[Federal Register Volume 62, Number 195 (Wednesday, October 8, 1997)]
[Notices]
[Pages 52578-52599]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-11008]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the

[[Page 52579]]

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 September 15, 1997, through September 26, 
1997. The last biweekly notice was published on September 24, 1997 (62 
FR 50000).

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 Chief, Rules and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, 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 November 7, 1997, 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 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.

[[Page 52580]]

    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 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, 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, 
Units Nos. 1, 2, and 3, Maricopa County, Arizona

    Date of amendments request: March 18, 1997, as supplemented by 
letters dated July 28, 1997 and September 9, 1997
    Description of amendments request: The amendments would revise the 
operating licenses for Palo Verde Units 1, 2 and 3 to reflect approval 
of Amendment 42 to the Palo Verde Nuclear Generating Station (PVNGS) 
Physical Security Plan. Amendment 42 would revise the methods used to 
search materials, packages and personnel prior to their entry into the 
protected area, as described within the security plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated. The 
``accident'' as it relates to the Security Plan would have to be an 
impact to the Design Basis Threat (DBT) postulated for PVNGS. This 
change does not decrease the overall security systems (as described 
in paragraph's (b) through (h) of 10 CFR 73.55) ability to protect 
PVNGS with the objective of high assurance against the DBT of 
radiological sabotage as stated in 73.1(a). This change does not 
delete or contradict any regulatory requirements.
    The applicable design basis threat is described in 10 CFR 73.1. 
Based on that threat, the probability of an external determined 
violent assault by stealth, or deceptive actions, of several persons 
is unaffected by the requested changes to the search requirements. 
Similarly, an internal threat of an insider, including an employee 
(in any position) is no more likely to occur as a result of the 
search techniques. The probability of an attack with a four-wheel 
drive land vehicle bomb is unaffected. Theft or diversion of formula 
quantities of strategic special nuclear material is a threat of 
removal from the inside of the protected area, which is not within 
the scope of this change that only affects searches of material 
entering the protected area.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The possibility of an accident of a new or different kind has 
not been created because the DBT (as described in the Security Plan 
and 10 CFR 73.1) would not be changed as a result of these changes. 
The changes supplement regulatory requirements and commitments 
already described in the PVNGS Physical Security Plan.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety. These changes to the personnel, material and 
package search criteria are not specifically considered in the basis 
for any margin of safety. The DBT considers inside assistance by a 
knowledgeable individual, however, these changes would not assist 
this individual in either sabotage or theft of nuclear material.
    The NRC staff has reviewed the licensee's 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 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    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
    NRC Project Director: William H. Bateman

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

    Date of amendment request: September 15, 1997
    Description of amendment request: The proposed license amendments 
would revise the Technical Specifications (TS) to:
    1. Revise the reactor coolant system heatup limitation curves in 
Figure 3.4-2, which are applicable only to the first 10 effective full-
power years (EFPYs). The revised curves would be (a) applicable to the 
first 15 EFPYs; (b) include the latest radiation surveillance capsule 
results; (c) remove instrument margins by relocating them to a 
licensee-controlled document, ``Pressure Temperature Limit Report;'' 
and (d) administratively delete certain unneeded footnotes that exist 
in the current figure.
    2. Modify the actual surveillance capsule identification listed in 
Table 4.4-5, ``Reactor Vessel Material Surveillance Program - 
Withdrawal Schedule'' (for Unit 2 only) and update each units lead 
factors and withdrawal time.
    3. Revise the power-operated relief valve (PORV) setpoints in 
Section 3.4.9.3.a to less than or equal to 400 pounds per square inch 
gauge (psig) (as left calibrated), allowable value less than or equal 
to 425 psig (as found).
    4. Make editorial changes to improve consistency among various TS 
sections to conform with the Westinghouse Improved Standard Technical 
Specifications, and update applicable Code references.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below.
    1. Will the changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?No. 
No previously evaluated accident was considered to originate from 
use of the heatup curves (change 1. above), the testing and use of 
surveillance capsules (change 2. above), the setpoint of PORVs 
(change 3. above), and editorial changes to the TS. Also, these 
items did not have any role in previously analyzed accident 
scenarios and thus no impact on accident consequences. Therefore, 
these proposed changes will have

[[Page 52581]]

no impact on the consequences or probabilities of any type of 
previously evaluated accidents.
    2. Will the changes create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. No actual plant equipment or operating procedure will be 
affected by the proposed changes. Hence, no new equipment failure modes 
or accidents from those previously evaluated will be created.
    3. Will the changes involve a significant reduction in a margin 
of safety?
    No. The margin of safety is associated with confidence in the 
design and operation of the plant. The changes to the TS do not 
involve any change to plant design or operation. Thus, the margin of 
safety previously analyzed and evaluated is maintained.
    On the basis of this analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location:  York County Library, 138 East 
Black Street, Rock Hill, South Carolina
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North 
Carolina
    NRC Project Director: Herbert N. Berkow

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: June 14, 1997
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) for the Crystal River Nuclear 
Electric Generating Plant Unit 3 (CR-3). The proposed TS changes 
reflect the operational limitations in mitigating certain Small break 
loss-of-coolant-accident (SBLOCA) events. The licensee also proposed 
changes to the associated licensing and design bases.
    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 are 
addressed in three major parts: (1) SBLOCA Mitigation, (2) Emergency 
Diesel generator (EDG) upgrade and (3) EDG Load Rejection Test and 
Steady State Loads.
    SBLOCA Mitigation
    The licensee's revised SBLOCA analyses show that for certain 
sized breaks, a combination of emergency core cooling system (ECCS) 
flow to the reactor vessel and emergency feedwater (EFW) flow to the 
once through steam generators (OTSG) is needed to provide for 
adequate core decay heat removal. Due to load capacity limits on the 
A EDG, the length of time that the motor-driven emergency feed 
pump-1 (EFP-1) would be available is limited. To ensure adequate EFW 
system flow and core decay heat removal, several actions would have 
to be initiated. They include A EDG load management, and EFW flow 
through the turbine-driven emergency feedwater pump-2 (EFP-2) by 
opening the cross tie valve, flow through both the high pressure 
injection (HPI) pumps and EFP-1. The proposed TS changes reflect the 
operational limitations and other associated required actions to 
ensure adequate ECCS and EFW cooling capability remains. These 
changes for system cross train dependencies and EDG load management 
are required for the remainder of current Cycle 11 only.
    1. The proposed Technical Specification changes, modifications, 
and operator actions involving SBLOCA mitigation will not result in 
a significant increase in the probability of an accident previously 
evaluated. In addition, the portions of the change involving cross-
train dependencies and load management are being requested for the 
remainder of Cycle 11 only, which limits the impact on any 
previously established probabilities. The initiators of any design 
basis accident is not affected by the proposed Technical 
Specification changes, modifications, and operator actions involving 
SBLOCA mitigation. Consequently, there is no significant impact on 
any previously evaluated accident probabilities.
    The proposed Technical Specification changes, modifications and 
operator actions involving SBLOCA mitigation do not result in a 
significant increase in the consequences of SBLOCA mitigation-
related accidents previously evaluated. In this regard, the proposed 
Technical Specification changes, modifications and operator actions 
will not adversely affect the integrated ability of the EDGs and the 
EFW, SW [service water], RW [raw water], Control Complex Cooling, 
ECCS, DC [Decay Heat Closed Cycle Cooling Water System], Decay Heat 
Seawater, and Electrical Distribution Systems to perform their 
intended safety functions. Therefore, the combined ability of these 
components and systems and actions to mitigate the consequences of a 
SBLOCA will continue to be maintained. In fact, the collective 
impact of these Technical Specification changes, modifications and 
operator actions represents a restoration of the ability to mitigate 
the consequences of a SBLOCA, which are consistent with the 
consequences assumed in licensing and design basis for CR-3. For 
example, the installation of EFW cavitating venturis and the 
improved operational range of the turbine driven feedwater pump 
increase the ability of the EFW system to mitigate the consequences 
of a SBLOCA. In addition, the Technical Specification changes, 
modifications and operator actions do not significantly affect the 
onsite or offsite doses which remain a small fraction of 10 CFR Part 
100 limits.
    2. The proposed Technical Specification changes, modifications 
and operator actions do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The Technical Specification changes, modifications, and operator 
actions do not involve a different initiator for any design basis 
accident and do not create new design basis scenarios. SBLOCA 
mitigation, utilizing a combination of automatic and manual actions, 
is already part of the CR-3 licensing basis. Manual operator actions 
necessary for the mitigation of SBLOCAs are currently addressed or 
are being addressed in EOPs [emergency operating procedures]. Also, 
these Technical Specification changes, modifications and operator 
actions restore the ability to mitigate the impact of a SBLOCA, 
which is consistent with the CR-3 licensing and design basis. Based 
on the above, a new or different kind of accident does not result 
from this submittal.
    3. The proposed Technical Specification changes, modifications 
and operator actions do not involve a significant reduction in the 
margin of safety for SBLOCA mitigation. The Technical Specification 
changes, modifications and operator actions for the EDGs and the 
EFW, SW, RW, Control Complex Cooling Systems represent a restoration 
of the overall margin of safety to a degree that it will be 
consistent with the existing plant design and licensing bases for 
SBLOCA mitigation.
    EDG upgrade
    This aspect of the proposed license amendment involves increases 
in the service ratings of the EDGs. The required amount of fuel oil 
in the EDG fuel day tank and fuel storage tank, and lube oil storage 
is being increased to ensure that adequate volume is available to 
support the new service ratings. The EDG refueling interval load 
test parameters are being revised to reflect the increased service 
ratings and to ensure that the minimum test load is equal to or 
greater than the expected maximum steady state accident load. 
Additionally, associated EDG Surveillance Requirements (SR) Bases 
are being revised.
    1. The proposed Technical Specification changes, modifications 
and operator actions do not involve a significant increase in the 
probability of an accident previously evaluated because neither the 
EDGs nor the EDGs fuel oil and lube oil systems serve as the 
initiator for any design basis accident and, therefore, do not 
significantly impact any previously evaluated accident 
probabilities.
    The proposed Technical Specification changes, modifications and 
operator actions do not involve a significant increase in the 
consequences of an accident previously evaluated because the ability 
of the EDGs and the EDG fuel oil and lube oil to perform their 
intended safety function has not been adversely affected. The EDGs 
and the EDG fuel oil and lube oil systems remain fully capable of 
performing their safety function for all design basis accidents. The 
increase in loading permitted under these changes will reflect the 
manufacturers certified capabilities of the EDGs. Also, the 
increase in the required fuel remains within the capabilities of the 
fuel tanks. The same potential design basis failures that existed 
prior to the EDG upgrades will continue to

[[Page 52582]]

exist subsequent to the modifications. It follows that the 
consequences of such failures will remain a small fraction of 10 CFR 
Part 100 limits.
    2. The proposed Technical Specification changes, modifications 
and operator actions do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
Also, the proposed Technical Specification changes, modifications 
and operator actions do not involve any new accident initiators, or 
a new or different kind of accident from any previously evaluated. 
In addition, the configuration and basic function of the EDGs and 
EDG's fuel and lube oil systems are unaffected by the changes. In 
fact, the EDG upgrades ensure that the previously evaluated 
accidents are consistent with system and component capabilities and 
the current design and licensing bases.
    3. The proposed Technical Specification changes, modifications 
and operator actions do not involve a significant reduction in the 
margin of safety. The EDGs and EDGs fuel and lube oil systems will 
continue to able to be perform their safety function for all design 
basis accidents. There is an increase in the net margin of safety 
for fuel and lube oil storage since required volumes have been 
recalculated and increased, additional margin has been added to the 
calculated results, and the required volumes are based on usable 
tank volumes instead of tank capacity. These volumes continue to 
bound the postulated worse-case accident scenario. The increase in 
fuel storage required by the changes remains within the capacity of 
the storage tanks. The Technical Specification changes, 
modifications and operator actions further ensure that margins 
provided in current design and licensing bases are satisfied.
    EDG Load Rejection Test and Steady State Loads
    The proposed changes for this part affects the TS Bases. The 
basis of the EDG load rejection test is being revised to bound the 
largest single load. A description of ``steady state'' is being 
provided with examples of short duration loads and loads imposed by 
the starting of motors. Also, addressed is the licensee's conclusion 
that the refueling interval EDG load test is not invalidated by 
loads imposed by the starting of motors.
    1. The proposed Technical Specification changes, modifications 
and operator actions do not involve a significant increase in the 
probability of an accident previously evaluated because the EDG load 
tests and load rejection test do not serve as the initiator for any 
design basis accident and, therefore, do not significantly impact 
any previously evaluated probabilities.
    The proposed Technical Specification changes, modifications and 
operator actions do not involve a significant increase in the 
consequences of an accident previously evaluated because the changes 
do not affect the ability of the EDGs to perform their intended 
safety function. Rather, the Technical Specification changes, 
modifications and operator actions provide further assurance that 
the EDGs are capable of performing their safety function. Failure of 
an EDG has the same consequences as it would if the changes were not 
made. It follows that the 10 CFR Part 100 consequences of such 
failures has not changed.
    2. The proposed Technical Specification changes, modifications 
and operator actions do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the changes do not affect the ability of the EDGs to perform 
their intended safety function. The configuration and basic function 
of the EDGs, including accurately describing the manufacturer 
certified EDGs service ratings and steady state loads, do no create 
a possibility for a new or different kind of accident. Although the 
load rejection test is for an increased EDG largest single load, the 
kind of accident addressed by both the load rejection test and the 
refueling load test remain the same.
    3. The proposed Technical Specification changes, modifications 
and operator actions do not involve a significant reduction in the 
margin of safety. The calculated loads imposed by the starting of 
motors are short duration, have a low probability of occurrence, and 
are expected to be within the manufacturer limits. In fact, the 
margin confirmed by EDG refueling load testing and load rejection 
testing will demonstrate a restoration of design and licensing 
margin and confirm that the EDGs remain fully capable of performing 
their safety function for all design basis accidents.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: September 16, 1997
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) would modify TS 3.7.1.1, ``Plant Systems 
Turbine Cycle Safety Valves.'' During its effort to verify the current 
design and licensing bases for Millstone, Unit 2, NNECO has determined 
that the maximum allowable power level high trip setpoints with 
inoperable steam line code safety valves specified in Table 3.7-1 of TS 
3.7.1.1 are incorrect. The trip setpoints were not changed to be 
consistent with a previously approved reduction in the maximum power 
level high trip setpoint. In addition, NNECO is also in the process of 
reanalyzing the inadvertent closure of the main steam isolation valve 
(MSIV) and the loss of electrical load events. The results of the 
reanalysis indicate that the MSIV event results in the highest peak 
pressure in the secondary system and that the formula currently 
contained in the TS Bases for TS 3.7.1.1 may not result in the correct 
trip setpoints.
    Specifically, NNECO proposes to: (1) delete TS Table 3.7.1 by not 
allowing operation in Mode 1 or 2 with inoperable steam line code 
safety valves, (2) modify the associated action statement in TS 
3.7.1.1, and (3) update the TS Bases to reflect the proposed changes 
and update the amendment history numbers to reflect previously approved 
amendments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change does not involve an SHC [significant hazards 
consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This proposed change will remove the ability to operate in Modes 
1 or 2 with inoperable main steam line code safety valves. Operation 
in Mode 3 will be retained, provided no more than three main steam 
line code safety valves per steam generator are inoperable.
    The primary function of the main steam line code safety valves 
is to prevent secondary system overpressurization. These valves will 
also provide reactor core heat removal and design basis accident 
mitigation. This proposed change does not affect the length of time 
the plant can operate with inoperable main steam line code safety 
valves before compensatory actions must be taken. (Four hours is 
still allowed to restore the valve(s) to operable status.) This 
proposed change does not affect the probability of occurrence of any 
design basis accident and does not affect how the main steam line 
code safety valves function to mitigate design basis accidents. 
Therefore, this change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not alter the way any structure, 
system, or component functions. The proposed change will 
conservatively change plant operation in Modes 1 and 2 by removing 
the ability to operate at power with inoperable main steam

[[Page 52583]]

line code safety valves as currently specified in Technical 
Specification 3.7.1.1. It does not introduce any new failure modes 
and does not alter any assumption made in the safety analysis.
    Therefore, the change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    This proposed change to Technical Specification 3.7.1.1 will 
remove the ability to operate in Modes 1 or 2 with inoperable main 
steam line code safety valves. Operation in Mode 3 will be retained, 
provided no more than three main steam line code safety valves per 
steam generator are inoperable. The operability of the main steam 
line code safety valves ensures that the secondary system pressure 
will be limited to within 110% (1100 psig) of the design pressure of 
1000 psig during the most severe anticipated system operational 
transient. This change will not affect the operability requirements 
for the main steam line code safety valves and will not affect the 
length of time the plant can operate with inoperable main steam line 
code safety valves before compensatory actions must be taken. This 
will ensure the plant equipment required for design basis accident 
mitigation will be available. Therefore, there is no significant 
reduction in a margin of safety as defined in the Bases of Technical 
Specification 3.7.1.1.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut
    NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: August 29, 1997
    Description of amendment request: Based on a review and subsequent 
calculations of the cold overpressurization protection (COPS) enabling 
temperature and the emergency core cooling system (ECCS)/charging 
system Mode 3 requirements, NNECO proposes to reduce the COPS enabling 
temperature. As a result, NNECO proposed the following Technical 
Specifications (TS) changes: new heatup and cooldown pressure/
temperature limit curves and their associated requirements; new power 
operated relief valve (PORV) setpoint curves and their associated 
requirements; revisions to the reactor coolant loops and coolant 
circulation, ECCS, boration systems, and COPS to incorporate the lower 
enabling temperature and new restrictions for cold overpressure 
protection system (COPPS), PORV undershoot, and residual heat removal 
(RHR) relief valve bellows; addition of a footnote to allow a reactor 
coolant pump (RCP) to substitute for an RHR pump during heatup from 
Mode 5 to Mode 4, which is consistent with the improved standard 
technical specification (STS); reword TS 3/4.4.9.3 and its Bases 
section to be consistent with the improved STS; and revision of the 
affected Bases sections to be consistent with the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed revision does not involve [an] SHC because 
the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    Probability of Occurrence of Previously Evaluated Accidents
    Since the PORV setpoints and the COPS enabling temperature have 
been calculated in accordance with 10CFR50, Appendix G and ASME 
[American Society of Mechanical Engineers] Section XI, the change 
will not alter the probability that an overpressurization event will 
result in a loss of RV [reactor vessel] integrity. The new PORV 
setpoint curves are lower than the current curves in certain 
temperature ranges (below approximately 130F and above 
approximately 220 deg.F), and therefore the operating window is 
slightly decreased. However, the reduced operating window is still 
sufficient for normal anticipated pressure fluctuations. Below 
160 deg.F, operation of Reactor Coolant Pumps are prohibited if the 
PORVs are armed for COPPS; therefore, PORV actuation will not occur 
below 160F when the RCPs are running. In a water solid condition, 
RCS [reactor coolant system] pressure is maintained via the letdown 
low pressure control valve, which, when in automatic mode, maintains 
the RCS pressure in a relatively narrow range. When the RCPs are not 
running, the PORV COPPS system can be actuated. However, for this 
condition, the allowable pressure range is 0 to 418 psia [pounds per 
square inch atmospheric]. This pressure range is sufficient to 
accommodate normal anticipated pressure fluctuations.
    Above 220 deg.F, the minimum pressure range is from 300 psia to 
595 psia; this range is sufficient to accommodate normal anticipated 
pressure fluctuations. In this temperature range, a pressurizer 
bubble is normally present, which will minimize any pressure 
fluctuations, thereby limiting the possibility of a PORV actuation. 
Based on this, it is concluded that the proposed change will not 
impact the probability of occurrence that a PORV will be challenged.
    When the RHR relief valves are used for COPS there is no 
credible scenario which would result in excessive relief valve 
undershoot. This is because these valves are spring loaded relief 
valves which are designed to close whenever the RCS pressure 
decreases below the nominal setpoint of 440 psig [pounds per square 
inch gauge]. This provides assurance that there will be no damage to 
the seal of a running RCP.
    The proposed changes to the heatup/cooldown curves and the 
reduction in the enabling temperature for COPS only affect 
operational limits and can not be initiators of an event. The 
restrictions on RC [reactor coolant], RHR and ECCS pump operation 
can not result in an event initiator. Two separate operator actions 
are required to start an ECCS or RC pump. These two necessary 
actions as well as procedural controls are sufficient to prevent an 
inadvertent ECCS or RC pump start. De-energizing the RCPs when 
returning a loop to service can not initiate an event.
    The proposed change will provide an operable charging pump to 
ensure RCP seal flow and reactivity control will be available. When 
the RCP is in operation, the charging pump provides the preferred 
method for seal flow. The proposed change minimizes the time that 
this preferred method is interrupted. A loss of charging pump seal 
flow will not cause a malfunction of an RCP because the pump is 
designed to use RCS flow as an alternate method at these conditions. 
Not allowing two charging pumps to run simultaneously and requiring 
at least one pump to be in pull-to-lock, assures a second pump will 
not start on an inadvertent SI [safety injection] and exceed the 
assumptions in the Appendix G analysis or initiate a Boron Dilution 
or CVCS [chemical and volume control system] Malfunction event. If 
an operator were to inadvertently start the second pump, a failure 
of the charging throttle valve, FCV-121, and one relief valve 
credited for COPS would be necessary to exceed the assumptions in 
the Appendix G analysis. In addition, the actual time allowed for 
swapping the charging pumps is short. The remainder of the hour 
allows for documented verification of the disabling of the required 
pump. The proposed change will not change any control systems for 
these pumps or alter the system configuration that would affect the 
probability of an uncontrolled increase in charging flow. The 
procedure requirements to swap pumps and the likelihood of these

[[Page 52584]]

multiple failures occurring during the short duration allowed in 
this footnote provide adequate assurance that an overpressurization 
event will not occur. Maintaining at least one pump always operable 
makes the system more reliable for reactivity control than the 
current method which disables both pumps simultaneously.
    The proposed change to maintain one charging pump operable in 
Mode 4 [cannot] initiate an event because of the stable reactivity 
condition of the reactor, the emergency power supply requirement for 
the operable charging pump, and the fact that the plant is 
procedurally required to be borated to the highest required boron 
concentration for Modes 3, 4, or 5 prior to entering Mode 4. These 
changes do not effectively change the availability of plant 
equipment or the way that the plant is operated.
    The proposed change to substitute an RCS loop for an RHR loop 
during a planned heatup, can not initiate an event. The RCP will be 
verified as operating properly prior to stopping the RHR pump and as 
such will not initiate a loss of decay heat removal (by heating up 
to steam the SGs [steam generators])/loss of flow. While the RCP is 
in operation, it performs the RHR boron mixing function and the 
decay heat removal function is not required for heatup. Using the 
RCP to perform this function will not affect the probability that 
the RCP could fail because it will be operated within its normal 
operating design conditions. Aligning RHR in the ECCS lineup will 
not affect the probability of a RHR pump to start. The pump will be 
operable in this lineup. Currently in Mode 5, RHR is lost on a LOP 
[loss of offsite power] and is manually restarted once the diesel is 
running. With the proposed change, the RCP will be lost on a LOP and 
the RHR pump will have to be manually started. Thus, the proposed 
change does not affect the probability that the RHR pump could fail. 
Since the current response to a LOP is to manually restart the RHR 
pump, operator action is needed independent of this change. The 
proposed change allows normally open valves to be closed in Mode 5 
to align RHR for ECCS injection. This introduces additional manual 
actions which could extend the time required to establish flow. In 
addition, if one diesel generator were to fail, manual operation of 
a valve in the ESF [engineered safety features] building would be 
necessary. The mechanistic 'failure to open' of valves that is 
introduced by the change as well as the need for manual operator 
action to realign these valves increases the time to establish heat 
removal. However, there is sufficient time to re-establish RHR 
because this note applies only for a heatup in which the plant will 
have been shutdown for at least several hours which causes decay 
heat to be low (as compared to high decay heat immediately following 
a plant trip). Thus, it is concluded that there is no impact on the 
probability of failure of RHR to perform its required function.
    The proposed change to the ECCS wording does not result in any 
new failure modes that could initiate an event since manual 
realignment from the control room is currently allowed. Nor can the 
manual alignment of RHR valves initiate an event because this 
alignment is only for accident mitigation.
    Therefore, the proposed changes do not increase the probability 
of occurrence of previously evaluated accidents.
    Consequences of Previously Evaluated Accidents
    The revised Pressure/Temperature curves were calculated in 
accordance with 10CFR50, Appendix G, ASME Section XI, and Regulatory 
Guide 1.99, Revision 2. This provides assurance that an inadvertent 
overpressurization event will not result in a loss of RV integrity. 
The restrictions on RCP operation and the requirement to de-energize 
the RCPs in Modes 5 and 6 when returning a loop to service are 
consistent with the assumptions made in this Appendix G analysis and 
the RCPs are not required for accident mitigation for any previously 
evaluated accidents and therefore do not affect the consequences.
    The COPS relieving capability is greater than the maximum RCS 
pressurization rate resulting from any allowed pump combinations, 
and the PORV setpoints have been adjusted to take into account 
instrumentation effects. This will provide assurance that COPS will 
continue to perform its safety function. Since the COPS enabling 
temperature has been demonstrated to be conservative at 275F, 
allowing SI pump operability above 275F will have no impact on 
vessel non-ductile failure.
    The restriction between 275F and 350F on the SI and charging 
pumps, has been appropriately moved to the reactor coolant loop 
section to provide protection for the RHR system (RCS protective 
boundary) and to the cold overpressure protection section to provide 
protection for the RHR relief valves and the RCP seals. By 
incorporating this requirement previously located in the ECCS TS, 
RCS integrity is ensured.
    With the RCS less than 160F, the consequences of the PORV 
undershoot from the proposed PORV setpoints are that the RCS 
pressures may drop below the minimum requirement for RCP seal 
integrity. However, no seal damage will occur since a requirement 
has been added prohibiting the operation of RCPs below 160 deg.F 
with the PORVs not isolated while in the low setpoint mode. With 
cold overpressure relief valves in service above the COPS enable 
temperature (275 deg.F), restrictions are placed on the startup of 
an RCP and the number of ECCS pumps capable of injecting into the 
RCS to prevent unacceptable mass or energy addition transients. This 
provides assurance that the RHR relief valve capacity will not be 
exceeded and that PORV undershoot will not challenge the RCP 
<greek-i>1 seal. The restriction on the maximum number of ECCS pumps 
ensures that the integrity of the RHR relief valve bellows and the 
RCP seals during mass injection transients (i.e., inadvertent SI).
    The restrictions on RCS/SG secondary side temperature mismatch 
ensure that an unanalyzed energy addition event does not occur when 
an RCS loop is placed in operation.
    The consequences of a small break LOCA [loss of coolant 
accident] in COPS Mode 4 are not affected because the plant will 
continue to maintain one charging pump operable in Mode 4. In 
addition, additional options are provided in the bases of TS 3/
4.4.9.3 for disabling the required charging and SI pumps that will 
allow faster restoration if required to mitigate a LOCA or loss of 
RHR in Modes 4, 5 and 6.
    An RHR pump will remain available in Mode 4 with manual 
realignment from the control room as required to perform its ECCS 
safety function. The changes have no impact on the capability of RHR 
to function in the ECCS mode. RHR is credited during a safety grade 
cold shutdown. The proposed change assures that the RHR system will 
be available to perform its heat removal function during a safety 
grade cold shutdown and thus, there is no change in the analysis 
assumptions or consequences.
    The changes also eliminate an inconsistency between the charging 
system operability requirements for boration and the charging system 
operability requirements for cold overpressure protection. The 
requirement to maintain two charging pumps operable in Mode 4 will 
be reduced to one charging pump. As stated in the proposed basis 
section, a second method of boration is not required to be OPERABLE 
in Mode 4 for single failure considerations based on the stable 
reactivity condition of the reactor, the emergency power supply 
requirement for the operable charging pump, and the fact that the 
plant is procedurally required to be borated to the highest required 
boron concentration for Modes 3, 4, or 5 prior to entering Mode 4. 
This provides assurance that reactivity control will be maintained 
and stable while only one charging pump is operable for cold 
overpressure concerns. These changes do not effectively change the 
availability of plant equipment or the way that the plant is 
operated. The changes will not adversely impact the assumption for 
the limiting dilution flow path and flow rate and therefore, the 
consequences of a boron dilution event are not affected.
    The proposed changes will maintain a charging pump operable for 
reactivity control while ensuring that the flow limits in the 
Appendix G analyses are not exceeded. Remaining within the bounds of 
the Appendix G limits ensures reactor vessel integrity in Mode 4. 
Since the change maintains the reactor vessel integrity, it does not 
introduce any means of releasing radionuclides post-accident. The 
consequences of a small break LOCA in Mode 4 are not affected 
because the plant will continue to maintain one charging pump 
operable in Mode 4. These changes are reflected in TS 3.1.2.1, 
3.1.2.2, 3.1.2.3 and 3.1.2.4. Adequate protection is provided for 
reactor vessel integrity while maintaining reactivity control 
operability.
    In Mode 5, RHR requirements are specified for decay heat removal 
in the case of a loss of offsite power but none are specified for 
ECCS accident mitigation. The first RHR train will be aligned for 
injection prior to taking the second train out of service. This 
provides assurance that this train will be available if needed in 
Mode 5. Currently in Mode 5, following a LOP the RHR system can be 
re-established by restarting the RHR pump once the diesel is 
running. No valve manipulations

[[Page 52585]]

are necessary. With the proposed change, when the operating RCP 
trips following a LOP, some of the RHR valves must be realigned from 
the ECCS to heat removal mode. If one diesel generator were to fail, 
manual operation of a valve in the ESF building would be necessary. 
Since this footnote is only applicable during a heatup, decay heat 
will be low. There is sufficient time to re-establish RHR even if 
action outside the control room is necessary. Since there are four 
operable RCS loops, a bubble drawn in the pressurizer and the RCS 
pressurized, the plant will heat up to Mode 4 and natural 
circulation will provide core cooling if the RHR system cannot be 
re-established. Thus, decay heat removal is assured and there is no 
affect on the consequences of a LOP.
    Since the structural integrity of the RCS is maintained and 
adequate core cooling and reactivity control will be available for 
design basis events, the proposed changes will have no adverse 
impact on the consequences of previously evaluated accidents.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The temperature/pressure limits will continue to meet the 
requirements of 10CFR50, Appendix G. Since the new limits continue 
to provide assurance of reactor vessel integrity, the proposed 
change does not create the possibility of an accident of a different 
type than previously evaluated. Adequate RCS pressure-relieving 
capabilities will continue to be maintained throughout the shutdown 
modes. No new malfunctions will be introduced which could result in 
a new accident postulated in Modes 3-5.
    The restrictions on RCP operation do not create the potential 
for unanalyzed heat injection transient as a result of an 
inadvertent RCP start because two operator actions are required to 
start a pump. The requirement to have all RCPs de-energized, prior 
to unisolating a loop adds additional assurance that an energy 
addition transient will not occur.
    The proposed change to allow 2 charging pumps to be operable 
does not create an accident of a different type because there will 
be adequate controls to ensure that the second pump does not 
inadvertently start and initiate an increase in RCS inventory or a 
boron dilution. Procedural controls will minimize the amount of time 
that both charging pumps are operable and at no time will two pumps 
be out of pull-to-lock.
    The proposed footnote to TS 3.4.1.4.1 to remove RHR heat removal 
from operation allows normally open valves to be closed in Mode 5 to 
align RHR for ECCS injection. This introduces 'failure to open' as a 
potential mechanistic failure malfunction in the RHR system. This is 
a malfunction of a different type since previously stroking of these 
valves was not needed to establish RHR. The current response to a 
LOP is to manually restart the RHR pump only, with no valve 
manipulations required. The proposed change adds the manual action 
of realigning the valves. Since operator action to re-establish RHR 
following a LOP is required independent of the proposed changes, 
crediting operator action does not create the potential for a 
malfunction of a different type. Allowing both trains of RHR to be 
out of service does not create a different accident because 
additional requirements have been specified for RCS loop operability 
and at least one RHR pump is operable for ECCS when the core cooling 
requirement is being met by crediting RCS loop operability. Meeting 
the Mode 4 TS conditions prior to heatup, ensures two diesels are 
operable. As such, a single failure would only require one valve to 
be manually realigned in the ESF building. Adequate time is 
available to accomplish these actions since this note only applies 
during heatup, when decay heat is very low. Further, with four RCS 
loops operable and a bubble drawn in the pressurizer and the RCS 
pressurized, the steam generators can be used for core cooling via 
natural circulation once the plant heats up to Mode 4, in the event 
the RHR cannot be re-established. Since core cooling will be assured 
if a LOP occurred during heatup in Mode 5, the change in plant 
response to this event does not constitute an accident of a 
different type.
    The proposed changes to TS 3.5.3.f to manually realign the ECCS 
valves is no different from what is currently evaluated. During a 
Mode 4 LOCA adequate procedural guidance is provided to ensure that 
RHR will be realigned for injection. The proposed change allows RHR 
to be aligned to perform its safety grade cold shutdown heat removal 
function.
    Therefore, the proposed revision does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The new proposed curves raises the lower bound on RCS 
temperature, resulting in increased RCS ductility and therefore 
increased structural margin against non-ductile failure. The new 
curves take into account the dynamic pressure effects identified in 
NRC Information Notice 93-58 and are calculated in accordance with 
10CFR50 Appendix G, ASME Section XI and Regulatory Guide 1.99, 
Revision 2. These changes to the P/T [pressure/temperature] limits 
are reflected in TS 3.4.9.1. Additional restrictions have been 
placed on RCP operation to ensure that assumptions used in 
developing the curves remain bounding. These are also reflected in 
TS 3.4.1.3, 3.4.1.4.1, 3.4.1.4.2 and 3.4.1.6. As such, the curves 
will continue to provide the required assurance for reactor vessel 
integrity.
    The COPS enable temperature is proposed to be lowered from the 
current 350 deg.F to 275 deg.F which provides a margin of 31F above 
that required by NRC Branch Technical Position RSB 5-2. The 
reduction of the COPS enabling temperature eliminates the need for 
COPS to be operable in Mode 3. This will simplify the transition 
between Mode 3 and Mode 4.
    Additional changes have been made to the Overpressure Protection 
TS to ensure that the assumptions made in the Appendix G 
calculations remain bounding. These include additional restrictions 
on charging pump and SI pump operability and the modification of the 
PORV setpoints. The pump requirements have been transferred from the 
ECCS specification and expanded to cover Modes 4, 5 and 6. In 
addition, these same pump restrictions have been included in TS 
3.4.1.3 whenever RHR is in service. This provides added assurance 
that the RHR piping will not be overpressurized by an inadvertent 
actuation of an SI or charging pump. Additional actions and 
surveillances have been provided to assure that assumptions on 
charging pump and SI pump operability will be met. The additional 
options for assuring the inoperability of the SI and charging pumps 
require two distinct operator actions to restore injection 
capability from these pumps. Thus, these options are equivalent in 
providing assurance that an inadvertent injection will not occur 
while at the same time allowing faster restoration if needed to 
mitigate a loss of RHR.
    A requirement to have all RCPs de-energized, prior to 
unisolating a loop is added to TS 3.4.1.6.c, to ensure that loop 
flow will not be initiated which results in an energy addition 
transient from the secondary side of the SG being unisolated. This 
change will preclude RCS overpressurization when an idled loop is 
returned to service and SG secondary side temperature is greater 
than the RCS temperature.
    The PORV setpoints were established to ensure that the P/T limit 
curves are not exceeded as a result of a single operator action or 
as a result of a single equipment malfunction, as required by the 
current system design basis criteria (i.e., SRP [standard review 
plan] Branch Technical Position RSB 5-2).
    A clarification of the hydrostatic and leak test requirements 
ensures a uniform reactor vessel temperature for the test. A 72 hour 
time limit is placed on the performance of engineering evaluations 
of out of specification condition. This provides added assurance for 
RPV [reactor pressure vessel] integrity.
    The changes also eliminate an inconsistency between the charging 
system operability requirements for boration and the charging system 
operability requirements for cold overpressure protection. These are 
reflected in TS 3.1.2.1, 3.1.2.2, 3.1.2.3 and 3.1.2.4. The Bases 
requirement to maintain two charging pumps operable in Mode 4 will 
be reduced to one charging pump. As stated in the proposed basis 
section, a second method of boration is not required to be OPERABLE 
in Mode 4 for single failure considerations based on the stable 
reactivity condition of the reactor, the emergency power supply 
requirement for the operable charging pump, and the fact that the 
plant is procedurally required to be borated to the highest required 
boron concentration for Modes 3, 4, or 5 prior to entering Mode 4. 
This provides assurance that reactivity control will be maintained 
and stable while only one charging pump is available. The additional 
options for disabling the charging pump (provided in the bases for 
TS 4.4.9.3.5) will allow for faster restoration when needed while 
maintaining two distinct operator

[[Page 52586]]

actions to prevent a second pump from being started. This provides 
added assurance that reactor vessel integrity will be maintained.
    Procedures will minimize the amount of time that both charging 
pumps are operable and having at least one pump in pull-to-lock will 
ensure that the second pump does not inadvertently start and exceed 
the Appendix G analysis limits and thus, ensure reactor vessel 
integrity.
    The TS bases for requiring RHR in Mode 5 is to remove decay heat 
and provide RCS circulation. Since the RCP can perform the RHR 
circulation function and the decay heat removal function is not 
required during heatup, the proposed change is consistent with the 
bases. Since this option is only allowed during heatup where decay 
heat is low, sufficient time will be available to re-establish RHR 
heat removal as required to mitigate a LOP in Mode 5. Further, with 
the RCS pressurized, four RCS loops operable and the SG filled, core 
cooling can be accomplished by the steam generators via natural 
circulation once the plant heats up to Mode 4, in the event that RHR 
cannot be re-established. Therefore, the design basis analyses 
remain limiting and the margin of safety is not reduced.
    The original plant design allows the RHR pumps to be available 
for both heat removal while shutdown and ECCS. As such, an 
allowance, TS 3.5.3.f, was provided to allow manual realignment from 
heat removal to ECCS mode. The specific wording of TS 3.5.3.f 
implies that this realignment only involves the suction valves. 
Since discharge valves must also be realigned, the TS is being 
reworded to apply for the discharge as well as suction valves. 
Therefore, this change is a clarification of the existing TS.
    The proposed changes do not impact the protective boundaries 
(reactor vessel integrity) nor any of the design basis accidents.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: July 25, 1997
    Description of amendment request: The proposed amendment request 
would implement 10 CFR Part 50 Appendix J, Option B by revising the 
Technical Specifications (TS) to allow the frequency of conducting 
integrated leak rate testing (ILRT) and local leak rate testing (Type B 
and C) to be based on component performance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change implements Option B of 10 CFR Part 50 
Appendix J on performance-based containment leakage testing. The 
proposed change does not involve a change to the plant design or 
operation. As a result, the proposed change does not affect any 
parameters or conditions that contribute to the initiation of any 
accidents previously evaluated. The proposed change potentially 
affects the leak-tight integrity of the containment structure 
designed to mitigate the consequences of a Loss-of-Coolant Accident 
(LOCA). The function of the containment is to maintain functional 
integrity during and following the peak transient pressures and 
temperatures and limit fission product leakage following the design 
basis LOCA. Because the proposed change does not alter the plant 
design, only the frequency of measuring Type A, B, and C leakage, 
the proposed change does not directly result in an increase in 
containment leakage.
    Test intervals will be established based on the performance 
history of components being tested. The frequency of monitoring the 
relatively few containment isolation valves and/or containment 
penetrations subject to above normal leakage will not decrease by 
implementing Option B of Appendix J. A performance based program 
will identify those valves and penetrations which must continue to 
be tested each refueling outage.
    The risk resulting from the proposed changes is characterized as 
follows, based primarily on the results contained in NUREG-1493 
``Performance-Based Containment Leakage Test Program,'' the 
principal Technical Support Document used by the NRC as the basis 
for the Appendix J Final Rule:
    Type A Testing
    NUREG-1493 found that the effect of containment leakage on 
overall accident risk is minimal since risk is dominated by accident 
sequences that result in failure or bypass of the containment. 
Industry wide, Integrated Leak Rate Tests (ILRTs) have only found a 
small fraction of the leaks that exceed current acceptance criteria. 
Only three percent of all leaks are detectable only by ILRTs, and 
therefore, by extending the Type A testing intervals, only three 
percent of all leaks have a potential for remaining undetected for 
longer periods of time. In addition, when leakage has been detected 
by ILRTs, the leakage rate has been only marginally above existing 
requirements. The Fort Calhoun Station Unit No. 1 Type A testing 
confirms the industry-wide experience that a majority of the leakage 
experienced during Type A testing is through components tested by 
Type B and C tests.
    NUREG-1493 found that these observations, together with the 
insensitivity of reactor accident risk to the containment leakage 
rate, show that increasing the Type A leakage test intervals would 
have a minimal impact on public risk.
    Type B and C Testing
    NUREG-1493 found that while Type B and C tests can identify the 
vast majority (greater than 95 percent) of all potential leakage 
paths, performance-based alternatives to current local leakage-
testing requirements are feasible without significant risk impacts. 
The risk model used in NUREG-1493 suggests that the number of 
components tested would be reduced by about 60 percent with less 
than a three-fold increase in the incremental risk due to 
containment leakage. Since, under existing requirements, leakage 
contributes less than 0.1 percent of overall accident risk, the 
overall impact is very small. In addition, the NRC's Final 
Regulatory Impact Analysis concluded that while the extended testing 
intervals for Type B and C tests led to minor increases in potential 
offsite dose consequences, the beneficial expected decrease in 
onsite worker dose received during ILRT and local leak rate testing 
exceeds (by at least an order of magnitude) the potential off-site 
dose consequences.
    Therefore, the proposed change will not result in a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There will be no physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
implementation of setpoints or limits as a result of this proposed 
change. As a result, the proposed change does not affect any of the 
parameters or conditions that could contribute to initiation of any 
accidents.
    This change involves the reduction of Type A, B, and C test 
frequency. Except for the method of defining the test frequency, the 
methods for performing the actual tests are not changed. No new 
accident modes are created by extending the testing intervals. No 
safety-related equipment or safety functions are altered as a result 
of this change. Extending the test frequency has no influence on, 
nor does it contribute to, the possibility of a new or different 
kind of accident or malfunction from those previously analyzed. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.

[[Page 52587]]

    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change only affects the frequency of Type A, B, and 
C testing. Except for the method of defining the test frequency, the 
methods for performing the actual tests are not changed.
    The frequency of monitoring the relatively few containment 
isolation valves and/or containment penetrations subject to above 
normal leakage will not decrease by implementing Option B of 
Appendix J. A performance based program will identify those valves 
and penetrations which must continue to be tested each refueling 
outage. NUREG-1493 has determined that, under several different 
accident scenarios, the increased risk of radioactivity
    release from containment is negligible with the implementation 
of these proposed changes.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to containment 
leakage rate. The containment isolation system is designed to limit 
leakage to La, which is stated in the Fort Calhoun Station Unit No. 
1 Technical Specifications to be 0.1 percent by weight of the 
containment air per 24 hours at 60 psig.
    The limitation on containment leakage rate is designed to ensure 
that total leakage volume will not exceed the value assumed in the 
accident analyses at the peak accident pressure. The margin to 
safety for the offsite dose consequences of postulated accidents 
directly related to the containment leakage rate is maintained by 
meeting the 1.0 La acceptance criteria. The La value is not being 
modified by this proposed change.
    Except for the method of defining the test frequency, no change 
in the method of testing is being proposed. The Type B and C tests 
will continue to be done at 60 psig or greater. Other programs are 
in place to ensure that proper maintenance and repairs are performed 
during the service life of the primary containment and systems and 
components penetrating the primary containment.
    Therefore, the proposed change will not result in a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of amendment request: August 26, 1997
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 4.6.5.3.1b, for the Filtration, 
Recirculation and Ventilation System (FRVS), Ventilation Subsystem, and 
TS 4.6.5.3.2b for the FRVS Recirculation Subsystem. The revised TSs 
would state that the heaters should be ``operating (automatic heater 
modulation to maintain relative humidity)'' instead of ``on'' when 
performing the 10-hour, monthly test.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS revisions involve no hardware changes and no 
changes to existing structures, systems or components. Conducting TS 
Surveillance Requirements 4.6.5.3.1.b and 4.6.5.3.2.b with the FRVS 
recirculation unit and ventilation unit heaters in automatic 
modulation to maintain the relative humidity within the design 
requirements, meets the intent of the USNRC Regulatory Guide 1.52, 
position C.4.d, in reducing adsorber and HEPA filter moisture 
levels. In the unlikely event that the adsorber and HEPA filters, 
that are enclosed and isolated in a confined space should reach an 
equilibrium at the maximum design operating humidity level, the 10 
hour run with heaters energized would reduce the humidity to 
acceptable levels. Therefore, the proposed changes do not change the 
post-accident performance characteristics of the FRVS adsorber or 
HEPA filters below the design requirements and does not increase the 
consequences of accidents previously identified. Since there are no 
changes to the operation of FRVS in normal or post-accident 
operating conditions, there is no increase in the probability of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes contained in this submittal will not 
adversely impact the operation of any safety related component or 
equipment. PSE&G has concluded that [the] method of performing the 
monthly FRVS recirculation unit and ventilation unit surveillances 
with the heaters modulating adequately maintains and demonstrates 
operability of FRVS. Since the proposed changes involve: 1) no 
hardware changes; 2) no changes to FRVS operation in normal 
operating or post-accident conditions; and 3) no changes to existing 
structures, systems or components, there can be no impact on the 
potential occurrence of any accident. Furthermore, there is no 
change in plant testing proposed in this change request which could 
initiate an event. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The revisions to TS Surveillance Requirements 4.6.5.3.1.b and 
4.6.5.3.2.b provide a more accurately defined basis for performing 
this surveillance test. The proposed changes reflect PSE&G's 
position on satisfying USNRC Regulatory Guide 1.52, position C.4.d. 
Since PSE&G has concluded that performing TS Surveillance 
Requirements 4.6.5.3.1.b and 4.6.5.3.2.b with the FRVS recirculation 
unit and ventilation unit heaters in automatic moduation [sic] 
[modulation] to maintain the relative humidity within the design 
requirements, adequately reduces adsorber and HEPA filter moisture 
levels, the proposed changes do not significantly reduce a margin of 
safety in FRVS. Since the FRVS recirculation units and ventilation 
units will continue to be tested with the heaters: 1) operable; and 
2) set at the demand necessary to ``reduce the buildup of 
moisture,'' PSE&G believes that the proposed changes to clarify the 
TS are justified.
    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: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038
    NRC Project Director: John F. Stolz

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

    Date of amendment request: August 19, 1997
    Description of amendment request: The proposed amendment would 
revise the Ginna Station Improved Technical Specifications (ITS) by 
revising the Emergency Core Cooling System Accumulators Surveillance 
Requirement 3.5.1.2 to correct the specified accumulator borated water 
volume values in order to match the associated accumulator percent 
level values.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 52588]]

issue of no significant hazards consideration, which is presented 
below:
    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The change is 
only to correct a conversion error with respect to accumulator 
borated water volume. This does not increase the probability of any 
accident previously evaluated since the accumulator water volume 
provides mitigation capability only (i.e., does not initiate any 
accident). The affected accident analyses with respect to the 
accumulator (e.g., small and large [loss-of-coolant] LOCA and steam 
line break) have been re-evaluated using the correct accumulator 
water volume values with acceptable results. Therefore, these 
changes do not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. Ginna Station 
operators verify accumulator water volume via percent level (versus 
cubic feet) which remains unchanged. Thus, this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes only correct a conversion error. The 
error has been re-evaluated with acceptable results. As such, no 
question of safety is involved, and the change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Project Director: Alexander W. Dromerick, Acting Director

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: February 14, 1997, as supplemented by 
letters dated June 20, August 5, and September 22, 1997
    Description of amendments request: The proposed amendments would 
change the maximum reactor core power level for facility operation from 
2652 megawatts thermal (MWt) to 2775 MWt in the Farley, Units 1 and 2, 
Facility Operating Licenses. In addition, the proposed amendments would 
involve the following Technical Specification (TS) changes.
    The defined rated thermal power for Farley; departure from nucleate 
boiling (DNB) parameters for reactor coolant system (RCS) average 
temperature (T<INF>avg</INF>); pressurizer pressure; and RCS flow would 
be changed.
    The reactor trip system interlock setpoint for power range neutron 
flux (P-8) and engineered safety features (ESF) actuation trip setpoint 
for steam generator water high-high level for turbine trip and 
feedwater isolation (P-14), and ESF actuation system interlock for low-
low T<INF>avg</INF> (P-12) would be modified to reflect analytical 
results.
    An evaluation of additional reactor trip system and ESF actuation 
system safety analysis limits and trip setpoints would result in 
changes to the allowable values for several functions.
    On the basis of the results of new containment analyses, the 
maximum peak calculated containment internal pressure for a loss-of-
coolant accident (LOCA) event would be revised. The main steamline 
isolation valve closure time requirement would be revised. Surveillance 
requirements for emergency core cooling systems (ECCS) would be 
modified to reflect reduced ECCS flows. The number of secondary system 
hydrostatic pressure tests (Table 5.7-1) would be increased. For Farley 
Unit 2 only, the steam generator F* distance would be revised.
    Changes to the plant design features and administrative controls 
are also proposed. These changes would revise the RCS fluid volume 
contained in Section 5.4 and the addition of the NRC-approved 
references for best estimate LOCA listed in Section 6.9.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    DEFINITION, DESIGN FEATURE AND ADMINISTRATIVE CONTROL CHANGES
    * * * *
    1. The proposed changes to the rated thermal power definition, 
RCS fluid volume, and COLR [Core Operating Limit Report] references 
do not increase the probability or consequences of an accident 
previously evaluated in the FSAR [Final Safety Analysis Report]. The 
comprehensive analytical efforts performed to support the proposed 
uprating included a review and evaluation of all components and 
systems (including interface systems and control systems) that could 
be affected by this change. The revised power uprate value and RCS 
fluid volume were inputs to applicable safety analyses. All systems 
will function as designed, and all performance requirements for 
these systems have been evaluated and found acceptable. None of 
these proposed changes directly initiate any accident; therefore, 
the probability of an accident has not increased. All dose 
consequences have been analyzed or evaluated with respect to these 
parameters, and all acceptance criteria continue to be met. 
Therefore, the consequences of an accident previously evaluated in 
the FSAR have not increased.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident than any accident already evaluated in 
the FSAR. No new accident scenarios, failure mechanisms or limiting 
single failures are introduced as a result of the proposed changes. 
The proposed technical specification changes have no adverse effects 
on any safety-related system and do not challenge the performance or 
integrity of any safety-related system. Therefore, the possibility 
of a new or different kind of accident is not created.
    3. The proposed operating license and technical specification 
changes do not involve a significant reduction in a margin of 
safety. All analyses supporting the proposed power uprate reflect 
the RCS fluid volume and rated thermal power values. The use of NRC 
approved BELOCA [best estimate LOCA] methodology must be referenced 
since BELOCA will now be the LBLOCA [large break LOCA] analysis 
licensing basis for FNP [Farley Nuclear Plant]. All acceptance 
criteria (including LOCA peak clad temperature, DNB criteria, 
containment temperature and pressure, and dose limits) continue to 
be met. Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    DNB PARAMETERS CHANGES
    * * * *
    1. The proposed technical specification changes for DNB 
parameters do not involve a significant increase in the probability 
or consequences of an accident previously evaluated in the FNP FSAR. 
The mechanical design features associated with VANTAGE 5 fuel and 
the improved methodologies (such as Revised Thermal Design 
Procedure) provide capability for relaxation of analytical input 
parameters such that increased DNBR [DNB ratio] margin can be 
generated without violation of any acceptance criteria. The 
indicated DNB parameters bound the analytical values used to support 
the proposed uprating. In each case, the appropriate design and 
acceptance criteria are met. All performance requirements for any 
system or component have been evaluated and support the revised 
analysis assumptions. Overall plant integrity is not reduced. 
Furthermore, the parameter changes are associated with features used 
as limits or mitigators to assumed accident scenarios and are not 
accident initiators. Therefore, the probability of an accident has 
not significantly increased.

[[Page 52589]]

    The radiological consequences of accidents previously evaluated 
in the FSAR have been assessed due to the proposed technical 
specification changes. Evaluations have confirmed that the doses 
remain within previously approved acceptable limits as well as those 
defined by 10 CFR [Part] 100. Therefore, the radiological 
consequences to the public resulting from any accident previously 
evaluated in the FSAR has not significantly increased.
    2. The proposed technical specification changes do not create 
the possibility of a new or different kind of accident from any 
previously evaluated in the FSAR. No new accident scenarios, failure 
mechanisms, or limiting single failures are introduced as a result 
of the revised DNB parameters. The revised analytical assumptions 
have no adverse effect and do not challenge the performance of any 
other safety-related system. This has been verified in WCAP 12771, 
Rev. 1. Therefore, the possibility of a new or different kind of 
accident is not created.
    3. The proposed technical specification changes do not involve a 
significant reduction in the margin of safety. The margin of safety 
for fuel-related parameters (such as DNB and Kw/ft) are defined in 
the Bases to the Technical Specifications. The uncertainties 
associated with the proposed DNB parameter changes are included in 
the core safety limits. Performance of analyses and evaluations with 
the reactor core safety limits defined by RTDP [Revised Thermal 
Design Procedure] have confirmed that the operating envelope defined 
by the Technical Specifications continues to be bounded by the 
revised analytical basis, which in no case exceeds the acceptance 
limits. Therefore, the margin of safety provided by the analyses in 
accordance with these acceptance limits is not reduced.
    MISCELLANEOUS OPERATION AND MARGIN ENHANCEMENT CHANGES
    * * * *
    1. The proposed changes do not increase the probability or 
consequences of an accident previously evaluated in the FSAR. 
Explicit modeling of these parameters is included in the uprate 
analyses and evaluations. The comprehensive analytical effort 
performed to support the proposed uprating has included a review and 
evaluation of all components and systems (including interface 
systems and control systems) that could be affected by this change. 
In addition LOCA and non-LOCA analyses and evaluations have verified 
that all acceptance criteria continue to be met. All systems will 
function as designed. None of these proposed changes can directly 
initiate any accidents; therefore, the probability of an accident 
has not been increased. All dose consequences have been analyzed or 
evaluated with respect to these parameters, and all acceptance 
criteria continue to be met. Therefore, the consequences of an 
accident previously evaluated in the FSAR have not increased.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident than any accident already evaluated in 
the FSAR. No new accident scenarios, failure mechanisms or limiting 
single failures are introduced as a result of the proposed changes. 
The proposed technical specification changes have no adverse effects 
on any safety-related system and do not challenge the performance or 
integrity of any safety-related system. Therefore, the possibility 
of a new or different kind of accident is not created.
    3. The proposed technical specification changes do not involve a 
significant reduction in a margin of safety. All analyses supporting 
the proposed power uprate reflect these proposed values. All 
acceptance criteria (including LOCA peak clad temperature, DNB 
criteria, containment temperature and pressure, and dose limits) 
continue to be met. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.
    ALLOWABLE VALUES AND TRIP SETPOINTS FOR REACTOR TRIP SYSTEM AND 
ESFAS [ENGINEERED SAFETY FEATURE ACTUATION SYSTEM]
    * * * *
    1. The proposed changes do not increase the probability or 
consequences of an accident previously evaluated in the FSAR. The 
comprehensive engineering effort performed to support the proposed 
uprating has included evaluations or reanalysis of all accident 
analyses including all dose related events. Setpoint calculations 
have verified acceptability of the proposed setpoints and allowable 
value changes. All systems will function as designed, and all 
performance requirements on these systems have been verified to be 
acceptable. Neither allowable values nor the setpoints initiate any 
accident; therefore, the probability of an accident has not been 
increased. All dose consequences have been analyzed or evaluated 
with respect to these parameters, and all acceptance criteria 
continue to be met. Therefore the consequences of an accident 
previously evaluated in the FSAR have not increased.
    2. The proposed setpoints and allowable value changes do not 
create the possibility of a new or different kind of accident than 
any accident already evaluated in the FSAR. No new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed changes. The proposed 
technical specification changes have no adverse effects on any 
safety-related system and do not challenge the performance of 
integrity of any safety-related system. The specified trip setpoints 
associated with the respective RTS [Reactor Trip System] and ESFAS 
functions ensure all accident analyses criteria continue to be met. 
Therefore, the possibility of a new or different kind of accident is 
not created.
    3. The proposed technical specification changes do not involve a 
significant reduction in a margin of safety. All analyses supporting 
the proposed power uprate reflect these proposed values. Setpoint 
calculations demonstrate that margin exists between the setpoint and 
the corresponding safety analysis limits. The calculations are based 
on FNP instrumentation and calibration/functional test methods and 
include allowances for uprated power conditions. All acceptance 
criteria (including LOCA peak clad temperature, DNB criteria, 
containment temperature and pressure, and dose limits) continue to 
be met. Therefore, the proposed changes do 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: 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: Herbert N. Berkow

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

    Date of amendment request: September 17, 1997 (TS 97-02)
    Description of amendment request: The proposed changes would revise 
Section 4.6.2.1 of the Sequoyah Technical Specifications (TS) to change 
the parameters to be monitored during the inservice inspection 
surveillance testing of the containment spray system pumps. The changes 
would also adopt provisions in the Westinghouse Improved Standard TS 
(NUREG-1431) that affect that section of the Sequoyah TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revisions to the containment spray system 
surveillances for the pumps, valves, and nozzles do not change the 
intent of the current TS requirements. These revisions only affect 
the TS operability testing requirements without changing the system 
functions. These functions are not considered to be accident 
initiators. The proposed surveillance wording is not based on 
changes to the plant although a modification to flow orifices for 
the containment spray pumps created the need to revise the 
surveillance that verifies pump developed head. The revisions 
primarily provide flexibility for required methods to verify system 
operability as well as utilizing less prescriptive operability 
limits and conditions for testing. The testing flexibility and less 
prescriptive requirements do not

[[Page 52590]]

relax the intent to properly verify operability of the containment 
spray system but do allow for changes in testing that continue to 
ensure the appropriate operability requirements. Since these 
revisions are not directly related to modifications of the plant or 
result in different methods for operating the plant, there is no 
change that could increase the probability of an accident. In 
addition, the consequences of an accident are not increased because 
there has not been a change that would impact the safety functions 
of the containment spray system. These revisions will continue to 
properly verify the operability of the containment spray system.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The containment spray system functions are not changed as 
discussed above and the operating practices for the plant remain the 
same. The testing methods can be modified as a result of the 
proposed revisions but will continue to maintain appropriate 
verifications of system operability. These testing methods as well 
as the containment spray system are not considered to be a potential 
initiator of accidents. Therefore, these revisions will not impact 
the operation of systems that could initiate an accident and the 
possibility of a new or different kind of accident is not created.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed revisions do not directly change the limits for 
containment spray system operability although they do provide the 
flexibility to properly revise limits resulting from system 
modifications. This type of limit revision would be necessary to 
adequately verify system operability. The appropriate limits 
continue to be required by the proposed TS surveillance 
requirements. Therefore, the proposed revisions do not allow 
inappropriate changes to setpoints or operating requirements that 
maintain the margin of safety and no reduction in this margin is 
involved in this request.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: August 26, 1997
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.2, ``Power Distribution 
Limits.'' The DNB Parameters Limiting Condition for Operation would be 
modified consistent with an industry notification.
    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 Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit No. 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
assumptions or probabilities are affected by the proposed change. 
The proposed change corrects a nonconservative Technical 
Specification Action statement by removing provisions which allow 
continued Mode 1 plant operation in the event the Reactor Coolant 
System flow rate is less than the required value. Under the proposed 
change, a power reduction to less than 5 percent of rated thermal 
power (Mode 2) will be required if the Reactor Coolant System flow 
rate is less than the required Technical Specification value.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed change does not 
affect any equipment, accident conditions, or assumptions which 
could lead to a significant increase in radiological consequences of 
an accident. The proposed change will ensure accident analyses 
remain valid if the Reactor Coolant System flow rate becomes less 
than the required value.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators will be introduced by the proposed change. No 
equipment or operations will be affected.
    3. Not involve a significant reduction in a margin of safety 
because under the proposed Technical Specification Action statement 
a power reduction to less than 5 percent of rated thermal power 
(Mode 2) will be required if degraded Reactor Coolant System flow 
develops. The proposed Action statement ensures accident analyses' 
assumptions are maintained.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: August 22, 1997, as supplemented by 
letter dated September 18, 1997
    Description of amendment request: The proposed amendment would 
revise the Vermont Yankee Technical Specifications (TSs) to address the 
new low pressure C0<INF>2</INF> suppression system for the East and 
West Switchgear Rooms and more clearly describe the separation of the 
rooms.
    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 proposed changes support the use of a newly installed low 
pressure C0<INF>2</INF> suppression system for the East and West 
Switchgear Rooms, to meet the C0<INF>2</INF> concentration 
requirements of NFPA 12 (1993) following detection of a fire 
condition in one of the associated rooms. The new low pressure 
C0<INF>2</INF> system consists of a 6 ton storage tank, piping, 
valves, associated instrumentation and controls.
    The FSAR [Final Safety Analysis Report] was reviewed for impact 
as a result of this proposed amendment with none being found. The 
initiators of the four design basis accidents, as defined in section 
14.6 of the FSAR, were reviewed with respect to the new low pressure 
C0<INF>2</INF> system. The low pressure C0<INF>2</INF> system is not 
an initiator of any of the Chapter 14.6 accidents. The low pressure 
C0<SUP>T22</SUP> suppression system is classified as a Non Nuclear 
Safety (NNS) related system. However, the C0<INF>2</INF> dispersion 
headers have been seismically mounted to preclude the possibility of 
their failure affecting safety related equipment during a seismic 
event. Although the Switchgear Room (East and West) low pressure 
C0<INF>2</INF> system is not used as a mitigator of any accident 
listed in section 14.6 of the FSAR, the switchgear contained in the 
aforementioned rooms is used to

[[Page 52591]]

mitigate the consequences of the section 14.6 accidents.
    The new low pressure C0<INF>2</INF> system, which meets NFPA 12 
(1993), provides fire suppression for the affected room by raising 
the C0<INF>2</INF> concentration to a 50% level and maintains this 
concentration for a 20 minute duration upon initiation. As a result, 
this C0<INF>2</INF> system prevents a fire in the affected room from 
spreading to adjacent rooms and adversely impacting the adjacent 
room's safety related equipment. Consequently, the unaffected rooms 
and associated trains of equipment remain functional to perform 
their intended safety functions if required. The proposed amendment 
also reflects the separation of the switchgear room into two fire 
areas with equivalent detection and suppression.
    Based on the above, use of the low pressure C0<INF>2</INF> 
system for East or West Switchgear Room fire suppression does not 
create new initiators, nor degrade the effectiveness of equipment 
relied upon to perform mitigative functions assumed for the 
previously evaluated design basis accidents. Therefore, the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated:
    The NNS low pressure C0<INF>2</INF> system, which meets NFPA 12 
(1993), provides fire suppression for the East and West Switchgear 
Rooms by raising the C0<INF>2</INF> concentration to a 50% level and 
maintains this concentration for a 20 minute duration upon 
initiation. As a result, this C0<INF>2</INF> system prevents a fire 
in the affected switchgear rooms from spreading to adjacent rooms 
and adversely impacting the adjacent rooms associated equipment. The 
switchgear room is more clearly depicted as two separate fire areas 
in the proposed amendment with equivalent protection. The 
C0<INF>2</INF> suppression header piping located in the switchgear 
rooms is seismically supported, which precludes the possibility of 
this piping failing during a seismic event and affecting safety 
related equipment located nearby.
    The new low pressure C0<INF>2</INF> system does not introduce 
new accident initiators. The low pressure C0<INF>2</INF> system is 
fulfilling the fire suppression function previously performed by the 
existing high pressure C0<INF>2</INF> system. The previous 
separation of the switchgear room into two separate fire areas, 
provides separation of redundant equipment and equivalent fire 
detection and suppression for that equipment. The low pressure 
C0<INF>2</INF> system consists of a 6 ton storage tank, piping, 
valves, and associated instrumentation and controls. There are no 
failure mechanisms, associated with the new low pressure 
C0<INF>2</INF> equipment, which cannot be categorized under at least 
one of the three failure mechanisms identified in section 14.4.3 of 
the FSAR. Consequently, the proposed amendment will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    Technical Specifications 3.13.D/4.13.D were reviewed with 
respect to the proposed amendment to determine if the changes would 
result in a reduction in a margin of safety. The proposed amendment, 
to allow use of a low pressure C0<INF>2</INF> suppression system for 
the East or West Switchgear Rooms, does not degrade the existing 
fire protection program. The level of protection provided by the 
switchgear room C0<INF>2</INF> fire protection system is enhanced by 
the introduction of the new low pressure system which meets NFPA 12 
(1993) and provides fire suppression for the East or West Switchgear 
Rooms by raising the C0<INF>2</INF> concentration to a 50% level and 
maintains this concentration for a 20 minute duration upon 
initiation. Consequently, the pre-established levels of system 
operability in the event of a fire and the assurance of a safe 
reactor shutdown, as provided by the fire protection systems, have 
not been degraded. An analysis has been performed to ensure that 
either a failure of the low pressure C0<INF>2</INF> storage tank 
outside the switchgear rooms, or a continuous discharge of the 
entire tank contents within the switchgear room, will not adversely 
affect either control room habitability or emergency diesel 
operation. The designation of separate fire areas for the switchgear 
room with equivalent protection does not decrease safety for this 
equipment. As a result, the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Attorney for licensee: R. K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624
    NRC Project Director: Ronald B. Eaton, Acting Project Director

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

    Date of amendment request: July 16, 1997
    Description of amendment request: The proposed amendment would add 
new minimum reactor vessel pressure versus reactor vessel metal 
temperature (P/T) curves, applicable to 12 EFPY (effective full power 
years). These changes are necessary to support leak and hydrostatic 
testing in accordance with the American Society for Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code (Code) Section XI.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed 12 EFPY curve was developed using the same 
methodology as that used in the current 32 EFPY curve and the 8 EFPY 
curve. This methodology is consistent with the guidance provided in 
Regulatory Guide 1.99, Revision 2.
    Assumptions and parameters were the same as those used in the 8 
EFPY curve calculation. However, fluence values used in the 
calculation were those for 12 EFPY.
    Use of the 12 EFPY curves on or before attainment of 12 EFPY of 
operation is equivalent to the previously approved use of the 32 
EFPY curves on or before attainment of 32 EFPY of operation.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change introduces no credible mechanism for 
unacceptable radiation release.
    The proposed change does not require physical modification to 
the plant.
    The 12 EFPY curves are consistent with the previously approved 
32 and 8 EFPY curves.
    Inservice hydrostatic or leak testing is not assumed to be an 
initiator of analyzed events. Since approval of the proposed 
amendment will ensure adequate protection of the reactor pressure 
vessel, it will not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The accident analyses for the plant as described in the FSAR are 
not affected by this proposed change.
    The 12 EFPY curves were developed using the same methodology as 
the 32 and 8 EFPY curves and thus involve no reduction in the margin 
of safety as previously evaluated.
    The margin of safety, relative to the available heat sink in the 
Reactor Coolant System, is actually increased by the use of the 
proposed curves due to the lower allowed test temperature.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

[[Page 52592]]

    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William H. Bateman

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

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

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: August 6, 1997, as supplemented August 
26, 1997
    Brief description of amendments: The proposed amendments would 
address an unreviewed safety question associated with handling of the 
spent fuel shipping cask at the Brunswick Steam Electric Plant, Units 1 
and 2.Date of publication of individual notice in Federal Register: 
September 17, 1997 (62 FR 48897)
    Expiration date of individual notice: October 17, 1997
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

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

    Date of application for amendment: August 26, 1997
    Brief description of amendment request: The proposed amendments 
would approve a modification to the Diablo Canyon Power Plant, Unit 
Nos. 1 and 2 auxiliary saltwater (ASW) system to bypass approximately 
800 feet of Unit 1 and 200 feet of Unit 2 Class 1 ASW pipe, a portion 
of which is buried below sea level in the tidal zone outside the intake 
structure. This modification was completed on Unit 1 during the 
refueling outage completed this year.Date of individual notice in 
Federal Register: September 16, 1997 (62 FR 48677)
    Expiration date of individual notice: October 16, 1997
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

Southern Nuclear Operating Company, Inc., et al., Docket No. 50-
348, Joseph M. Farley Nuclear Plant, Unit No. 1, Houston County, 
Alabama

    Date of amendment request: September 3, 1997
    Description of amendment request: The proposed amendment would 
allow a reduction in the number of required available movable detector 
thimbles (flux map paths) for Cycle 15 operation.Date of publication of 
individual notice in Federal Register: September 10, 1997 (62 FR 47695)
    Expiration date of individual notice: October 10, 1997
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama

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 amendment request: September 17, 1997
    Description of amendment request: The proposed amendments would 
modify Technical Specification 3/4.4.9, ``Specific Activity,'' and 
associated Bases to reduce the limit associated with dose equivalent 
iodine-131.Date of publication of individual notice in Federal 
Register: September 24, 1997 (62 FR 49998)
    Expiration date of individual notice: October 24, 1997
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama

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

    Date of application for amendment: August 14, 1997
    Brief description of amendment: The proposed amendment would change 
the Perry Nuclear Power Plant design basis as described in the Updated 
Safety Analysis Report. The change will add a description of the 
methodology utilized for determining the systems and components that 
are considered to require protection from tornado missiles.Date of 
individual notice in Federal Register: September 16, 1997 (62 FR 
48674).
    Expiration date of individual notice: October 16, 1997
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081

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

[[Page 52593]]

amendment, (2) the amendment, and (3) the Commission's related letter, 
Safety Evaluation and/or Environmental Assessment as indicated. All of 
these items are available for public inspection at the Commission's 
Public Document Room, the Gelman Building, 2120 L Street, NW., 
Washington, DC, and at the local public document rooms for the 
particular facilities involved.

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

    Date of application for amendment: December 27, 1996, as 
supplemented by letter dated August 22, 1997
    Brief description of amendment: The amendments change Technical 
Specification 3/4.6.1.3.b and its associated Bases sections to reflect 
an increase in the peak containment internal pressure for the design 
basis loss-of-coolant accident (LOCA) from 49.5 psig to 52 psig.
    Date of issuance: September 11, 1997
    Effective date: September 11, 1997, to be implemented within 30 
days from its date of issuance.
    Amendment No.: Unit 1 - 113; Unit 2 - 106; Unit 3 - 85
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27794) The August 22, 1997, supplemental letter provided additional 
clarifying information and did not change the staff's original no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated September 11, 1997.No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

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 application for amendments: July 8, 1997, as supplemented 
August 22, 1997
    Brief description of amendments: These amendments remove the 
suppression chamber water volume band from Technical Specification 
3.6.2.1.a.1 while retaining the equivalent water level band. The 
amendments additionally revised the volume band to account for the 
displacement of water due to the installation of larger emergency core 
cooling system suction strainers.
    Date of issuance: September 17, 1997
    Effective date: September 17, 1997
    Amendment Nos.: 188 and 219
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43366) The August 22, 1997, submittal provided a correction to the 
Bases to reflect a change authorized by a previous amendment and did 
not alter the initial no significant hazards determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated September 17, 1997.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

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

    Date of application for amendments: July 1, 1997
    Brief description of amendments: The amendments revise Technical 
Specification definition 1.4, Channel Calibration, to allow an 
alternative method of calibrating thermocouples and resistance 
temperature detector sensors. The amendments also make editorial and 
administrative corrections to TS Table 3.3.2-1, Table 3.3.6-1, and 
Bases Section 3/4.3.1.
    Date of issuance: September 15, 1997
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 102 and 104
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40848) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 15, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of application of amendments: October 30, 1996, as 
supplemented by letters dated April 22, July 2, September 3, and 
September 4, 1997
    Brief description of amendments: The amendments revise the Reactor 
Building Structural Integrity Technical Specifications regarding the 
tendon surveillance program.
    Date of Issuance: September 15, 1997
    Effective date: The license amendments are effective as of the date 
of issuance and the change to the facilities shall be implemented prior 
to the Unit 1 end-of-cycle 17 outage. Implementation of the amendments 
shall include the provisions that the licensee provide in the facility 
Updated Final Safety Analysis Report (specifically the Selected 
Licensee Commitment Manual) the prescribed lower limit and the minimum 
required value of Reactor Building Post-Tensioning System tendon forces 
for each group of tendons prior to performing the seventh tendon 
surveillance for Unit 1. In addition, the portion of the Selected 
Licensee Commitment Manual related to the establishment of these limits 
will be submitted as soon as available.
    Amendment Nos.: 225, 225, 222
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications, License Conditions, 
and Appendix C.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64383) The April 22, July 2, September 3, and September 4, 1997, 
letters provided clarifying information that did not change the scope 
of the October 30, 1996, application and the initial proposed no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated September 15, 1997.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of application of amendments: June 12, 1997
    Brief description of amendments: The amendments change the name 
``Duke Power Company'' to ``Duke Energy Corporation'' in the Oconee 
facility operating licenses and Technical Specifications.
    Date of Issuance: September 16, 1997

[[Page 52594]]

    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 226, 226, 223
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications and Operating Licenses 
including Appendix C.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35849) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 16, 1997, and 
Environmental Assessment dated August 21, 1997 (62 FR 44495).No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

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

    Date of application for amendments: September 6, 1996, as 
supplemented May 23, 1997 and August 13, 1997.
    Brief description of amendments: These amendments revise Item 7.c 
of Beaver Valley Power Station, Unit No. 1 (BVPS-1) Technical 
Specification (TS) Table 3.3-3 and Item 7.d of Beaver Valley Power 
Station, Unit No. 2 (BVPS-2) TS Table 3.3-3 to reflect that a safety 
injection (SI) signal starts all auxiliary feedwater (AFW) pumps. The 
notation on BVPS-1 TS Table 3.3-5 is revised to state that the response 
time is for all AFW pumps on all SI signal starts. Items 7.d of BVPS-2 
TS Tables 3.3-4 and 4.3-2 is revised to reflect that an SI signal 
starts all AFW pumps.
    The amendments also revise and reformat TSs 3/4.7.1.2 to more 
closely resemble the wording contained in the NRC's ``Standard 
Technical Specifications Westinghouse Plant,'' (NUREG-1431, Revision 
1). These changes require three AFW trains to be operable and describe 
what constitutes an operable train. The mode applicability for these 
TSs is expanded to include Mode 4 when the steam generator(s) is relied 
upon for heat removal.
    Date of issuance: September 18, 1997
    Effective date: Both units, as of the date of issuance, to be 
implemented within 60 days
    Amendment Nos.: 206 and 85
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1996 (61 
FR 58902) The May 23, 1997, and August 13, 1997, letters provided minor 
editorial changes that did not change the initial proposed no 
significant hazards consideration determination or expand the amendment 
request beyond the scope of the November 19, 1996, Federal Register 
notice. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 18, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

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

    Date of application for amendment: May 29, 1997
    Brief description of amendment: The amendments consist of changes 
to the Technical Specifications (TS) which correct typographical 
errors, remove outdated material, incorporate minor changes in text, 
make editorial corrections, and resolve other inconsistencies in the 
Unit 1 and 2 TS.
    Date of Issuance: September 22, 1997
    Effective Date: September 22, 1997
    Amendment Nos.: 152 and 89
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40849) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 22, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596

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

    Date of amendment request: July 8, 1996
    Brief description of amendments: The amendments allowed that the 
component cooling water system surge tank level instrumentation can be 
demonstrated operable, by performing a channel calibration test, during 
any plant mode of operation.Date of issuance: September 23, 1997
    Effective date: September 23, 1997, to be implemented within 30 
days of issuance.
    Amendment Nos.: Unit 1 - Amendment No. 91; Unit 2 - Amendment No. 
78
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44358) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 23, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

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

    Date of application for amendment: June 19, 1997
    Brief description of amendment: Technical Specification 3/4.7.1.3 
requires sufficient water to be available for the auxiliary feedwater 
system to maintain the reactor coolant system at hot standby for 10 
hours before cooling down to hot shutdown in the next 6 hours. The 
amendment increases the required volume of water when the demineralizer 
water storage tank and condensate storage tank are being credited, 
makes editorial changes, and expands the descriptions in Bases Sections 
3/4.7.1.2 and 3/4.7.1.3.
    Date of issuance: September 11, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 150
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40853) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 11, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

[[Page 52595]]

Northern States Power Company, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County, Minnesota

    Date of application for amendments:  May 7, 1997, as supplemented 
May 30, July 29, and September 12, 1997
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.8, including TS 3.8.D.1 and TS 3.8.D.3, to change 
TS limitations on crane operations in the spent fuel pool enclosure 
relating to spent fuel pool special ventilation system operability. 
These changes are necessary to allow movement of loads over spent fuel 
stored in the spent fuel pool enclosure with the spent fuel pool 
special ventilation system inoperable. The staff denied the proposed 
change to TS 3.8.D.2. A separate notice of denial has been sent to the 
Federal Register for publication.
    Date of issuance: September 15, 1997
    Effective date: September 15, 1997, with full implementation within 
30 days. License Condition 4 of Appendix B is effective immediately 
upon issuance of the amendments.
    Amendment Nos.: 130 and 122
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35850) The July 29 and September 12, 1997, letters provided clarifying 
information within the scope of the original application and did not 
change the staff's initial proposed no significant hazards 
considerations determination. The Commission's related evaluation of 
the amendments is contained in a Safety Evaluation dated September 15, 
1997.No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: July 3, 1997
    Brief description of amendment: This amendment makes changes to 
Technical Specification Table 3.6.3-1, ``Primary Containment Isolation 
Valves'' to add valves to the list, therein.
    Date of issuance: September 15, 1997
    Effective date: Effective as of the date of issuance, to be 
implemented within 60 days.
    Amendment No.: 102
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43375) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 15, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: July 7, 1997
    Brief description of amendment: The amendment changes Technical 
Specification (TS) 3/4.8.4.2, ``Motor Operated Valves - Thermal 
Overload Protection (BYPASSED),'' to relocate the list of applicable 
valves (TS Table 3.8.4.2-1) to the Hope Creek Generating Station 
Updated Final Safety Analysis Report.
    Date of issuance: September 16, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 103
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications and the License.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43375) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 16, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: April 1, 1997, as supplemented 
by letter dated May 30, 1997
    Brief description of amendment: The amendment changed Technical 
Specifications (TSs) 4.6.1.1, ``Primary Containment Integrity;'' 3/
4.6.1.2, ``Primary Containment Leakage;'' 3/4.6.1.3, ``Primary 
Containment Air Locks;'' 4.6.1.5.1, ``Primary Containment Structural 
Integrity;'' and 4.6.1.8.2, ``Drywell and Suppression Chamber Purge 
System.'' This amendment also changed the Bases for 3/4.6.1.2, 
``Primary Containment Leakage;'' 3/4.6.1.3, ``Primary Containment Air 
Locks;'' 3.4.6.1.5, ``Primary Containment Structural Integrity;'' 
Section 6, ``Administrative Controls;'' and License Condition 2.D of 
Facility Operating License NPF-57. A new TS, 6.8.4.f, ``Primary 
Containment Leakage Rate Testing Program,'' was added. These changes 
modify the TSs and the Facility Operating License to adopt the 
performance based containment leak rate testing requirements (Option B) 
of 10 CFR Part 50, Appendix J.Date of issuance: September 18, 1997
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 104
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications and the License.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43375) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 18, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

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

    Date of application for amendment: March 26, 1997
    Brief description of amendment: The amendment changes the 
definition of ``Core Alteration.''
    Date of issuance: September 17, 1997
    Effective date: September 17, 1997
    Amendment No.: 138
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27800) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 17, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

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

    Date of application for amendments: September 26, 1996, as 
supplemented on August 12, 1997 (TS 96-04)
    Brief description of amendments: The amendments change the 
Technical

[[Page 52596]]

Specifications (TS) by relocating the fire protection program details 
to the Updated Final Safety Analysis Report and Fire Protection Plan in 
accordance with Generic Letters 86-10 and 88-12.
    Date of issuance: September 23, 1997
    Effective date: September 23, 1997
    Amendment Nos.: 228 and 219
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35843)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 23, 1997.No significant hazards 
consideration comments received: No.
    Local Public Document Room location:  Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of application for amendment: April 30, 1997, as supplemented 
June 18, July 21 (3 letters), August 7 and 21, 1997
    Brief description of amendment: The proposed amendment would change 
the design features section of the Technical Specifications (TS) to 
provide for insertion of Lead Test Assemblies containing Tritium 
Producing Burnable Absorber Rods in the Watts Bar Nuclear Plant reactor 
core during Cycle 2.
    Date of issuance: September 15, 1997
    Effective date: September 15, 1997
    Amendment No.: 8
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30644) The TVA letters dated June 18, July 21, August 7 and 21, 1997 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in an Environmental Assessment dated September 8, 1997, and in a Safety 
Evaluation dated September 15, 1997.No significant hazards 
consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402

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

    Date of application for amendment: January 16, 1996, supplemented 
December 6, 1996, and August 15, 1997
    Brief description of amendment: The amendment extended the test 
interval for the drywell bypass leakage rate test from 18 months to 10 
years. Also, some surveillances for the drywell air locks were 
increased from 18 months to 24 months.
    Date of issuance: September 22, 1997
    Effective date: September 22, 1997
    Amendment No.: 88
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1996 (61 FR 
3951) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 22, 1997.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment: May 2, 1997
    Brief description of amendment: The amendment revises an existing 
exception to Limiting Condition for Operation (LCO) 3.0.4 as it applies 
to LCO 3.6.1.9 for the main steam isolation valve (MSIV) leakage 
control system (LCS) by making the exception permanent and clarifying 
that it only applies for the inboard MSIV LCS subsystem.
    Date of issuance: September 24, 1997
    Effective date: September 24, 1997
    Amendment No.: 89
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33135) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 24, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment: August 14, 1997
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.6 by adding a note that would extend the 
surveillance interval to perform the inservive testing (IST) full 
stroke exercise of primary containment isolation check valve TIP-V-6 
until the 1998 refueling outage, scheduled to begin no later than May 
15, 1998, or until a plant shutdown of sufficient duration occurs to 
allow TIP-V-6 testing, whichever occurs first.
    Date of Issuance: September 18, 1997
    Effective date: September 18, 1997, to be implemented within 30 
days of issuance.
    Amendment No.: 152
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: Yes (62 FR 45280 dated August 26, 
1997). 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 September 25, 1997, but 
indicated 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 and 
final no significant hazards consideration determination are contained 
in a Safety Evaluation dated September 18, 1997.
    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

    Date of application for amendments: January 16, 1997 (TSCR-191), as 
supplemented on April 17, August 7, and August 27, 1997

[[Page 52597]]

    Brief description of amendments: These amendments increase the 
minimum volume and boron concentration for the refueling water storage 
tanks and the boric acid storage tanks. Additionally, these amendments 
increase the minimum concentration of boric acid in the safety 
injection accumulator, the reactor coolant system during refueling 
operations, and the reactor coolant system during positive reactivity 
changes made when containment integrity is not maintained.
    Date of issuance: September 23, 1997
    Effective date: September 23, 1997, with full implementation within 
45 days
    Amendment Nos.: 180 and 184
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19836) The April 17, August 7, and August 27, 1997, submittals provided 
clarifying information within the scope of the original application and 
did not change the staff's initial proposed no significant hazards 
considerations determination. The Commission's related evaluation of 
the amendments is contained in a Safety Evaluation dated September 23, 
1997.No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241

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

    Date of amendment request: July 3, 1997
    Brief description of amendment: The amendment changes the 
definition for an alteration of the reactor core to one that is 
consistent with the intent of the Improved Standard Technical 
Specifications.
    Date of issuance: September 18, 1997
    Effective date: September 18, 1997
    Amendment No.: 109
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40861) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 18, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

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

    Date of amendment request: July 3, 1997
    Brief description of amendment: The amendment modifies Technical 
Specifications 5.3.1, ``Fuel Assemblies'' and 6.1.9.6, ``CORE OPERATING 
LIMITS REPORT (COLR)'' to add ZIRLO as fuel material and the use of 
limited zirconium alloy filler rods in place of fuel rods.
    Date of issuance: September 22, 1997
    Effective date: September 22, 1997, to be implemented within 30 
days of issuance.
    Amendment No.: 110
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40860) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 22, 1997.No 
significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental

[[Page 52598]]

assessment need be prepared for these amendments. If the Commission has 
prepared an environmental assessment under the special circumstances 
provision in 10 CFR 51.12(b) and has made a determination based on that 
assessment, it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By November 7, 1997, 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 contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendment: September 15, 1997, as 
supplemented by letter dated September 16, 1997
    Brief description of amendment: The amendment revised the 
applicability requirement in Technical Specifications (TSs) Sections 
3.4.2, ``Safety/Relief Valves'' (Action c), 4.4.2, and 3.3.7.5, 
``Accident Monitoring Instrumentation'' (TS Table 3.3.7.5-1, Action 
80). The change to the referenced TSs adds the following applicability 
footnote: Compliance with these requirements for the ``S'' SRV acoustic 
monitor is not required for the period beginning September 12, 1997, 
until the next unit shutdown of sufficient duration to allow for 
containment entry, not to exceed the 10th refueling and inspection 
outage.
    Date of issuance: September 23, 1997
    Effective date: September 23, 1997
    Amendment No.: 169
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No. On September 17, 1997, the staff 
issued a Notice of Enforcement Discretion, which was immediately 
effective and remained in effect until this amendment was issued.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, consultation with the State of Pennsylvania, 
and final no significant hazards consideration determination are 
contained in a Safety Evaluation dated September 23, 1997.

[[Page 52599]]

    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    NRC Project Director: John F. Stolz
    Dated at Rockville, Maryland, this 1st day of October 1997.
    For the Nuclear Regulatory Commission
John N. Hannon,
Acting Director, Division of Reactor Projects - III/IV, Office of 
Nuclear Reactor Regulation
[Doc. 97-26502 Filed 10-7-97; 8:45 am]
BILLING CODE 7590-01-F