[Federal Register Volume 62, Number 78 (Wednesday, April 23, 1997)]
[Notices]
[Pages 19825-19845]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10423]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 29, 1997, through April 11, 1997. The 
last biweekly notice was published on April 9, 1997 (62 FR 17223).

Notice of Consideration Of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, and Opportunith 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 
Review 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 May 23, 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

[[Page 19826]]

the applicant on a material issue of law or fact. Contentions shall be 
limited to matters within the scope of the amendment under 
consideration. The contention must be one which, if proven, would 
entitle the petitioner to relief. A petitioner who fails to file such a 
supplement which satisfies these requirements with respect to at least 
one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of amendment request: March 17, 1997
    Description of amendment request: The proposed change would revise 
eight specifications for 18-month tests to delete a conditional 
statement that the testing be done while the unit is shut down and to 
clarify that Harris Nuclear Plant (HNP) may take credit for tests on 
some components which are performed while the unit is at power.
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes permit HNP to evaluate the conditions 
required to safely perform a test, but the changes do not directly 
affect the functioning or operation of any plant equipment. Since no 
equipment operation is involved there is no increase in the 
probability or consequence of any previously identified accident.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the conditional statements on the 
surveillance frequencies do not involve any physical alterations or 
additions to plant equipment or alter the manner in which any 
safety-related system performs itsfunction or is operated. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes to the conditional statements on the 
surveillance frequency allows HNP to evaluate the conditions needed 
to safely perform the required testing. There is no change in the 
frequency of testing or in the testing which is required. There is 
no change in the responsibility of HNP toperform tests in a safe and 
responsible manner, and any changes to procedures will have to be 
individually evaluated to ensure that they do not reduce 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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Mark Reinhart, Acting

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

    Date of amendment request: January 30, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 1.0, ``Definitions;'' TS 3/4.6.1, 
``Primary Containment'' and associated Bases; and TS 5.4.2, ``Reactor 
Coolant System Volume'' for Byron and Braidwood to support steam 
generator replacement. ComEd will be replacing the original 
Westinghouse D4 steam generators at Byron and Braidwood with Babcock 
and Wilcox International steam generators. The replacement steam 
generators increase the Reactor Coolant System volume which results in 
a higher calculated peak containment pressure (Pa) value.
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Each of the [replacement steam generators] RSGs has a larger 
[reactor coolant system] RCS side volume than the original steam 
generators (OSGs). As a result of the RCS

[[Page 19827]]

volume increase, the mass and energy release during the blowdown 
phase of the large break loss of coolant accident (LBLOCA) is 
increased. Additionally, the heat transfer rate of the RSGs is 
greater than the OSGs, and the RSGs will operate at a slightly 
higher pressure than that for the OSGs. Consequently, the steam 
enthalphy exiting the break during the reflood period, with the RSG, 
will be greater than that for the OSG. This results in an increase 
in the containment building peak pressure, Pa.
    The proposed revisions to the Technical Specifications involve 
the specified value of Unit 1 RCS volume and the defined value of 
Unit 1 Pa. Several editorial changes are also being made 
to improve clarity and consistency of the TS.
    RCS volume is not an initiator for any event and an increase in 
volume does not affect any operating margin or requirements. 
Therefore, increasing the primary volume does not increase the 
probability of any event previously analyzed.
    The revised value of Pa continues to be less than the 
design basis pressure for the containment building structure. The 
change represents only a revision to the containment test pressure 
for containment leakage testing. Such testing is only performed with 
the affected unit in the shutdown condition. Therefore, the proposed 
change in Pa does not involve a significant increase in 
the probability of an accident previously evaluated.
    All accidents in the Updated Final Safety Analysis Report 
(UFSAR) were evaluated to determine the effect of an increase in 
primary volume on accident consequences. The events identified that 
may be impacted by an increase in primary volume are the Waste Gas 
System Leak or Failure and LBLOCA. For the Waste Gas System Leak or 
Failure, the activity of the decay tank is controlled to Technical 
Specification limits which are unaffected by RCS volume. Therefore, 
an increase in RCS volume would not increase the offsite dose.
    The offsite dose calculation for the LBLOCA is unaffected by the 
proposed change. The license basis offsite dose calculation is in 
accordance with NRC Reg Guide 1.4 ``Assumptions Used for Evaluating 
The Potential Radiological Consequences of a Loss of Coolant 
Accident for Pressurized Water Reactors.'' This Regulatory Guide 
states, in part, ''...a number of appropriately conservation 
assumptions, based on engineering judgment and on applicable 
experimental results from safety research programs conducted by the 
AEC.'' These conservatisms include (but are not limited to) the 
following assumptions:
     Twenty five percent of the equilibrium radioactive full 
power inventory is immediately available for leakage from the 
primary containment.
     100% of the equilibrium full power radioactive noble 
gas inventory is immediately available for leakage from the primary 
containment.
     The primary containment should be assumed to leak at 
the (maximum) leak rate specified in the technical specifications 
for the first 24 hours and at 50% of this value for the remaining 29 
days of the accident duration.
    The design basis leakage corresponding to a peak containment 
pressure of 50 psig utilized in the design basis accident analysis 
is 0.10% per day of the containment free air mass. Therefore, the 
offsite dose calculation was performed with a leakage of .1% per day 
for day one and .05% per day for days two through 30. Isotopic 
inventories are unaffected by the increase in reactor coolant 
volume. Thus, the offsite dose is unaffected by the increase in the 
peak containment pressure. Therefore, this proposed change to 
Pa does not involve a significant increase in the 
consequences of an accident previously evaluated.
    The editorial changes proposed are for clarity and consistency 
within the Technical Specifications and do not affect either 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 in RCS volume is a change in a plant 
parameter within the ``Design Features'' section of the Technical 
Specifications. Increasing the RCS volume does not create any new or 
different failure modes. The existing RCS design requirements 
continue to be met.
    The revised value of Pa continues to be less than the 
design basis pressure for the containment building structure. The 
change represents only a revision to the test pressure for 
containment leakage testing. Such testing is only performed with the 
affected unit in the shutdown condition. Therefore, no new or 
different failure modes are being introduced by modification of the 
testing parameters.
    The editorial changes proposed are for clarity and consistency 
within the Technical Specifications and do not result in any 
physical changes to the facility or how it is operated. No new or 
different failure modes are being introduced by these changes.
    Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Changing the RCS volume in the Technical Specifications does not 
reduce the margin of safety. RCS volume is a design feature. The 
change in RCS volume does not involve a change to any setpoint or 
design requirements. An evaluation of all UFSAR accidents was 
performed to determine the effect of an increase in RCS volume. This 
evaluation is summarized as follows:
    An evaluation of the Chemical and Volume Control System 
Malfunction was performed to determine the effect of the increased 
RCS volume due to the RSGs. The larger RCS volume of the RSGs 
reduces the reactivity insertion for a given dilution flow rate. 
Therefore, the UFSAR analyses remain bounding for Byron Unit 1 and 
Braidwood Unit 1 with the RSGs and there is no reduction in the 
margin of safety.
    An evaluation of the Inadvertent Actuation of the Emergency Core 
Cooling System During Power Operation Event was performed to 
determine the effect of the increased RCS volume due to the RSGs. 
For this event, the injection of borated water causes a negative 
reactivity insertion, which increases DNBR. For a given Refueling 
Water Storage Tank (RWST) boron concentration, the larger RCS volume 
will cause a reduction in the negativity insertion rate as compared 
to the current UFSAR analysis. However, negative reactivity would 
still be inserted, no fuel pins would experience DNB, and there is 
no reduction in the margin of safety.
    An evaluation of the Small Break LOCA was performed to determine 
the effect of increased RCS volume. The additional RCS volume will 
cause a delay in the loop seal clearing which in turn delays the 
core uncovery as compared with the UFSAR analysis. A delay in core 
uncovery reduces the amount of core heatup which results in a lower 
peak clad temperature (PCT) because the core decay heat would be 
less than in the UFSAR analysis. The benefit is considered small, 
but there is still a benefit. Therefore, the increased RCS volume 
does not result in a reduction in the margin of safety.
    An evaluation of the Large Break LOCA was performed to determine 
the effect of increased RCS volume. For a LBLOCA, the increased RCS 
volume causes the blowdown phase of the event to be longer. 
Increased blowdown phase, alone, could potentially result in a 
higher PCT. However, the RSGs also have less resistance to flow due 
to increased primary side steam generator flow area, which results 
in a higher blowdown flow compared to the OSGs. The increased 
blowdown flow more than compensates for the longer blowdown phase 
associated with the increased RCS volume. The net effect is a 
decrease in PCT for the RSG compared to the OSG. Therefore, there is 
no reduction in the margin of safety.
    An evaluation of the Gas Waste System Leak or Failure was 
performed to determine the effect of the increased RCS volume. 
Because the activity of the decay tank is controlled within 
Technical Specification limits, an increase in RCS volume would not 
change the results of the event. Therefore, there is no reduction in 
the margin of safety.
    An evaluation was performed to determine the effect of the 
increased RCS volume on the peak containment pressure following a 
LBLOCA. The increased RCS volume caused the peak containment 
pressure to increase to 47.8 psig. This is still below the 
containment design pressure of 50.0 psig. Therefore, there is no 
reduction in the margin of safety.
    This proposed change involves testing requirements designed to 
demonstrate adequate leakage rates are maintained. If adequate 
leakage rates are maintained as outlined in the Technical 
Specifications, there will be no reduction in the margin of safety. 
In the event of degradation of a containment seal that results in 
unacceptable leakage, plant shutdown will occur as required by 
Technical Specifications and administrative requirements in 
accordance with approved plant procedures. Therefore, this proposed 
change does not involve a significant reduction in a margin of 
safety.
    The editorial changes proposed are for clarity and consistency 
within the Technical Specifications and do not result in any 
physical changes to the facility or how it is

[[Page 19828]]

operated. Therefore, the changes have no effect on the margin of 
safety.
    Thus, this amendment request does not result in any decrease 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: March 27, 1997
    Description of amendment request: The proposed amendment would 
alter the company name in the Facility Operating License DPR-20 and 
Technical Specifications for the Palisades Plant. Specifically, the 
proposed amendment would revise the name from ``Consumers Power 
Company'' to ``Consumers Energy Company.''
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Since the proposed changes do not alter the technical content of 
any Facility Operating License or Technical Specifications 
requirements, they do not alter any feature of plant equipment, 
settings, operation, or configuration.
    Therefore, they cannot involve a significant increase in the 
probability of an accident previously evaluated.
    The proposed changes alter the company name in the Facility 
Operating License and Technical Specifications to reflect the change 
from ``Consumers Power Company'' to ``Consumers Energy Company''. 
The proposed change will not affect any obligations. The company 
will continue to own all of the same assets, will continue to serve 
the same customers, and will continue to honor all existing 
obligations and commitments. The proposed changes will not alter 
plant operation or configuration, or its ability to respond to 
accidents.
    Therefore, they will not involve a significant increase in the 
consequences of any accident previously evaluated.
    B. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    Since the proposed changes do not alter the technical content of 
any Facility Operating License or Technical Specifications 
requirements, they do not alter any feature of plant equipment, 
settings, operation or configuration.
    Therefore, they cannot create the possibility of a new or 
different kind of accident from any previously evaluated.
    C. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Since the proposed changes do not alter the technical content of 
any Facility Operating License or Technical Specifications 
requirements, they do not alter any feature of plant equipment, 
settings, operation, or configuration.
    Therefore, they cannot 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: Van Wylen Library, Hope 
College, Holland, Michigan 49423
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: John N. Hannon

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

    Date of amendment request: March 31, 1997 (TSC 96-10)
    Description of amendment request: The proposed amendments would 
modify and clarify the High Pressure Injection (HPI) System operability 
requirements in Specification 3.3.1, impose additional HPI system 
operability requirements for operation above 75 percent power, 
incorporate the new Standard Technical Specifications format for the 
HPI system, revise Specification 3.3.2 to clarify that the Reactor 
Building Emergency Sump isolation valves are remote-manually operated 
valves, and add new specifications and a surveillance test to address 
operability requirements of the atmospheric dump valves. In addition, 
corresponding Bases changes would be incorporated.
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. None of the proposed changes has any impact upon the 
probability of any accident which has been evaluated in the UFSAR 
[Updated Final Safety Analysis Report]. The only potential change in 
operating configuration is allowing operation with the HPI [High 
Pressure Injection] System pump discharge header cross-
    connected. This operating mode does not affect the probability 
of a LOCA [Loss-of-Coolant Accident] or of any other accident 
evaluated in the UFSAR.
    None of these changes have any impact upon the ability of the 
HPI System to add soluble poison to the Reactor Coolant System, as 
addressed by Specification 3.2. The remaining potential impact is 
upon the ability to mitigate the consequences of a small break LOCA, 
which is addressed below. The small break LOCA is the limiting 
design basis accident with respect to HPI System operability 
requirements.
    The proposed changes to Specification 3.3.1 provide appropriate 
actions to address any degradation in the operability of the HPI 
System. The operability requirements for the HPI System are 
supported by a spectrum of small break LOCA analyses based on the 
approved Evaluation Model described in FTI [Framatome Technologies, 
Incorporated] topical report BAW-10192P. These small break LOCA 
analyses demonstrate that the acceptance criteria of 10CFR 50.46 are 
not violated.
    Two trains of HPI are required to mitigate a small break LOCA 
above 75% FP [full power]. Operability requirements in the proposed 
Technical Specifications assure that the HPI System can withstand 
the worst single failure and still result in two HPI pumps injecting 
through two trains. The full power small break LOCA analyses 
supporting this proposed license amendment have been performed in 
accordance with the approved Evaluation Model described in FTI 
topical report BAW-10192P. The proposed Technical Specifications 
limit operation above 75% FP with a degraded HPI System to 72 hours 
before a power reduction to less than 75% FP (or a reactor shutdown) 
must be initiated. The required actions depend on the HPI System 
components that are inoperable. The 72 hour completion time is 
consistent with the time requirements for HPI specified in NUREG-
1430.
    When at or below 75% FP, one HPI train provides sufficient flow 
to mitigate a small break LOCA. The 75% power level is justified by 
analyses using the Evaluation Model described in FTI topical report 
BAW-10192P, considering the worst case break location and size 
described in LER [Licensee Event Report] 269/90-15 and Attachment 3 
to this submittal. The proposed Technical Specifications require two 
HPI trains to be operable at or below 75% FP. These requirements 
ensure that, following the worst single failure, one train of HPI 
would remain

[[Page 19829]]

available to mitigate a small break LOCA. Operation with less than 
two HPI trains operable is restricted to 72 hours before shutdown 
requirements are imposed. This completion time is consistent with 
the time requirements specified for an HPI System in NUREG-1430.
    The additional HPI system restriction that requires the HPI pump 
discharge header to be cross-connected when all three HPI pumps are 
operable does not increase the consequences of a small break LOCA. 
If a single failure prevents one HPI train from actuating, this 
lineup results in at least two HPI pumps initially injecting through 
the automatically actuating train. This increases the amount of 
cooling flow initially delivered to the core as compared to the 
current system configuration.
    The impact of this alignment has been evaluated, considering the 
potential single active failures, including the failure of any 
powered component to operate and any single failure of electrical 
equipment.
    It has been determined that, when each of the three HPI pumps is 
either running or is capable of automatic actuation upon an 
Engineered Safeguards signal, cross-connection of the HPI pump 
discharge header does not introduce susceptibility to any single 
failure. Therefore, the potential consequences of a small break LOCA 
are not increased. If fewer than three HPI pumps are either running 
or are capable of automatic actuation, and the HPI pump discharge 
header were cross-connected, a single failure of one pump could 
cause a single pump to be aligned to both HPI trains. In this 
condition, the single pump could experience runout conditions prior 
to corrective operator action. However, proposed Specification 3.3.1 
requires the discharge header to be isolated between the two 
remaining operable HPI pumps. The proposed BASES provide guidelines 
to ensure that the requirements for redundancy are properly 
implemented. Therefore, the proposed specifications ensure that the 
consequences of a small break LOCA are not increased by allowing the 
HPI pump discharge header to be cross-connected.
    In addition, proposed Specification 3.4.7 requires new 
operability requirements for the main steam atmospheric dump valves. 
These operability requirements do not impact the probability or 
consequences of any accident. The proposed specification for the 
atmospheric dump valves provides additional assurance that these 
valves will be operable in the event of a small break LOCA.
    In summary, the proposed Technical Specifications provide 
adequate controls to assure that operability of the HPI System is 
maintained in a manner consistent with the requirements of the 
design basis accidents. Therefore, it is concluded that this 
amendment request will 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 kind of accident previously evaluated:
    No. Of the proposed substantive changes, only cross-connection 
of the HPI pump discharge header represents any change to the way in 
which the facility is normally operated. Operation with the 
discharge header cross-connected is not a new configuration, as it 
has always been used for HPI pump testing both at power and during 
shutdown conditions. Potential failure modes have already been 
considered as described earlier. No new initiating events or 
potentially unanalyzed conditions have been created. Therefore, this 
proposed amendment will not create the possibility of any new or 
different kind of accident.
    (3) Involve a significant reduction in a margin of safety.
    No. The HPI restrictions associated with the proposed Technical 
Specifications are supported by analyses which demonstrate that the 
acceptance criteria of 10 CFR 50.46 are not violated for any small 
break LOCA. These analyses were performed in accordance with the 
Evaluation Model described in FTI topical report BAW-10192P. 
Therefore, it is concluded that the proposed amendment request will 
not result in a significant decrease in the margin of safety.
    Duke has concluded, based on the above, that there are no 
significant hazards considerations involved in this amendment 
request.
    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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: March 10, 1997
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.4.5, ``Steam Generators,'' and 
associated Bases to allow repair of steam generator tubes by 
installation of sleeves with the tungsten inert gas (TIG) welded sleeve 
developed by ABB Combustion Engineering. In addition, the proposed 
amendment would delete the option for using the kinetic sleeving 
methodology previously approved for use at Beaver Valley, but is not 
currently recommended by Framatome Technologies, Inc.
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment allows the ABB Combustion Engineering 
(ABB/CE) tungsten inert gas (TIG) welded tubesheet sleeves and tube 
support plate sleeves to be used as an alternate steam generator 
tube repair method. The sleeve configuration was designed and 
analyzed in accordance with the criteria of Regulatory Guide (RG) 
1.121 and Section III of the ASME [American Society of Mechanical 
Engineers] Code. Fatigue and stress analyses of the sleeved tube 
assemblies produce acceptable results for both types of sleeves as 
documented in ABB/CE Topical Report CEN-629-P, Revision 02 and CEN-
629-P Addendum 1. Mechanical testing has shown that the structural 
strength of the sleeves under normal, faulted, and upset conditions 
is within the acceptable limits specified in RG 1.121. Leakage rate 
testing for the tube sleeves has demonstrated that primary to 
secondary leakage is not expected during any plant condition. The 
consequences of leakage through the sleeved region of the tube is 
fully bounded by the existing steam generator tube rupture (SGTR) 
analysis included in the Updated Final Safety Analysis Report 
(UFSAR).
    The sleeves are designed to allow inservice inspection of the 
pressure retaining portions of the sleeve and parent tube. Inservice 
inspection is performed on all sleeves following installation to 
ensure that each sleeve has been properly installed and is 
structurally sound. Periodic inspections are performed in subsequent 
refueling outages to monitor sleeve degradation on a sample basis. 
The eddy current technique used for inspection will be capable of 
detecting both axial and circumferential flaws. Specific guidance 
for steam generator sleeve inspection is provided in the current 
technical specification surveillance requirements. Tubes that 
contain defects in a sleeve, which exceed the repair limit, will be 
removed from service. This ensures that sleeve and tube structural 
integrity is maintained.
    The proposed TS change to support the installation of TIG welded 
sleeves does not adversely impact any previously evaluated design 
basis accident. The effect of sleeve installation on the performance 
of the SG [steam generator] was analyzed for heat transfer, flow 
restriction, and steam generation capacity. The sleeves reduce the 
risk of primary to secondary leakage in the SG. The installation of 
ABB/CE sleeves results in a hydraulic flow restriction that is 
dependent on the number and types of sleeves installed. The 
reduction in primary system flow rate is a small percentage of the 
flow rate reduction seen from plugging one tube and is a preferable 
alternative when considering core margins based on minimum reactor 
coolant system flow rates. The sleeving installation will result in 
a resistance to primary coolant flow through the tube for other 
evaluated accidents. The results of the analyses and testing, as 
well as industry operating experience, demonstrate that the sleeve 
assembly is an acceptable

[[Page 19830]]

means of maintaining tube integrity. In summary, installation of 
sleeves does not substantially affect the primary system flow rate 
or the heat transfer capability of the steam generators.
    Installation of the sleeves can be used to repair degraded tubes 
by returning the condition of the tubes to their original design 
basis condition for tube integrity and leak tightness during all 
plant conditions. The tube bundle overall structural and leakage 
integrity will be increased with the installation of the sleeves 
reducing the risk of primary to secondary leakage in the SG while 
maintaining acceptable reactor coolant system flow rates. Therefore, 
sleeving will not increase the probability of occurrence of an 
accident previously evaluated.
    Removal of the kinetically welded sleeve process as an approved 
SG tube repair methodology will have no effect on plant operations. 
There are currently no kinetically welded sleeves installed in the 
steam generators. Had there been, plant operations would have still 
been bounded by the existing SGTR analysis in the UFSAR.
    Therefore, these proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The implementation of the proposed sleeving process will not 
introduce significant or adverse changes to the plant design basis. 
Stress and fatigue analyses of the repair has shown the ASME Code 
Section III and RG 1.121 allowable values are met. Implementation of 
TIG welded sleeving maintains overall tube bundle structural and 
leakage integrity at a level consistent with that of the originally 
supplied tubing. Leak and mechanical testing of the sleeves support 
the conclusions that the sleeve retains both structural and leakage 
integrity during all conditions. Repair of a tube with a sleeve does 
not provide a mechanism that would result in an accident outside of 
the area affected by the sleeve.
    Any hypothetical accident as a result of potential tube or 
sleeve degradation in the repaired portion of the tube is bounded by 
the existing SGTR analysis. The SGTR analysis accounts for the 
installation of sleeves and the impact on current plugging level 
analyses. The sleeve design does not affect any other component or 
location of the tube outside of the immediate area repaired.
    The current primary to secondary leakage limit ensures that SG 
tube integrity is maintained in the event of an MSLB [main steam 
line break] or LOCA [loss-of-coolant accident]. The limit will 
provide for leakage detection and a plant shutdown in the event of 
the occurrence of an unexpected single crack resulting in excessive 
tube leakage. The leakage limit also provides for early detection 
and a plant shutdown prior to a postulated crack reaching critical 
crack lengths for MSLB conditions.
    Inservice inspections are performed following sleeve 
installation to ensure proper weld fusion has occurred to maintain 
structural integrity. The post installation inspection also serves 
as baseline data to be used for comparison during future 
inspections. Periodic eddy current inspections monitor the pressure 
retaining portions of the sleeve and parent tube for degradation. 
Eddy current techniques will be employed that are sensitive to axial 
and circumferential degradation.
    Increasing the sample size of tubes repaired using either 
sleeving process during each scheduled inservice inspection will 
increase the monitoring of these tubes for any further degradation. 
The improved monitoring and evaluation of the tube and the sleeves 
assures tube structural integrity is maintained or the tube is 
removed from service.
    Corrosion testing of typical sleeve-tube configurations was 
performed to evaluate local stresses, sleeve life, and resistance to 
primary and secondary side corrosion. The tests were performed on 
stress relieved and as-welded (non-stress relieved) sleeve-tube 
joints. Using the corrosion test data in conjunction with finite 
element analyses of the local stress, the stress relieved joint life 
was determined to be in excess of 40 years. The ABB/CE TIG welded 
sleeve operating experience in the industry has shown no sleeve 
failures due to service induced degradation in sleeves that were 
installed with acceptable inspection results. This experience 
includes the stress relieved and as-welded sleeve configurations. 
All sleeves will be stress relieved as specified in the topical 
report.
    Removal of the kinetically welded sleeve process as an approved 
SG tube repair methodology and not completing the additional 
corrosion testing necessary to establish the design life for the 
kinetically welded sleeve in the presence of a crevice will not 
create the possibility of a new or different type of accident from 
any accident previously evaluated.
    Repair of an SG tube with a kinetically welded sleeve would not 
have provided a mechanism that resulted in an accident outside of 
the area affected by the sleeve. Any hypothetical accident as a 
result of potential tube or sleeve degradation in the repaired 
portion of the tube would have been bounded by the existing SGTR 
analysis. The SGTR analysis accounts for the installation of sleeves 
and the impact on current plugging level analyses. The sleeve design 
does not affect any other component or location of the tube outside 
of the immediate area repaired. Furthermore, there are currently no 
kinetically welded sleeves installed in either plant.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The TIG welded sleeving repair of degraded steam generator tubes 
has been shown by analysis to restore the integrity of the tube 
bundle to its original design basis condition. The safety factors 
used in the design of the sleeves for the repair of degraded tubes 
are consistent with the safety factors in the ASME Boiler and 
Pressure Vessel Code Section III used in steam generator design. The 
design of the ABB/CE SG sleeves has been verified by testing to 
preclude leakage during normal and postulated accident conditions.
    The portion of the installed sleeve assembly which represents 
the reactor coolant pressure boundary can be monitored for the 
initiation and progression of sleeve/tube wall degradation, thus 
satisfying the requirement of RG 1.83. The portion of the SG tube 
bridged by the sleeve joints is effectively removed from the 
pressure boundary, and the sleeve then forms the new pressure 
boundary. The sleeve enhances the safety of the plant by 
reestablishing the protective boundaries of the steam generator. 
Keeping the tube in service with the use of a sleeve instead of 
plugging the tube and removing it from service increases the heat 
transfer efficiency of the steam generator. During each scheduled 
inservice inspection, each sleeve inspected and found to have 
unacceptable degradation shall be removed from service.
    The effect on the design transients and the accident analyses 
have been revised based on the installation of sleeves equal to the 
tube plugging level coincident with the minimum reactor coolant flow 
rate. Evaluation of the installation of sleeves was based on the 
determination that LOCA evaluations for the licensed minimum reactor 
coolant flow bound the combined effect of tube plugging and sleeving 
up to an equivalent of the actual plugging limit. Sleeving results 
in a fractional amount of the plugging limitation of one tube and is 
a preferable alternative when considering core margins based on 
minimum reactor coolant system flow rates. The sleeving installation 
will result in a resistance to primary coolant flow through the 
tube. The primary coolant flow through the ruptured tube is reduced 
by the influence of the installed sleeve; therefore, the 
consequences to the public due to an SGTR event have not increased.
    As SG sleeve removes an indication of a possible leak source 
from the reactor coolant system (RCS) pressure boundary, eliminating 
the potential of a primary-to-secondary leak. The structural 
integrity of the tube is maintained by the sleeve and sleeve-to-tube 
joint.
    Installation of either tube sheet or tube support plate sleeves 
will increase the protective boundaries of the steam generators and 
will not reduce the margin of safety.
    Removal of the kinetically welded sleeve process as an approved 
SG tube repair methodology will not result in a reduction in the 
margin of safety. There are currently no kinetically welded sleeves 
installed in either plant. SG tube integrity will be maintained by 
applying an alternate NRC approved repair methodology or removing 
the SG tube from service by plugging.
    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.

[[Page 19831]]

    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: March 10, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specifications 3.4.5, ``Steam Generators,'' and 
associated Bases to allow repair of steam generator tubes by 
installation of sleeves with the Electrosleeving process developed by 
Framatome Technologies, Inc. (FTI).
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The Electrosleeve configuration has been designed and analyzed 
in accordance with the requirements of the ASME [American Society of 
Mechanical Engineers] Code. The applied stresses and fatigue usage 
for the Electrosleeve are bounded by the limits established in the 
ASME Code. Minimum material property values are used for the 
structural and plugging limit analysis. Mechanical testing has shown 
that the structural strength of nickel Electrosleeves under normal, 
upset, and faulted conditions provides margin to the acceptance 
limits. These acceptance limits bound the most limiting (3 times 
normal operating pressure differential) burst margin recommended by 
Regulatory Guide 1.121. Leakage testing has shown that the 
Electrosleeve is essentially leaktight during all plant conditions.
    The Electrosleeve nominal wall thickness depth-based plugging 
limit is determined using the guidance of Regulatory Guide 1.121 and 
the pressure stress equation of Section III of the ASME Code. The 
limiting requirement of Regulatory Guide 1.121 for the 
Electrosleeve, which applies to part through wall degradation, is 
the minimum acceptable wall thickness to maintain a safety factor of 
three against tube failure under normal operating conditions. A 
bounding set of design and transient loading input conditions was 
used for the minimum wall thickness evaluation in the generic 
evaluation. Evaluation of the minimum acceptable wall thickness for 
normal, upset and postulated accident condition loading per the ASME 
Code indicates these conditions are bounded by the design minimum 
wall thickness.
    Bounding tube wall degradation growth rate per cycle and 
nondestructive examination uncertainty has been assumed for 
determining the Electrosleeve technical specification plugging 
limit. Electrosleeve wall degradation extent determined by 
nondestructive examination, which would require plugging 
Electrosleeved tubes, is developed using the guidance of Regulatory 
Guide 1.121 and is defined in FTI Topical Report BAW-10219P, 
Revision 1, to be 20% throughwall of the nominal sleeve wall 
thickness.
    The effect of Electrosleeving and plugging will remain below the 
plugging limit assumed in the UFSAR [Updated Final Safety Analysis 
Report]. The proposed change will not increase the consequences of 
these accidents.
    The results of the analyses and testing demonstrate that the 
Electrosleeve is an acceptable means of maintaining tube integrity. 
Further, per Regulatory Guide 1.83 recommendations, the 
Electrosleeved tube can be monitored through periodic inspections 
with present NDE [nondestructive examination] techniques. These 
measures demonstrate that installation of Electrosleeves spanning 
degraded areas of the tube will restore the tube to a condition 
consistent with its original design basis.
    Since the main steamline break post-accident primary-to-
secondary leakage is not increased by the presence of 
Electrosleeves, the consequences of an accident previously evaluated 
in the UFSAR are not increased. Conformance of the Electrosleeve 
design with the applicable sections of the ASME Code and results of 
the leakage and mechanical tests support the conclusion that 
installation of Electrosleeves does not increase the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Electrosleeving will not adversely affect any plant component. 
Stress and fatigue analysis of the repair has shown that the ASME 
Code and Regulatory Guide 1.121 criteria are not exceeded. 
Implementation of Electrosleeving maintains overall tube bundle 
structural and leakage integrity at a level consistent with that of 
the original tubing during all plant conditions. Leak and mechanical 
testing of Electrosleeves support the conclusions of the 
calculations that each Electrosleeve retains both structural and 
leakage integrity during all conditions. Electrosleeving of tubes 
does not provide a mechanism resulting in an accident outside of the 
area affected by the Electrosleeves. Any accident resulting from 
potential tube or Electrosleeve degradation in the repaired portion 
of the tube is bounded by the existing tube rupture accident 
analysis.
    Implementation of Electrosleeving will reduce the potential for 
primary-to-secondary leakage while not significantly impacting 
available primary coolant flow area in the event of a LOCA. By 
effectively isolating degraded areas of the tube through repair, the 
potential for steamline break leakage is reduced. These degraded 
intersections now are returned to a condition consistent with the 
Design Basis. While the installation of an Electrosleeve reduces 
primary coolant flow, the reduction is far below that caused by 
plugging. Greater primary coolant flow area is maintained through 
Electrosleeving versus plugging. Therefore, the possibility of a new 
or different kind of accident from any accident previously evaluated 
is not created.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The Electrosleeve repair of degraded steam generator tubes has 
been shown by analysis to restore the integrity of the tube bundle 
consistent with its original design basis condition. The tube/
Electrosleeve operational and faulted condition stresses are bounded 
by the ASME Code requirements and the Electrosleeved tubes are 
leaktight. The safety factors used in the design of Electrosleeves 
for the repair of degraded tubes are consistent with the safety 
factors in the ASME Code used in steam generator design. The 
portions of the installed Electrosleeve assembly which represent the 
reactor coolant pressure boundary can be monitored for the 
initiation and progression of Electrosleeve/tube wall degradation, 
thus satisfying the requirements of Regulatory Guide 1.83. The 
portion of the tube bridged by the Electrosleeve is effectively 
removed from the pressure boundary, and the Electrosleeve then forms 
the new pressure boundary. The areas of the Electrosleeved tube 
assembly which require inspection are defined in Framatome 
Technologies Inc. Topical Report BAW-10219P, Revision 1.
    In addition, since the installed Electrosleeve represents a 
portion of the pressure boundary, a baseline inspection of these 
areas is required prior to operation with Electrosleeves installed. 
The effect of sleeving on the design transients and accident 
analyses has been reviewed based on the installation of 
Electrosleeves up to the level of steam generator tube plugging 
coincident with the minimum reactor coolant flow rate and UFSAR and 
has been found acceptable.
    It is concluded that the proposed license amendment request does 
not result in a significant reduction in the margin of safety as 
defined in the UFSAR or technical specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: March 27, 1997

[[Page 19832]]

    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) would modify the limiting condition for 
operation (LCO) and surveillance requirements (SR) for the ultimate 
heat sink. The ultimate heat sink for Millstone Unit No. 2 is the Long 
Island Sound that transfers heat from safety-related systems during 
normal and accident conditions. Specifically, TS LCO 3.7.11 would be 
changed to indicate that the ultimate heat sink is operable at a water 
temperature of less than or equal to 75  deg.F instead of an average 
value. TS SRs 4.7.11.a and .b would also delete the use of average when 
verifying the water temperature and delete the reference to a specific 
monitoring location, the Unit No. 2 intake structure. These proposed 
changes do not change the ultimate heat sink temperature limit, which 
remains at a maximum of 75  deg.F.
    The TS Bases 3/4.7.11 would also be modified to reflect the above 
changes and to identify the various locations that the ultimate heat 
sink temperature can be measured.
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes do not involve an SHC [significant hazards 
consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes remove the reference to a monitoring 
location where the temperature of the ultimate heat sink is measured 
and eliminate the use of an average ultimate heat sink temperature. 
The instruments used provide information to the operators which will 
permit them to ensure that the plant is operated within the design 
basis of the plant. The subject instruments will provide an accurate 
representation of the ultimate heat sink temperature. This role is 
passive; thus, these instruments cannot initiate or mitigate any 
accident.
    The locations used to monitor the ultimate heat sink temperature 
will be maintained in the Bases. This is a licensee controlled 
document which is maintained under the requirements of 10CFR50.59. 
The details being removed from the Technical Specifications are not 
assumed to be an initiator of any analyzed event. Since any changes 
to the relocated details will be evaluated per 10CFR50.59, any 
possible increase in the probability or consequences of an accident 
previously evaluated will be addressed.
    The proposed changes do not revise the ultimate heat sink 
temperature limit of 75  deg.F. The current analysis is based on the 
ultimate heat sink temperature limit of 75  deg.F. Therefore, there 
is no effect on the consequences of any accident previously 
evaluated.
    Thus, the license amendment request does not impact the 
probability of an accident previously evaluated nor does it involve 
a significant increase in the consequences of an accident previously 
evaluated.
    2. Created the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes remove the reference to a monitoring 
location where the temperature of the ultimate heat sink is measured 
and eliminate the use of an average ultimate heat sink temperature. 
The instruments used provide information to the operators which will 
permit them to ensure that the plant is operated within the design 
basis of the plant. The subject instruments will provide an accurate 
representation of the ultimate heat sink temperature. This role is 
passive, thus, these instruments cannot initiate or mitigate any 
accident.
    The proposed changes will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. They will not alter assumptions 
made in the safety analysis and licensing basis.
    The locations used to monitor the ultimate heat sink temperature 
will be maintained in the Bases. This is a licensee controlled 
document which is maintained under the requirements of 10CFR50.59. 
Thus, adequate control of information will be ensured.
    Therefore, the changes 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.
    The proposed changes remove the reference to a monitoring 
location where the temperature of the ultimate heat sink is measured 
and eliminate the use of an average ultimate heat sink temperature. 
They do not change the ultimate heat sink temperature limit of 75 
deg.F, which is assumed by the current analysis. Therefore, there is 
no effect on the consequences of any accident previously evaluated 
and no significant impact on offsite doses associated with 
previously evaluated accidents. Thus, there is no significant 
reduction in the margin of safety for the design basis accident 
analysis. The license amendment request does not result in a 
reduction of the margin of safety as defined in the Bases for 
Technical Specification 3.7.11. The instruments used provide 
information to the operators which will permit them to ensure that 
the plant is operated within the design basis of the plant. The 
subject instruments will provide an accurate representation of the 
ultimate heat sink temperature. The proposed changes do not alter 
the way any structure, system, or component functions and do not 
alter the manner in which the plant is operated. They do not have 
any impact on the protective boundaries (e.g., fuel matrix and 
cladding, reactor coolant system pressure boundary, and primary and 
secondary containment), or on the safety limits for these 
boundaries.
    The locations used to monitor the ultimate heat sink temperature 
will be maintained in the Bases. The Bases are a licensee controlled 
document which is maintained under the requirements of 10CFR50.59. 
Since any future changes to this license controlled document will be 
evaluated per the requirements of 10CFR50.59, any possible reduction 
(significant or insignificant) in a margin of safety will be 
addressed.
    Thus, the license amendment request does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270NRC 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: March 31, 1997
    Description of amendment request: The proposed amendment would 
modify Technical Specification Surveillance Requirement 4.7.1.2.1.b 
which requires the testing of the auxiliary feedwater motor-driven and 
turbine-driven pumps on recirculation flow at least once per 92 days. 
The proposed amendment would also makes changes to the appropriate 
Bases section.
    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 changes in accordance with 10CFR 
50.92 and has concluded that the changes do not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR 50.92(c) are not 
satisfied. The proposed changes do not involve [an] SHC because the 
changes would not:

[[Page 19833]]

    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes to Technical Specification Surveillance 
4.7.1.2.1.b to increase the required test parameter for the motor 
driven pumps from 1460 psid to 1468 psid and replacing the current 
parameters for the motor driven and turbine driven pumps from 
differential pressure measured in psid [pounds per square inch 
differential] to total head measured in feet are consistent with 
equipment design criteria and does not significantly increase the 
probability of an accident previously evaluated.
    The proposed changes to increase the required test parameter for 
the motor driven pumps from 1460 psid to 1468 psid and replacing the 
current parameters for the motor driven and turbine driven pumps 
from differential pressure measured in psid to total head measured 
in feet provides the necessary assurance that the pumps will 
function as required in accident analyses and does not significantly 
increase the consequence of an accident previously evaluated.
    The moving of the reference to Specification 4.0.5 in order to 
clarify that it applies to the testing of the motor driven and 
turbine driven pumps and the modifications to the bases section are 
administrative and do not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    Therefore, the proposed changes do 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 proposed changes to Technical Specification Surveillance 
4.7.1.2.1.b to increase the required test parameter for the motor 
driven pumps from 1460 psid to 1468 psid and replacing the current 
parameters for the motor driven and turbine driven pumps from 
differential pressure measured in psid to total head measured in 
feet does not change the operation of the auxiliary feedwater system 
or any of its components during normal or accident evaluations.
    The moving of the reference to Specification 4.0.5 in order to 
clarify that it applies to the testing of the motor driven and 
turbine driven pumps and the modifications to the bases section are 
administrative and do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, the proposed changes do 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 proposed change to Technical Specification Surveillance 
4.7.1.2.1.b to increase the referenced total head of the motor
    driven auxiliary feedwater pumps during surveillance testing 
provides an acceptable margin between the required surveillance and 
design pump performance to provide assurance that the pumps will 
operate consistent with system evaluations and does not involve a 
significant reduction in a margin of safety.
    The change in the referenced units from differential pressure 
measured in psid to total head measured in feet for the motor driven 
auxiliary and turbine driven auxiliary feedwater pumps during 
surveillance testing is to account for the effect of water density 
on pump performance during each test and does not involve a 
significant reduction in a margin of safety.
    The moving of the reference to Specification 4.0.5 in order to 
clarify that it applies to the testing of the motor driven and 
turbine driven pumps and the modifications to the bases section are 
administrative and do not involve a significant reduction in a 
margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed changes do 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 06141-0270NRC 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: March 31, 1997
    Description of amendment request: The proposed amendment would 
separate the required testing of motor-operated valve thermal overload 
protection into two new surveillances.
    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 change in accordance with 
10CFR50.92 and has concluded that the change does not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed change does not involve a SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes to the surveillance testing of the motor-
operated valve thermal overload protection are consistent with 
equipment design criteria and performing surveillance testing does 
not significantly increase the probability of an accident previously 
evaluated. The proposed changes to the surveillance testing provides 
the necessary assurance that the motor operated valve thermal 
overload protection will function as required and does not involve a 
significant increase in the consequence of an accident previously 
evaluated.
    Therefore, the proposed changes do 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 proposed changes to the surveillance testing of the motor-
operated valve thermal overload protection does not change the 
operation of any system or system component during normal or 
accident evaluations.
    Therefore, the proposed change 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 proposed changes to the surveillance testing of the motor-
operated valve thermal overload protection are administrative in 
that the changes to the surveillance only clarify that following 
maintenance on the motor starter, a channel calibration is required 
only on that valve. The surveillance continues to require periodic 
representative sample testing.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed change 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 06141-0270NRC Deputy Director: Phillip F. McKee

[[Page 19834]]

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

    Date of amendment request: March 4, 1997
    Description of amendment request: The amendments would modify the 
Emergency Core Cooling System (ECCS) surveillance test acceptance 
criteria in Technical Specification 3/4.5.2 for the Centrifugal 
Charging (CH) and the Safety Injection (SI) pumps. The changes to the 
specified flow values would account for system alignments that effect 
the suction pressure to the pumps. In the recirculation mode, increased 
flow occurs when the CH and SI pumps take suction from the discharge of 
the Residual Heat Removal pumps.
    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 evaluations performed by Westinghouse determined 
that, with the proposed changes, the subject pumps remain operable 
and the safety analyses criteria remain valid.
    Previous conclusions under LCR [License Change Request] 91-03 
evaluating the consequences of the LOCA [loss-of-coolant-accident] 
considered in the Salem Units 1 & 2 licensing basis remain 
unchanged. With respect to the LOCA, the Peak Cladding Temperature 
(PCT) continues to conform to the 10CFR50.46 guidelines of less than 
2200*F. Evaluation of LOCA mass and energy releases previously found 
acceptable remain valid. Decreasing the acceptance window to 
accommodate the potential of an increase to pump runout flow, 
assures that the current limits on pump runout flows continue to be 
met. This change ensures pump integrity is maintained during the 
accident. The reduction of the flow by throttling valves to 
compensate for the potential suction boost remains within the 
current analyses and therefore more conservative values are being 
proposed. Additionally, the proposed change balances the pump flows 
more appropriately by differentiating between the hot and cold leg 
alignments. Flow to the reactor core is unaffected by the very 
slight reduction in the upper flow limits. Since the design 
limitations continue to be met and the integrity of the reactor 
coolant system pressure boundary is not challenged, offsite dose 
assumptions and calculations remain valid. Further, the ECCS is 
post-accident mitigation system and probability of an accident is 
not increased by this proposed change. Lastly, the correction of 
double use of the word ``the'' in Salem Unit 1 Technical 
Specification section 4.5.2.h.1.a is of editorial nature.
    Based on the above information, the proposed changes do not 
increase the risk 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. No new single failures are initiated. The proposed 
changes will therefore not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed change addresses suction boost by changing the 
Technical Specification surveillance acceptance criteria. The 
typographical correction is of editorial nature.
    3. The proposed change does not involve a significant reduction 
in a margin of safety. The evaluation of LOCA accident analysis 
previously performed by Westinghouse continues to be met and 
verifies that, with the proposed changes to the TS, plant operations 
will be maintained within the bounds of safe, analyzed conditions as 
defined in the UFSAR [Updated Final Safety Analysis Report] and that 
conclusions presented in the UFSAR remain valid. The peak cladding 
temperatures (PTC) remains unchanged as no effective differences in 
the operating parameters have occurred. The typographical correction 
is of editorial nature. The proposed changes will therefore not 
reduce 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: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendments request: March 7, 1997
    Description of amendments request: The proposed amendments would 
allow operability testing for the containment isolation valves listed 
in Table 3.6-1 of the Technical Specifications during a defueled 
status. These proposed changes are technically consistent with the 
requirements of NUREG-1431, Revision 1, ``Westinghouse Standard 
Technical Specifications,'' issued on April 7, 1995.
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    [1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.]
    The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the FSAR [Final Safety Analysis Report]. The proposed changes have 
no impact on the probability of an accident. The containment 
isolation valves will continue to require operability testing. 
Allowing the testing to be performed when the unit is in a defueled 
status will have no impact on any accidents previously evaluated. 
The net effect of these changes is not significant and, as a result, 
does not involve a significant increase in the consequences of an 
accident previously evaluated.
    [2. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.]
    The proposed changes to the Technical Specifications do not 
increase the possibility of a new or different kind of accident than 
any accident already evaluated in the FSAR. No new limiting single 
failure or accident scenario has been created or identified due to 
the proposed changes. Safety-related systems will continue to 
perform as designed. The proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    [3. Involve a significant reduction in a margin of safety.]
    The proposed changes do not involve a significant reduction in 
the margin of safety. Although, as a result of these proposed 
changes, the containment isolation valves could be tested for 
operability while the unit is in a defueled state, there is no 
impact in the accident analyses. These proposed changes are 
technically consistent with the requirements of NUREG-1431, Revision 
1 which has already received the requisite review and approval of 
the NRC staff. Thus 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

[[Page 19835]]

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

    Date of amendment request: August 22, 1996, as supplemented on 
March 28, 1997 (TS 96-02)
    Description of amendment request: The proposed changes would revise 
Section 3.6.5 of the Sequoyah Technical Specifications (TS) and 
associated Bases to lower the minimum TS ice basket weight of 1,155 
pounds to 1,071 pounds. This would reduce the overall weight of ice 
required in the ice condenser from 2,245,320 pounds to 2,082,024 
pounds. The TVA license amendment request also proposed to extend the 
chemical analysis surveillance interval for the ice condenser ice bed 
from 12 months to 18 months based on the provisions of Generic Letter 
93-05.
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of Sequoyah Nuclear Plant (SQN) in accordance with the 
proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    TVA proposes to modify the SQN Unit 1 and Unit 2 TSs [Technical 
Specifications] to revise Surveillance Requirement (SR) 4.6.5.1.d to 
lower SQN's minimum TS basket weight from 1,155 pounds (lbs) to 
1,071 lbs, thus lowering the overall ice condenser weight from 
2,245,320 lbs to 2,082,024 lbs.
    The ice condenser system is provided to absorb thermal energy 
release following a loss-of-coolant accident (LOCA) or high energy 
line break (HELB) and to limit the peak pressure inside containment. 
The current containment analysis for SQN is based on a minimum of 
993 lbs of ice per basket evenly distributed throughout the ice 
condenser at the end of an 18-month refueling cycle. The revised 
containment analysis shows that for the predicted sublimation rate 
of 15 percent for 18 months, an average basket weight of 922 lbs at 
the end of the 18-month period would ensure containment design 
pressure is not exceeded.
    Based on TVA's evaluation and the revised containment analysis, 
TVA considers the reduction of ice weight to be acceptable for 
satisfying the safety function of the ice condenser for an 18-month 
ice weighing interval. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    TVA is also proposing to extend the surveillance interval as it 
pertains to the ice bed chemical analysis. Based on test results, 
both at SQN and the industry, the average boron concentration and pH 
changes are minimal; therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Elimination of the temperature at which the pH of the ice bed is 
determined is an administrative change. Future testing will be 
accomplished in accordance with American Society for Testing and 
Materials Standards recommendations. Therefore, this change cannot 
increase the probability of an accident and the consequences of an 
accident will not increase.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    TVA's request to lower the TS limit for ice weight at the start 
of the surveillance interval will not result in a new or different 
kind of accident from that previously analyzed in SQN's Final Safety 
Analysis
    Report. SQN's ice condenser serves to limit the peak pressure 
inside containment following a LOCA. TVA has evaluated the revised 
containment pressure analysis for SQN (Enclosure 4, Westinghouse 
WCAP-12455, Revision 1) and determined that sufficient ice would be 
present at all times to keep the peak containment pressure below 
SQN's containment design pressure of 12 pounds per square inch gage 
(psig). Therefore, this change would not result in a new or 
different kind of accident from any previously analyzed.
    The proposed reduced testing frequency of the chemical 
composition of the ice bed does not change the manner in which the 
plant is operated. Additionally, the ice condenser is a passive 
system that reacts to an accident, but does not support plant 
operation on a daily basis. The reduced testing frequency of the ice 
bed chemical composition does not generate any new accident 
precursors; therefore, the possibility of a new or different kind of 
accident from any previously analyzed is not created.
    Elimination of the temperature at which the pH of the ice bed is 
determined is an administrative change. This change cannot create 
the possibility of a new or different kind of accident.
    3. Involve a significant reduction in a margin of safety.
    The ice condenser system is provided to absorb thermal energy 
release following a LOCA and to limit the peak pressure inside 
containment. The current ice condenser analysis for SQN is based on 
a minimum of 993 lbs of ice per basket. The revised containment 
analysis changes the minimum ice weight assumed in the analysis to 
922 lbs per basket.
    The revised containment analysis shows that using an average 
basket weight of 1,071 lbs and a sublimation allowance of 15 
percent, all bays would have an average basket weight of 922 lbs at 
the end of the 18-month surveillance interval. The revised analysis 
utilizes new mass and energy releases (refer to Westinghouse WCAP-
10325-P-A), which substantially delays ice-bed meltout and limits 
the initial containment peak pressure to approximately 7.15 psig 
during the blowdown phase. The ice-bed meltout delay allows the 
second containment pressure peak, which is driven mainly by the 
decay heat, to be limited to approximately 11.45 psig, which is 
below the containment design pressure of 12 psig.
    Based on TVA's evaluation and the revised containment analysis, 
TVA considers the reduction of the average basket weight to be 
acceptable for satisfying the safety function of the ice condenser 
for the current 18-month interval. Therefore, the proposed change 
does not involve a significant reduction in the margin of safety.
    The proposal to extend the surveillance from 12 to 18 months 
does not change the boron concentration or pH requirements. 
Experience at Duke Power Company, as stated in NUREG-1366, indicates 
that these parameters do not change appreciably when verified every 
9 months. SQN has a similar experience with a 12-month interval. 
Since the boron concentration and the post-LOCA pH requirements 
remain essentially the same, there is no reduction in the margin of 
safety.
    Elimination of the temperature at which the pH of the ice bed is 
determined is an administrative change. Future testing will be 
accomplished in accordance with ASTM recommendations. The difference 
between the pH values determined at the current TS specified 
temperature and the temperature currently recommended by the ASTM 
standards is insignificant. Therefore, there is no reduction in the 
margin of safety.
    The NRC 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: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: September 12, 1996
    Description of amendment request: The proposed change to the 
Technical Specifications is administrative in nature in that it would 
add the NRC standard fire protection license condition to each unit's 
Operating License and relocate the fire protection requirements from 
the Technical Specifications to the Updated Final Safety Analysis 
Report (UFSAR).
    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:

[[Page 19836]]

    Specifically, operation of Surry Power Station with the proposed 
amendment will not:
    1. Involve a significant increase in either the probability of 
occurrence or consequences of any accident or equipment malfunction 
scenario that is important to safety and which has been previously 
evaluated in the UFSAR. The requirements of the Fire Protection 
Program have not been changed by theproposed amendment. Relocation 
of the Fire Protection Program requirements into the UFSAR and 
station procedures does not decrease any portion of the program. The 
same fire protection requirements exist as before the change.
    2. Create the possibility of a new or different type of accident 
than those previously evaluated in the safety analysis report. The 
requirements of the Fire Protection Program have not been changed by 
the proposed amendment. This is an administrative change to relocate 
the Fire Protection Program requirements from the Technical 
Specifications to the UFSAR and station procedures. Consequently, 
the possibility of a new or different kind of accident from any 
accident previously evaluated has not been created.
    3. Involve a significant reduction in a margin of safety. 
Implementation of the Fire Protection Program requirements is 
assured by the UFSAR and station procedures. Since the rogram is 
being retained intact, there is no 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: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: Mark Reinhart, Acting

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

    Date of amendment request: January 16, 1997
    Description of amendment request: The proposed amendments (Point 
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request 
(TSCR) 191) would revise the minimum boron concentration required in 
the refueling water storage tank(s)(RWST), boric acid storage tanks 
(BAST), and safety injection (SI) accumulators during normal operation; 
the minimum boron concentration of primary coolant during refueling 
conditions; and the minimum boron concentration in the reactor when 
positive reactivity could be added and/or boron dilution could occur 
and containment integrity is not intact. These changes are necessary to 
accommodate the planned extension of the operating cycle from 12 months 
to 18 months. The licensee proposes to change TS 15.3.2, ``Chemical and 
Volume Control System,'' TS 15.3.3, ``Safety Injection and Residual 
Heat Removal Systems,'' TS 15.3.6, ``Containment System,'' TS 15.3.8, 
``Refueling,'' and associated 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:
    1. Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probabilities of accidents previously evaluated are based on 
the probability of initiating events for these accidents. Initiating 
events for accidents previously evaluated are described in the PBNP 
FSAR [final safety analysis report].
    In effect, the proposed changes will result in: (1) higher boron 
concentrations of primary coolant during refueling, and (2) higher 
boron inventories in the RWSTs, BASTs, and SI accumulators. These 
changes do not require hardware changes or changes to the operation 
of accident-mitigating equipment. These changes relate to the 
performance capability of particular accident mitigation systems; 
equipment that is not postulated to cause accidents. Therefore, 
these proposed changes do not cause an increase in the probabilities 
of any accidents previously evaluated.
    The consequences of accidents previously evaluated in the PBNP 
FSAR are determined by the results of analyses that are based on 
initial conditions of the plant, the type of accident, transient 
response of the plant, and the operation and failure of equipment 
and systems.
    In effect, the proposed changes will result in: (1) higher boron 
concentrations of primary coolant during refueling, and (2) higher 
boron inventories in the RWSTs, BASTs, and SI accumulators. These 
increased boron concentrations do not increase the probability that 
engineered safety features equipment will fail, nor do these changes 
affect the capability of this equipment to operate as required for 
the accidents previously evaluated in the PBNP FSAR. These changes 
do not require hardware changes or changes to the operation of 
accident-mitigating equipment.
    The consequential effects of a lower containment spray pH will 
not affect the capability of the containment spray to remove 
elemental iodine during design basis LOCA [loss-of-coolant accident] 
accidents. Also, the consequential reduction in containment sump 
water pH will not affect the fluid's capability to retain elemental 
iodine, nor will it adversely increase the potential corrosion rates 
for materials inside containment if the sump water is sprayed into 
containment during the recirculation phase of a LOCA.
    Another consequence of injecting a higher concentration boric 
acid solution into the core during a LOCA may be an abbreviated 
onset to boron precipitation in the post-LOCA core. An incremental 
change in the boron injection concentration would not have 
significant effect on the postulated onset, but each core reload 
safety evaluation will continue to verify that the existing 
emergency operating procedures accommodate the potential for boron 
precipitation.
    Therefore, this proposed license amendment does not affect the 
consequences of any accident previously evaluated in the PBNP FSAR, 
because the factors that are used to determine the consequences of 
accidents are not changed.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    New or different kinds of accidents can only be created by new 
or different accident initiators or sequences. New and different 
types of accidents (different from those that were originally 
analyzed for Point Beach) have been evaluated and incorporated into 
the licensing basis for PBNP. Examples of different accidents that 
have been incorporated into the PBNP licensing basis include 
anticipated transients without scram and station blackout.
    The changes proposed by this TSCR do not create any new or 
different accident initiators or sequences because these changes to 
minimum boron concentrations will not cause failures of equipment or 
accident sequences different than the accidents previously analyzed. 
No new equipment interfaces are created, and no new materials or 
fluids are introduced. The incremental increase in boron 
concentrations will not create a failure mechanism not previously 
known and evaluated. Therefore, these proposed TS changes do not 
create the possibility of an accident of a different type than any 
previously evaluated in the PBNP FSAR.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The margins of safety for Point Beach are based on the design 
and operation of the reactor and containment and the safety systems 
that provide their protection. Plant safety margins are established 
through Limiting Conditions for Operation, Limiting Safety System 
Settings and Safety Limits specified in the Technical 
Specifications. The proposed Technical Specification changes to 
refueling water storage tank (RWST), SI accumulator, and BAST boron 
inventory requirements have all been evaluated to preserve the 
shutdown capability described in the associated bases (boration from 
just critical, hot zero or full power, peak xenon with control rods 
at the

[[Page 19837]]

insertion limit, to xenon-free cold shutdown with the highest worth 
control rod assembly fully withdrawn). Similarly, the proposed TS 
change to the minimum boron concentration of the primary coolant 
system for refueling operations have been evaluated to preserve the 
subcriticality margin described in the associated TS bases (i.e., 5% 
[delta] k/k in the cold condition with all rods inserted).
    Because there are no changes to any of these margins, the 
proposed license amendment does not involve a reduction in any 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: January 21, 1997
    Description of amendment request: The proposed amendments (Point 
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request 
195) would revise TS Section 15.6.11, ``Radiation Protection Program,'' 
to update all references to 10 CFR Part 20, ``Standards for Protection 
Against Radiation,'' to restore consistency between 10 CFR Part 20 
regulations and the PBNP 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:
    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendments are administrative in nature, providing 
consistency between the Point Beach licenses and Commission 
regulations. The amendments do not affect the operation or 
maintenance of any PBNP structure[,] system or component. In 
addition, the regulations and proposed changes provide more 
conservative determinations of high radiation areas, thereby 
potentially resulting in lower personnel radiation exposures during 
normal operation and post accident. The consequences of an accident 
related to personnel radiation exposures may be reduced.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendments are administrative only and do not 
affect the operation or maintenance of any structure[,] system or 
component at Point Beach Nuclear Plant. No new systems or components 
are introduced. Therefore, no new accident initiators or sequences 
result from any previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create a significant reduction in a 
margin of safety.
    The proposed amendments are administrative and reflect 
regulatory requirements that are more conservative than those 
presently reflected in the PBNP Technical Specifications. These more 
conservative requirements result in more conservative designation of 
high radiation areas thereby providing additional margins of safety 
related to the control of radiation exposures to personnel. No 
structure[,] system or component at PBNP at PBNP is changed[,] 
thereby maintaining the margins of safety for the operation of the 
Point Beach Nuclear Plant.
    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: January 24, 1997
    Description of amendment request: The proposed amendments (Point 
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request 
(TSCR) 193) would revise TS 15.5.4, ``Fuel Storage,'' to increase fuel 
assembly enrichment limits to 5.0 w/o U-235 while maintaining Keff in 
the storage pools (spent fuel pool and new fuel storage racks) less 
than 0.95.
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a change to structures, 
systems, or components which would affect the probability or 
consequences of an accident previously evaluated in the PBNP Final 
Safety Analysis Report (FSAR). The only relevant concern with 
respect to increasing enrichment limits in the spent fuel pool and 
new fuel storage racks is one of criticality. The proposed changes 
use the same criticality limit used in the current Technical 
Specifications. Therefore, margin to safe operation of Units 1 and 2 
is maintained. The probability and consequences of an accident 
previously evaluated are dependent on this criticality limit. 
Because the limit will not change, the probability and consequences 
of those accidents previously evaluated will not change.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve a change to plant design. 
The proposed increase in spent fuel pool and new fuel storage racks 
fuel assembly enrichment limits maintains the margin to safe 
operation of Units 1 and 2 because the criticality limit for the 
spent fuel pool and new fuel storage racks will not change. These 
changes do not affect any of the parameters or conditions that 
contribute to the initiation of any accidents. Because the 
criticality limit remains the same, these changes have no effect on 
plant operation, design, or initiation of any accidents. Therefore, 
the proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The proposed changes maintain the margin to safe operation of 
Units 1 and 2. The margin of safety is based on the criticality 
limit of the spent fuel pool and the new fuel storage racks. Because 
this limit will not change, the margin of safety will not be 
affected. Therefore, the proposed changes will not create 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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241

[[Page 19838]]

    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: February 12, 1997, as supplemented on 
March 11, 1997
    Description of amendment request: The proposed amendments (Point 
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request 
196) would relocate turbine overspeed protection specifications, 
limiting conditions for operation, surveillance requirements, and 
associated bases from TS Section 15.3.4, ``Steam and Power Conversion 
System,'' and Section 15.4.1, ``Operational Safety Review,'' to the 
Final Safety Analysis Report (FSAR) in accordance with Generic Letter 
95-10, ``Relocation of Selected Technical Specifications Requirements 
Related to Instrumentation.''
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of Point Beach Nuclear Plant in accordance with the 
proposed amendments will not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments administratively relocate turbine 
overspeed protection Specifications to the Point Beach Final Safety 
Analysis Report (FSAR). The Specifications will be transferred 
verbatim, except for the turbine load limit with the crossover steam 
dump system inoperable, which has already been evaluated under 10 
CFR 50.59 and will be conservatively reduced. In addition, the 
regulatory requirements of 10 CFR 50.55a, ``Codes and Standards, '' 
will still apply to the relocated Specifications. Therefore, 
operation of Point Beach Nuclear Plant in accordance with the 
proposed amendments cannot create an increase in the probability or 
consequences of an accident previously evaluated.
    2. Operation of Point Beach Nuclear Plant in accordance with the 
proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendments administratively relocate Specifications 
to the FSAR and in one case result in a more conservative operating 
limit. Therefore, operation of Point Beach Nuclear Plant in 
accordance with the proposed amendments cannot create a new or 
different kind of accident from any accident previously evaluated.
    3. Operation of Point Beach Nuclear Plant in accordance with the 
proposed amendments will not create a significant reduction in a 
margin of safety.
    The proposed changes are administrative in nature. There is no 
physical change to the facility, its systems, or its operation, 
except for the conservative reduction of the turbine load limit with 
the crossover steam dump system inoperable which has already been 
justified via 10 CFR 50.59. Therefore, operation of PBNP in 
accordance with the proposed amendments cannot 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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: February 17, 1997; supersedes March 24, 
1995, as supplemented by letter dated August 16, 1995, amendment 
request.
    Description of amendment request: This amendment request proposes 
to revise Technical Specification 1.7, ``Containment Integrity,'' 
Technical Specification 3/4.6.1, ``Containment Integrity,'' and 
Technical Specification 3/4.6.3, ``Containment Isolation Valves.'' 
These proposed changes would relocate Technical Specification Table 
3.6-1, ``Containment Isolation Valves,'' to the Wolf Creek Generating 
Station (WCGS) procedures. This proposed change is in accordance with 
the guidance provided in Generic Letter 91-08, ``Removal of Component 
Lists from Technical Specifications,'' dated May 6, 1991. In addition, 
this request proposes that the August 16, 1996, supplemental submittal 
that provided an additional footnote allowing an increased outage time 
for certain component cooling water system valves be withdrawn. The 
determination that the additional footnote is not required supersedes 
the staff's proposed no significant hazards consideration determination 
evaluation for the requested changes that was published on September 
27, 1995 (60 FR 49949).
    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 simplify the technical specifications, meet 
the regulatory requirements for control of containment isolation, 
and are consistent with the guidelines of GL 91-08. The procedural 
details of Technical Specification Table 3.6-1 have not been 
changed, but only relocated to a different controlling document. The 
proposed changes are administrative in nature, should result in 
improved administrative practices, and do not affect plant 
operations.
    The probability of occurrence of a previously evaluated accident 
is not increased because this change does not introduce any new 
potential accident initiating conditions. The consequences of an 
accident previously evaluated is not increased because the ability 
of containment to restrict the release of any fission product 
radioactivity to the environment will not be degraded by this 
change.
    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 are administrative in nature, do not result 
in physical alterations or changes to the operation of the plant, 
and cause no change in the method by which any safety-related system 
performs its function. Therefore, this proposed change 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 administrative change to relocate Technical Specification 
Table 3.6-1 to appropriate plant procedures does not alter the basic 
regulatory requirements for containment isolation and will not 
adversely affect containment isolation capability for Coordinator 
credible accident scenarios. Adequate control of the content of the 
table is assured by existing plant procedures.
    The proposed relocation of Technical Specification Table 3.6-1 
does not alter current technical specification requirements for 
containment isolation valve operability. The LCO and Surveillance 
Requirements would be retained in the revised technical 
specifications. Therefore, the proposed change will not affect the 
meaning, application, and function of the current technical 
specification requirements for the valves in Table 3.6-1.
    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.

[[Page 19839]]

    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: March 18, 1997
    Description of amendment request: This license amendment request 
revises Technical Specification Surveillance Requirement 4.5.2.c to 
clarify when a containment entry visual inspection is required. This 
proposed change to reduce the visual inspection requirement to at least 
once daily is in accordance with the guidance provided in Generic 
Letter 93-05, ``Line-Item Technical Specifications Improvements to 
Reduce Surveillance Requirements for Testing During Power Operation.''
    Basis for proposed no significant Hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Implementing the proposed change could potentially increase the 
chances of loose debris (trash, rags, clothing, etc.) being left in 
containment for some period of time greater than would be allowed 
under current surveillance requirements. However, the proposed 
change also clarifies that the visual inspection must be performed 
at least once daily. Therefore, the period of time that debris could 
be left uncontrolled inside containment would still be less than 24 
hours. Based on work controls placed on material entry/exit into 
containment and personnel training on housekeeping controls inside 
containment, and the results of past surveillances, it is unlikely 
that a significant amount of debris would be left uncontrolled 
inside containment for this period of time. Also, based on 
containment sump design, relatively small amounts of debris would 
not be sufficient to cause a significant amount of blockage of the 
sump screens.
    The probability of occurrence of a previously evaluated accident 
is not increased because the reduced frequency of the visual 
inspection does not cause a significant impact on the possibility of 
containment sump screen blockage. Therefore containment sump 
operability is not affected by the proposed change. In addition, the 
proposed change will not result in any changes to the design or 
operation of any plant systems or components.
    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 decreases the frequency of performing a 
visual inspection for loose debris in containment, but does not 
result in a change to the design or operation of any plant system or 
component. The purpose of the inspection is to ensure that there is 
no loose debris, left in containment following a containment entry, 
that could potentially block the containment sump screens during 
LOCA conditions. Delaying this inspection until the last containment 
entry each day will not result in a significant amount of debris 
being left in containment, based on housekeeping practices 
controlling the entry/removal of trash and material into/from 
containment; training of employees to increase awareness of material 
control in radiologically-controlled areas; and retaining the 
requirement to perform a visual inspection at least once per day 
when containment entries are made (during periods when containment 
integrity is established), thereby ensuring that trash and debris 
can be identified and removed on a daily basis (on days containment 
entries are made).
    Based on the above, this proposed change 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 purpose of performing a visual inspection of areas affected 
by a containment entry is to ensure any debris or trash generated by 
the activity during the containment entry is identified and removed 
from containment. This ensures that no trash or debris is left in 
containment that could be transported to and block the containment 
sump screens during LOCA conditions. Based on current material 
control and housekeeping practices imposed on containment entry/
exit, and past inspection results, reducing the surveillance 
requirement to a once per day basis, on days containment entries are 
made, would not result in a significant amount of trash or debris 
being left in containment following completion of the entry, and any 
such material would be identified and removed prior to the end of 
the day. The amount of trash or debris that could be left in 
containment for a 24 hour period would be significantly less than 
the amount that would be required to cause sump screen blockage 
sufficient to affect sump performance. Therefore, the proposed 
change will not result in a significant reduction in the margin of 
safety of any plant system or equipment.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: March 18, 1997
    Description of amendment request: This license amendment request 
revises Technical Specification Section 5.3.1, Fuel Assemblies, to 
allow the use of an alternate zirconium based fuel cladding material, 
ZIRLO. Wolf Creek Nuclear Operating Corporation (WCNOC) is planning to 
insert Westinghouse fuel assemblies containing ZIRLO fuel rod cladding 
during the ninth refueling outage, which is currently scheduled to 
begin in late September 1997.
    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 methodologies used in the accident analysis remain 
unchanged. The proposed changes do not change or alter the design 
assumptions for the systems or components used to mitigate the 
consequences of an accident. Use of ZIRLO fuel cladding does not 
adversely affect fuel performance or impact nuclear design 
methodology. Therefore accident analyses are not impacted.
    The operating limits will not be changed and the analysis 
methods to demonstrate operation within the limits will remain in 
accordance with NRC approved methodologies. Other than the changes 
to the fuel assemblies, there are no physical changes to the plant 
associated with this technical specification change. A safety 
analysis will continue to be performed for each cycle to demonstrate 
compliance with all fuel safety design basis.
    VANTAGE 5H with IFMs fuel assemblies with ZIRLO clad fuel rods 
meet the same fuel assembly and fuel rod design bases as other 
VANTAGE 5H with IFMs fuel assemblies. In addition, the 10 CFR 50.46 
criteria are applied to the ZIRLO clad rods. The use of these fuel 
assemblies will not result in a change to the reload design and 
safety analysis limits. The clad material is similar

[[Page 19840]]

in chemical composition and has similar physical and mechanical 
properties as Zircaloy-4. Thus, the cladding integrity is maintained 
and the structural integrity of the fuel assembly is not affected. 
ZIRLO cladding improves corrosion performance and dimensional 
stability. No concerns have been identified with respect to the use 
of an assembly containing a combination of Zircaloy-4 and ZIRLO clad 
fuel rods. Since the dose predictions in the safety analyses are not 
sensitive to fuel rod cladding material, the radiological 
consequences of accidents previously evaluated in the safety 
analysis remain valid.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident or 
malfunction of equipment important to safety previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    VANTAGE 5H with IFMs fuel assemblies with ZIRLO clad fuel rods 
satisfy the same design bases as those used for other VANTAGE 5H 
with IFMs fuel assemblies. All design and performance criteria 
continue to be met and no new failure mechanisms have been 
identified. Since the original design criteria are met, the ZIRLO 
clad fuel rods will not be an initiator for any new accident or 
malfunction of equipment important to safety. The ZIRLO cladding 
material offers improved corrosion resistance and structural 
integrity.
    The proposed changes do not affect the design or operation of 
any system or component in the plant. The safety functions of the 
related structures, systems or components are not changed in any 
manner, nor is the reliability of any structure, system or component 
reduced. The changes do not affect the manner by which the facility 
is operated and do not change any facility design feature, structure 
or system. No new or different type of equipment will be installed. 
Since there is no change to the facility or operating procedures, 
and the safety functions and reliability of structures, systems and 
components are not affected, the proposed changes do not create the 
possibility of a new or different kind of accident or malfunction of 
equipment important to safety from any accident or malfunction of 
equipment important to safety previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Use of ZIRLO cladding material does not change the VANTAGE 5H 
with IFMs reload design and safety limits. The use of these fuel 
assemblies will take into consideration the normal core operating 
conditions allowed in the Technical Specifications. For each cycle 
reload core, the fuel assemblies will be evaluated using NRC 
approved reload design methods, including consideration of the core 
physics analysis peaking factors and core average linear heat rate 
effects.
    The use of Zircaloy-4, ZIRLO or stainless steel filler rods in 
fuel assemblies will not involve a significant reduction in the 
margin of safety because analyses using NRC approved methodologies 
will be performed for each configuration to demonstrate continued 
operation within the limits that assure acceptable plant response to 
accidents and transients. These analyses will be performed using NRC 
approved methods that have been approved for application to the fuel 
configuration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    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: March 27, 1997
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications for the Brunswick Steam Electric 
Plant Units 1 and 2 to eliminate certain instrumentation response time 
testing requirements in accordance with NRC-approved BWR Owners Group 
Topical Report NEDO-32291-A, ``System Analysis for the Elimination of 
Selected Response Time Testing Requirements.''Date of publication of 
individual notice in Federal Register: April 1, 1997 (62 FR 15542)
    Expiration date of individual notice: May 1, 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

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

    Date of amendment request: March 25, 1997
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3/4.4.9, ``Specific Activity,'' and 
associated Bases to reduce the limit associated with dose equivalent 
iodine-131. The steady-state dose equivalent iodine-131 limit would be 
reduced by 40 percent from .5 [micro]Curie/gram to .3 [micro]Curie/
gram.
    Date of publication of individual notice in Federal Register: April 
4, 1997 (62 FR 16201)
    Expiration date of individual notice: May 5, 1997
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

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

[[Page 19841]]

assessment need be prepared for these amendments. If the Commission has 
prepared an environmental assessment under the special circumstances 
provision in 10 CFR 51.12(b) and has made a determination based on that 
assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: February 19, 1997, as 
supplemented April 3, 1997.
    Brief description of amendments: The amendments would delete the 
24/48 Volt direct current (Vdc), batteries, battery chargers and 
distribution systems from the Technical Specifications (TSs) for Unit 
3, by adding a footnote to indicate that these TSs are only applicable 
to Unit 2. All safety-related loads associated with the 24/48 Vdc 
batteries for Unit 3 will be relocated to other safety-related battery 
systems which are in the TSs.
    Date of issuance: April 10, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 156 and 151
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 5, 1997 (62 FR 
10088). The April 3, 1997, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
April 10, 1997. No significant hazards consideration comments received: 
No
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450

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

    Date of application for amendment: August 14, 1996, as supplemented 
September 13, 1996
    Brief description of amendment: The amendment revises Technical 
Specification Sections 3.3 and 6.9.1.9; and the basis of Section 3.3, 
3.6 and 3.10. The changes incorporate the best estimate approach into 
the licensing basis for the Indian Point Unit No. 2 loss-of-coolant 
accident analysis.
    Date of issuance: March 31, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 188
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4344) The September 13, 1996, supplemental letter did not change the 
initial proposed no significant hazards consideration.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated March 31, 1997.No significant hazards consideration comments 
received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610

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

    Date of application for amendment: February 14, 1997, as 
supplemented March 12, 1997.
    Brief description of amendment: The amendment revises Technical 
Specification Section 4.13-2 to allow a one-time extension of the 
interval for steam generator tube inspection. Specifically, the date 
for commencement of the steam generator tube inspection is extended 
from April 14, 1997 to May 2, 1997.
    Date of issuance: April 9, 1997
    Effective date: As of the date of issuance to be implemented by 
April 14, 1997.
    Amendment No.: 189
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 4, 1997 (62 FR 
9816) The March 12, 1997, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 9, 1997.No 
significant hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of application for amendment: November 7, 1996
    Brief description of amendment: The amendment revised Technical 
Specification 4.2.9, Service and Instrument Air System, to add an 
additional air compressor.
    Date of issuance: April 2, 1997
    Effective date: Effective the date of issuance.
    Amendment No.: 118
    Facility Operating License No. DPR-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66706) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 2, 1997.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770

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

    Date of application for amendments: January 3, 1997, as 
supplemented by letter dated March 20, 1997
    Brief description of amendments: The amendments revise Technical 
Specification Tables 3.3-2, 3.3-4, 3.3-5, 4.3-2 and Bases Sections 3/
4.3.1 and 3/4.3.2 to eliminate the safety injection signal on low steam 
line pressure.
    Date of issuance:  April 3, 1997
    Effective date: For Unit 1, as of the date of issuance to be 
implemented before startup from the next refueling outage; For Unit 2, 
as of the date of issuance to be implemented before startup from the 
current refueling outage
    Amendment Nos.: 158 and 150
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4345) The March 20, 1997, letter provided clarifying information that 
did not change the scope of the original January 3, 1997, 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 April 3, 1997.No significant hazards 
consideration comments received: No

[[Page 19842]]

    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendments: February 5, 1997
    Brief description of amendments: The amendments reflect replacement 
of the existing source and intermediate range nuclear instrumentation 
with a new source range and wide range nuclear instrumentation system 
that provides more channels and continuous coverage from the Source 
Range to above the Power Range.
    Date of issuance: March 31, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 223, 223, 220
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8796) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 31, 1997.No significant 
hazards consideration comments received:
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: October 16, 1996
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications by revising Table 4.3-1 to expand the 
applicability for Core Protection Calculator (CPC) operability and to 
allow the use of a cycle independent shape annealing matrix in the 
CPCs.
    Date of issuance: April 11, 1997
    Effective date: April 11, 1997, to be implemented within 60 days
    Amendment No.: 125
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6575) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 11, 1997No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: December 2, 1996 as supplemented by 
letter dated February 4 and March 14, 1997
    Brief description of amendment: The amendment changes the Technical 
Specifications to reflect the approval for the licensee to use of the 
new Containment Leakage Rate Testing Program as required by 10 CFR Part 
50 Appendix J, Option B for Waterford Steam Electric Station, Unit 3.
    Date of issuance: April 10, 1997
    Effective date: April 10, 1997
    Amendment No.: 124
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 15, 1997 (62 FR 
2189) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 1997.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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

    Date of application for amendment: December 9, 1996
    Brief description of amendment: This amendment modifies technical 
specifications for selected cycle-specific reactor physics parameters 
to refer to the St. Lucie Unit 1 Core Operating Limits Report for 
limiting values.

    Date of issuance:  April 1, 1997
    Effective date: April 1, 1997
    Amendment No.: 150
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 15, 1997 (62 FR 
2189) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 1, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of applications for amendment: June 20, 1995, as supplemented 
August 30, 1995, and January 17, 1996
    Brief description of amendment: The amendment relocates the 
applicable requirements of Technical Specification (TS) 3.6.3 for the 
main steam line isolation valves (MSIVs) to TS 3.7.1.5, ``Main Steam 
Line Isolation Valves.'' In addition, the Applicability section of TS 
3.7.1.5 is revised to indicate that Specification 3.7.1.5 is applicable 
in Mode 1 and in Modes 2, 3, and 4, except where all MSIVs are closed 
and deactivated (i.e., in Modes 2, 3, and 4, TS 3.7.1.5 is applicable 
only if the MSIVs are open). Also, the Action Statement for the 
Limiting Condition for Operation 3.7.1.5 has been revised using the 
guidance of the Improved Standard Technical Specifications for 
Westinghouse plants (NUREG-1431). The amendment also deletes a license 
requirement to submit responses to and to implement requirements of 
Generic Letter 83-28, because the requirement has been completed. 
Generic Letter 83-28 pertains to the Salem anticipated transient 
without scram event. In addition, the amendment incorporates TS Bases 
submitted by Northeast Nuclear Energy Company by letters dated June 20, 
1995, February 5, 1996, and March 21 and 26, 1997. Since all four Bases 
changes affect Section B 3/4.7 of the TS, the NRC staff is using them 
in a group to avoid errors in revising the TS.
    Date of issuance: April 10, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 136
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications and License Condition.
    Date of initial notice in Federal Register: August 2, 1995 (61 FR 
39445) and February 28, 1996 (61 FR 7555)The August 30, 1995, letter 
provided clarifying information that did not change the scope of the 
June 20, 1995, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 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,

[[Page 19843]]

Norwich, Connecticut and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, Connecticut 06385

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-
336, and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, 
and 3, New London, Connecticut

    Date of application for amendments: February 3, 1997
    Brief description of amendments: The amendments revise Section 6, 
``Administrative Controls,'' of the Millstone Unit Nos. 1, 2, and 3 
Technical Specifications to reflect organizational changes that have 
been implemented in the Nuclear Division.
    Date of issuance: April 10, 1997
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 99, 206, and 135
    Facility Operating License Nos. DPR-21, DPR-65, and NPF-49: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8800) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 10, 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

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

    Date of application for amendments: November 18, 1996
    Brief description of amendments: These amendments change the 
Technical Specifications for Susquehanna Steam Electric Station (SSES), 
Units 1 and 2 by increasing the maximum isolation times for reactor 
core isolation cooling inboard warm-up line isolation valves from 3 
seconds to 12 seconds, high pressure core injection inboard warm-up 
line siolation valves from 3 seconds to 6 seconds and reactor 
recirculation process sample line isolation valves from 2 seconds to 9 
seconds.
    Date of issuance: April 7, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 164 and 135
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 15, 1997 (61 FR 
2191) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 7, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

Pennsylvania Power and Light Company, Docket No. 50-388, 
Susquehanna Steam Electric Station, Unit 2, Luzerne County, 
Pennsylvania

    Date of application for amendment:  March 17, 1997
    Brief description of amendment: The amendment modifies the Design 
Features Section 5.3.1 of the Technical Specifications to reflect the 
Atrium-10 design and would include a Siemens Power Corporation topical 
report in Section 6.9.3.2 to reflect mechanical design criteria for 
this fuel. This change would allow this fuel to be loaded into the core 
only under Operational Condition 5 (refueling) and does not permit 
startup or power operation using the Atrium-10 fuel.
    Date of issuance: April 9, 1997
    Effective date: As of date of issuance to be implemented within 30 
days.
    Amendment No.: 136
    Facility Operating License No. NPF-22: This amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (62 FR 14167) March 25, 1997. 
That 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 April 24, 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 of the 
amendment, finding of exigent circumstances, and final determination of 
no significant hazards consideration are contained in a Safety 
Evaluation dated April 9, 1997.
    Attorney for licensee: Jay Silbert, 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

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

    Date of application for amendments: September 19, 1996, as 
supplemented December 17, 1996, January 23 and 31, March 21 and April 
4, 1997
    Brief description of amendments: The amendments revise the 
surveillance requirements addressing the reactor vessel pressure and 
temperature limits.
    Date of issuance: April 4, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 206 and 147
    Facility Operating
    Local Public- Document -Room locations: ments revised the Technical 
Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
128) The December 17, 1996, January 23 and 31, March 21, 1997, and 
April 4, 1997, letters provided clarifying information that did not 
change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 1997.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

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: October 18, 1996 as 
supplemented March 12, March 17, April 4, and April 9, 1997 (TS 96-05)
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) by revising TS 3/4.4.5 and 3.4.6.2 and 
associated Bases to permanently incorporate requirements for steam 
generator tube inspections and repair in the Sequoyah Nuclear Plant, 
Units 1 and 2 TS.
    Date of issuance: April 9, 1997
    Effective date: As of the date of issuance to be implemented no 
later than 45 days of its issuance.
    Amendment Nos.: 222 and 213
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications and license conditions.

[[Page 19844]]

    Date of initial notice in Federal Register: February 11, 1997 (62 
FR 6276) The March 12, March 17, April 4, and April 9, 1997, letters 
provided clarifying information that did not change the scope of the 
October 18, 1996, application and the initial proposed no significant 
hazards consideration determination.The Commission's related evaluation 
of the amendment is contained in a Safety Evaluation dated April 9, 
1997.No significant hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 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 May 23, 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. 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

[[Page 19845]]

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 20555-001, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-001, 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).

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

    Date of application for amendment: April 1, 1997
    Brief description of amendment: The amendment revises Technical 
Specification Table 3.3-3 to correct administrative errors associated 
with the start logic of the turbine driven auxiliary feedwater pump.
    Date of issuance: April 2, 1997
    Effective date: April 2, 1997
    Amendment No.: 119
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No.The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated April 2, 1997.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 200379
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    NRC Project Director: William H. Bateman
    Dated at Rockville, Maryland, this 16th day of April, 1997.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director ,Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 97-10334 Filed 4-22-97; 8:45 am]
BILLING CODE 7590-01-F