acid in the boric acid storage tanks (BASTs), the 
boric acid transfer pumps, and boric acid heat tracing. Changes to TS 
3.3 and Table TS 3.5.3 are also being proposed to add requirements 
associated with the emergency core cooling system (ECCS) accumulators, 
remove the requirements associated with the boric acid storage tanks, 
and to increase the minimum required boron concentration in the 
refueling water storage tank (RWST). Additionally, the surveillance 
requirements involving the BASTs, associated valves and heat tracing 
located in Table TS 4.1-1, Table TS 4.1-2 and Section 4.5 would be 
eliminated. Supporting analysis for the limiting design basis accident 
conditions have been performed using the proposed minimum RWST boron 
concentration of 2400 ppm.
    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:
    Significant Hazards Determination for Proposed Changes to 
Technical Specification (TS) 3.2 and Table TS 3.5-3.
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated;
    Neither the charging pumps, the high concentration boric acid, 
the BASTs, the boric acid transfer pumps nor the boric acid heat 
tracing system are accident initiators. Therefore, a change to these 
systems will not significantly increase the probability of an 
accident previously evaluated. The effect of a reduction in initial 
safety injection boron concentration on the accident analysis was 
evaluated. The limiting accidents were the Large-Break Loss-of-
Coolant Accident [[Page 509]] (LOCA) and the Steam Line Break (SLB) 
event. A decrease in the initial safety injection boron 
concentration from 20,000 ppm to 2400 ppm will not adversely affect 
the Large or Small-Break Loss-of-Coolant Accident analysis because 
the evaluation models used in analyzing these accidents do not take 
credit for the high concentration boric acid stored in the BASTs. 
However, the evaluation models did take credit for boron in 
maintaining the long term post LOCA reactor core sub-critical. An 
analysis was performed which concluded that the inventory contained 
in the BASTs would not be required provided the minimum RWST boron 
concentration was increased to 2400 ppm. The SLB event is the other 
design basis event that could be affected by the proposed 
elimination of the high boron concentration BASTs as a source of 
safety injection fluid. Analyses have been performed which conclude 
that the BASTs are not required and that a minimum RWST boron 
concentration of only 1950 ppm is sufficient to provide adequate 
protection for the SLB event although 2400 ppm will be maintained to 
address post-LOCA subcriticality thus providing further safety 
margin. The results of these analyses indicate that the departure 
from nucleate boiling (DNB) design basis continues to be met. (A 
minimum Departure from Nucleate Boiling Ratio (DNBR) of 1.45 can be 
maintained throughout the event.) Finally, the containment pressure 
and temperature remains within the acceptable containment design 
limits. Since these criteria have been satisfied, there will be no 
adverse effect on the health and safety of the public and the 
consequences of any accident previously evaluated have not 
significantly increased.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated;
    Neither the charging pumps, the removal of the BASTs from 
initial SI pump injection, nor the elimination of both the boric 
acid transfer pumps and the boric acid heat tracing system as 
safety-related components would create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Furthermore, the reactivity control function of the boron in the 
CVCS and SI systems is not being changed. Therefore, the proposed 
changes will not adversely affect the health and safety of the 
public or create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3) Involve a significant reduction in the margin of safety.
    The reduction in the initial concentration of boron injected 
into the reactor coolant system for accident mitigation has been 
analyzed. These analyses conclude that all applicable criteria for a 
LOCA are satisfied. A decrease in the initial safety injection boron 
concentration from 20,000 ppm to 2400 ppm will not adversely effect 
the Large-or Small-Break Loss-of-Coolant Accident analysis because 
the evaluation models used in analyzing these accidents do not take 
credit for the high concentration boric acid stored in the BASTs. 
However, in order to maintain the long term post LOCA reactor core 
sub-critical, a minimum RWST boron concentration of 2400 ppm is 
required. To meet this requirement, the RWST boron concentration is 
being raised to 2400 ppm. All criteria of 10 CFR 50.46 can be 
achieved for both the Large or Small-Break LOCA with no BASTs and 
2400 ppm boron in the RWST. Since all criteria of 10 CFR 50.46 are 
satisfied, there is no adverse effect on the health and safety of 
the public and there is not a significant reduction in the margin of 
safety for these casualties.
    Since both the core response and the containment response can be 
limiting in the SLB event, both were considered in the boron 
concentration reduction analysis. This analysis concludes that a 
minimum RWST boron concentration of 1950 ppm is sufficient to 
provide adequate protection for the SLB event, although a 2400 ppm 
boron solution will be maintained to provide protection for the post 
LOCA concerns. Since the containment pressure and temperature 
remains within the acceptable containment design limits, and a 
minimum DNBR of 1.45 can be maintained throughout the event, there 
is not a significant reduction in the margin of safety for this 
event and therefore there is no adverse effect on the health and 
safety of the public.
    These proposed changes involve the conversion of the TS to Word 
Perfect format now being used at WPSC. Minor typographical errors 
and format inconsistencies were corrected. These proposed changes 
are administrative in nature; accordingly, these proposed changes do 
not involve a significant hazards consideration.
    Additionally, the proposed changes are similar to example 
C.2.e.(i) in 51 FR 7751. Example C.2.e.(i) states that changes which 
are purely administrative in nature; i.e., to achieve consistency 
throughout the Technical Specifications, correct an error, or a 
change in nomenclature, are not likely to involve a significant 
hazard.
    Significant Hazards Determination for Proposed Changes to Table 
TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations and Test of 
Instrument Channels'' and Table TS 4.1-2 ``Minimum Frequencies for 
Sampling Tests''
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated, 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 the margin of safety.
    The above listed surveillance requirements insure BAST 
operability. The BASTs will no longer be relied upon as a source of 
boron for safety injection, and will serve no safety related 
function. Whether the BASTs are operable or not will have no effect 
on plant safety. Therefore, elimination of the surveillance 
requirements which insure BAST operability is possible without any 
adverse effect on the health and safety of the public and presents 
no significant hazards.
    Significant Hazards Determination for Proposed Changes to 
Technical Specification TS 3.3 and Section 4.5.
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Neither the RWST, the boron solution contained within the RWST 
nor valves SI-3, SI-4A/B are accident initiators. Therefore, a 
change to these systems will not significantly increase the 
probability of an accident previously evaluated. The effect of a 
reduction in initial Safety Injection boron concentration on the 
accident analysis was evaluated. The limiting accidents were the 
Large-Break Loss-of-Coolant Accident (LOCA) and the Steam Line Break 
(SLB) event. A decrease in the initial safety injection boron 
concentration from 20,000 ppm to 2400 ppm will not adversely effect 
the Large or Small-Break Loss-of-Coolant Accident analysis because 
the evaluation models used in analyzing these accidents do not take 
credit for the high concentration boric acid stored in the BASTs. 
However, the evaluation models did take credit for boron in 
maintaining the long term post LOCA reactor core sub-critical. An 
analysis was performed which concluded that the BASTs could be 
eliminated provided the minimum RWST boron concentration was 
increased to 2400 ppm. The SLB event is the other design basis event 
that could be affected by the proposed elimination of the high 
concentration BASTs as a safety-related source for reactivity 
control injection fluid. However, analyses have been performed which 
conclude that a minimum RWST boron concentration of only 1950 ppm is 
sufficient to provide adequate protection for the SLB event although 
2400 ppm will be maintained to address post-LOCA subcriticality thus 
providing further safety margin. The results of these analyses 
indicate that the departure from nucleate boiling (DNB) design basis 
continues to be met. (A minimum Departure from Nucleate Boiling 
Ratio (DNBR) of 1.45 can be maintained throughout the event.) 
Furthermore, maintaining the suction of the SI pumps to the RWST 
with valves SI-4A or SI-4B open with power removed places the system 
in a normal SI sequence and eliminates the requirement to switch 
suction from the BASTs to the RWST. This eliminates a potential 
failure mechanism and increases the overall reliability of the ECCS 
system. Finally, the containment pressure and temperature remains 
within the acceptable containment design limits.
    Since these criteria have been satisfied, there will be no 
adverse effect on the health and safety of the public and the 
consequences of any accident previously evaluated have not 
significantly increased.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This change to the Technical Specifications allows use of 2400 
ppm boron for safety injection. SI pump suction would be directly 
from the RWST. This eliminates [[Page 510]] the necessity of 
shifting suction from the BASTs to the RWST, reducing the complexity 
of the operation. Since the pumps remain connected to the RWST 
throughout the injection phase, there is no possibility of a new or 
different kind of accident from any accident previously evaluated.
    Neither the reduction in initial boron concentration for safety 
injection, nor the increase in the boron concentration in the RWST 
would create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Lastly, the reactivity control function of the boron in the CVCS 
and SI systems is not being changed. Therefore, the proposed changes 
will not adversely affect the health and safety of the public or 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3) Involve a significant reduction in the margin of safety.
    The change in concentration of boron injected into the primary 
system for accident mitigation has been analyzed. These analyses 
conclude that all applicable criteria for a LOCA are satisfied. A 
change in safety injection boron concentration to 2400 ppm will not 
adversely affect the Large or Small-Break LOCA analysis because the 
evaluation model codes used in analyzing these accidents did not 
take credit for boron. However, a minimum RWST boron concentration 
of 2400 ppm is required to maintain long term post LOCA reactor core 
sub-criticality. To meet this requirement, the RWST minimum boron 
concentration is being raised to 2400 ppm. All criteria of 10 CFR 
50.46 can be achieved for both the Large or Small-Break LOCA with 
2400 ppm boron in the RWST. Since all criteria of 10 CFR 50.46 are 
satisfied, there is no adverse effect on the health and safety of 
the public and there is not a significant reduction in the margin of 
safety for these casualties.
    Since both the core response and the containment response can be 
limiting in the SLB event, both were considered in the boron 
concentration reduction analysis. Although a minimum RWST boron 
concentration of 1950 ppm is sufficient to provide adequate 
protection for the SLB event, a 2400 ppm boron solution will be 
maintained to provide protection for the post large break LOCA 
concerns. Since the containment pressure remains below the design 
pressure, and a minimum DNBR of 1.45 can be maintained throughout 
the event, there is not a significant reduction in the margin of 
safety for this event.
    These proposed changes involve the conversion of the TS to Word 
Perfect format now being used at WPSC. Minor typographical errors 
and format inconsistencies were corrected. These proposed changes 
are administrative in nature; accordingly, these proposed changes do 
not involve a significant hazards consideration.
    Additionally, the proposed changes are similar to example 
C.2.e.(i) in 51 FR 7751. Example C.2.e.(i) states that changes which 
are purely administrative in nature; i.e., to achieve consistency 
throughout the Technical Specifications, correct an error, or a 
change in nomenclature, are not likely to involve a significant 
hazard.
    Significant Hazards Determination for Proposed Changes to 
Technical Specification (TS) Section 4.5 ``Emergency Core Cooling 
System and Containment Air Cooling System Tests.''
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated, 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 the margin of safety.
    The above listed surveillance requirements insure BAST 
operability. The BASTs will no longer be relied upon as a source of 
boron for safety injection, and will serve no safety related 
function. Whether the BASTs are operable or not will have no effect 
on plant safety. Therefore, elimination of the surveillance 
requirements which insure BAST operability is possible without any 
adverse effect on the health and safety of the public and presents 
no significant hazards.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: Leif J. Norrholm

Previously Published Notices Of Consideration Of Issuance Of 
Amendments ToFacility 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.

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

    Date of amendment request: November 29, 1994
    Brief description of amendment request: The proposed amendment 
would delete requirements to perform the full complement of steam 
generator surveillances as outlined in the technical specifications 
(TSs) when the steam generators are subjected to special inspections 
that are in addition to inspections that are required by the TSs.
    Date of individual notice in the Federal Register: December 5, 1994 
(59 FR 62416)
    Expiration date of individual notice: January 4, 1995
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801

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 
[[Page 511]] Assessment as indicated. All of these items are available 
for public inspection at the Commission's Public Document Room, the 
Gelman Building, 2120 L Street, NW., Washington, DC, and at the local 
public document rooms for the particular facilities involved.

Arizona Public Service Company, et al., Docket No. STN 50-528, Palo 
Verde Nuclear Generating Station, Unit No. 1, Maricopa County, 
Arizona

    Date of application for amendment:  November 22, 1994
    Brief description of amendment: The amendment adds a note to 
Technical Specification Table 3.7-2. The note allows continuous 
operation of Unit 1 during Cycle 5 at 100-percent maximum steady state 
power with one main steam safety valve inoperable per steam generator. 
This note applies only during the current fuel cycle (Cycle 5) for Unit 
1.
    Date of issuance: December 19, 1994
    Effective date: December 19, 1994
    Amendment No.: 87
    Facility Operating License No. NPF-41: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (59 FR 61907, dated December 2, 1994). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by December 19, 1994, but stated that, if the Commission makes 
a final no significant hazards consideration determination, any such 
hearing would take place after issuance of the amendment. The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, and final determination of significant hazards 
consideration is contained in a Safety Evaluation dated December 19, 
1994.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: October 7, 1994
    Brief description of amendment: The amendment revises the 
introduction to TS Section 6.9.3.3 to require the approved revision 
number for the referenced analytical methods to be listed in the Core 
Operating Limits Report. The methodology referenced in 6.9.3.3.b.f (XN-
NF-82-49(A)) has been updated to clarify that all supplements are 
included. New methodologies ANF-89-151(A) and EMF-92-081(A) will be 
added to TS Section 6.9.3.3.b.
    Date of issuance: December 12, 1994
    Effective date: December 12, 1994
    Amendment No.: 154
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55868)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 12, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College, Hartsville, South Carolina 29550

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

    Date of application for amendment: February 18, 1994, as 
supplemented by letters dated June 3, 1994, November 1, 1994, December 
2, 1994, December 14, 1994 and December 16, 1994.
    Brief description of amendment: The amendment revises surveillance 
intervals for the Vapor Containment Sump Discharge Flow and Temperature 
Channel, the Loss of Power Undervoltage and Degraded Voltage Relays, 
and the Control Rod Protection System Trip to accommodate a 24-month 
refueling cycle. In addition it changes the trip setpoint for the 
Control Rod Protection System Trip. These revisions are being made in 
accordance with the guidance provided by Generic Letter 91-04, 
``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle.''
    Date of issuance: December 20, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 179
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22003) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 20, 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.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: July 19, 1994
    Brief description of amendments: These amendments change Technical 
Specification 3.1.5 for each unit for the standby liquid control system 
(SLCS) to remove the operability requirement for the SLCS while the 
plant is in Operational Condition 5 (refueling) with any control rod 
withdrawn, and to delete the 18-month system surveillance requirement 
(Surveillance Requirement 4.1.5.d.3).
    Date of issuance: December 20, 1994
    Effective date: December 20, 1994
    Amendment Nos.: 136 and 106
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42344)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 20, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

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

    Date of application for amendments: July 22, 1994
    Brief description of amendments: The amendment removes the 
surveillance frequency details regarding 10 CFR Part 50, Appendix J, 
Types B and C testing from the Technical Specifications.
    Date of issuance: December 19, 1994
    ]Effective date: December 19, 1994
    Amendment Nos. 83 and 44
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47180) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 19, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
[[Page 512]] High Street, Pottstown, Pennsylvania 19464.

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: October 7, 1994
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 4.6E.4 and the associated Bases to establish that 
the manual cycling of reactor coolant system safety/relief valves 
(SRVs) during plant startups is to be accomplished within 12 hours 
after steam pressure and flow are adequate to perform the testing. TS 
4.6E.4 had previously required that this testing be performed within 12 
hours of continuous power operation at a reactor steam dome pressure of 
at least 940 psig. The amendment also makes several editorial changes 
to clarify the intent of TSs involving SRV testing and performance 
requirements.
    Date of issuance: December 16, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 219
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55889)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 16, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey Date of application for amendments: September 9, 
1994

    Brief description of amendments: The amendments revise the 
Technical Specification surveillance requirements regarding visual 
inspection of snubbers and are consistent with the guidance provided in 
Generic Letter 90-09, ``Alternative Requirements for Snubber Visual 
Inspection Intervals and Corrective Actions.''
    Date of issuance: December 12, 1994
    Effective date: December 12, 1994
    Amendment Nos. 161 and 142
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55889)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 12, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey Date of application for amendments: March 28, 
1994, as supplemented June 1, 1994, and August 24, 1994

    Brief description of amendments: The amendments revise the 
sustained degraded voltage relay trip setpoint and the allowable value 
due to changes in the switchyard configuration.
    Date of issuance: December 14, 1994
    Effective date: December 14, 1994
    Amendment Nos. 162 and 143
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29633) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 14, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendments: October 7, 1993 (TS 313)
    Brief description of amendments: The changes include the addition 
of the high range primary containment radiation monitors and recorders 
and the wide range gaseous effluent radiation recorder and monitor, 
which were installed at the Browns Ferry facility in response to NUREG 
0737 ``Clarification of TMI Action Plan Requirements'' and GL 83-36, 
into the technical Specifications (TS) for Units 1 and 3. Similar 
changes to the Unit 2 TS were issued previously (Amendment Nos. 125 and 
171). The amendment also clarifies that the high range primary 
containment radiation recorders and monitors are both part of the 
instrument loop. The amendment contains administrative typographical 
changes which provide consistency for the TS tables and footnotes for 
Units 1 and 3.
    Date of issuance: December 21, 1994
    Effective Date: December 21, 1994
    Amendment Nos.: 214, 230, 187
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67863)The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 21, 1994.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611

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

    Date of amendment request: November 11, 1994, as supplemented by 
letter dated November 16, 1994.
    Brief description of amendments: The proposed amendment would 
modify Comanche Peak Steam Electric Station Technical Specification 
Table 4.8-1, ``Diesel Generator Test Schedule,'' by excluding two valid 
failures of the Unit 2 Train B diesel generator from contributing 
towards an accelerated test schedule.
    Date of issuance: December 9, 1994
    Effective date: December 9, 1994
    Amendment Nos.:  Unit 1 - Amendment No. 33; Unit 2 - Amendment No. 
19
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.Public comments requested as to 
proposed no significant hazards consideration: Yes (59 FR 69399, dated 
November 23, 1994). The notice provided an opportunity to submit 
comments on the Commission's proposed no significant hazards 
consideration determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by December 23, 
1994, but stated that, if the Commission makes a final no significant 
hazards consideration determination, any such hearing would take place 
after issuance of the amendments.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration is contained in a Safety Evaluation dated 
December 9, 1994.
    Local Public Document Room location: University of Texas at 
Arlington library, Government [[Page 513]] Publications/Maps, 702 
College, P.O. Box 19497, Arlington, Texas 76019.

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

    Date of application for amendments: March 29, 1994
    Brief description of amendments: These amendments revise Point 
Beach Nuclear Plant Technical Specification 15.3.2, ``Chemical and 
Volume Control System,'' by eliminating the necessity for high 
concentration boric acid and removing the operability requirements for 
the associated heat tracing. The basis for Section 15.3.2 and 
applicable surveillances in Table 15.4.1-2 are also revised to support 
the above changes.
    Date of issuance: December 12, 1994
    Effective date: Date of issuance, to be implemented within 45 days.
    Amendment Nos.: 158 & 162
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37091) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 12, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

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

    Date of application for amendments: September 12, 1994
    Brief description of amendments: These amendments revise Point 
Beach Nuclear Plant Technical Specification (TS) 15.3.3, ``Emergency 
Core Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan 
Coolers, and Containment Spray,'' by incorporating allowed outage times 
similar to those contained in NUREG-1431, Revision 0, ``Westinghouse 
Owner's Group Improved Standard Technical Specifications,'' and by 
clarifying the operability requirements for the service water pumps. 
The changes also clarify the completion times for placing a unit in hot 
or cold shutdown, if a limiting condition for operation cannot be met.
    Date of issuance: December 21, 1994
    Effective date: Date of issuance, to be implemented within 45 days
    Amendment Nos.: 159 & 163
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 24, 1994 (59 FR 
53844)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 21, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Dated at Rockville, Maryland, this 27th day of December 1994.
    For The Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV, Office of 
Nuclear Reactor Regulation.
[Doc. 95-5 Filed 1-3-95; 8:45 am]
BILLING CODE 7590-01-F