[Federal Register Volume 61, Number 2 (Wednesday, January 3, 1996)]
[Notices]
[Pages 174-192]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10103]



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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 December 11, 1995, through December 20, 
1995. The last biweekly notice was published on December 20, 1995 (60 
FR 65672).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By February 2, 1996, 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.

[[Page 175]]

    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: November 1, 1995, as supplemented on 
December 1, 1995
    Description of amendments request: The proposed amendments would 
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 2 and 3, 
Technical Specifications (TSs) and supporting TS Bases relating to the 
electrical distribution system. The changes are necessary to 
accommodate the installation of a new safety-related emergency diesel 
generator (EDG) and a non-safety EDG. The non-safety EDG will be used 
as an alternate air conditioning source of power in case of a station 
blackout. In addition to reflecting the new plant configuration, the 
proposed TSs also reflect the upgraded electrical capacities of the 
existing EDGs, increased fuel oil storage, and fire protection system 
for the new EDG building.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The Engineered Safety Features (ESF) electrical system provides 
a reliable source of electrical power to the 4.16 kV ESF busses to 
operate the necessary accident mitigation equipment, should offsite 
power be lost. The proposed change to Units 1 and 2 Technical 
Specifications was prompted by two significant modifications to this 
system - the addition of No. 1A Emergency Diesel Generator (EDG) and 
the upgrade of the electrical capacity of two of the three existing 
Fairbanks Morse EDGs. The addition of No. 1A EDG provides the plant 
with an ESF electrical system configuration consisting of two EDGs 
dedicated to each unit, thereby eliminating reliance upon a 
``swing'' diesel capable of being aligned to either unit. The four-
EDG configuration provides a greater degree of flexibility when an 
EDG is being overhauled or tested during refueling outages. The 
increased electrical capacity of the existing Fairbanks Morse EDGs 
will give the operators greater flexibility in the choice of 
discretionary loads for the mitigation of accidents. Both 
modifications necessitate changes to the Technical Specifications.
    The ESF electrical system, including the four EDGs, is used to 
mitigate the consequences of an accident. The design of the new No. 
1A EDG is such that incorporation of this EDG into the existing ESF 
electrical system does not result in this system becoming an 
accident initiator. Furthermore, the modification to upgrade the 
capacity of the existing EDGs will enhance the plant operators' 
ability to mitigate accidents by allowing greater flexibility in the 
choice of discretionary loads, but will not change the configuration 
of the ESF electrical system or any support systems such that the 
EDGs would become an accident initiator. Therefore, the proposed 
change would not increase the probability of an accident previously 
evaluated.
    The addition of the safety-related No. 1A EDG to the ESF 
electrical system will enhance the ability to provide reliable 
electric power during all modes of operation and shutdown conditions 
of the plant. Number 1A EDG and its support systems are designed 
such that failure of a single component will not prevent the 
capability to safely shut down the plant and to maintain the plant 
in a safe shutdown condition. Furthermore, non-safety-related 
systems associated with No. 1A EDG are designed so that their 
failure will not result in the loss of function of any safety-
related system. The design of the Fire Protection System in the 
Diesel Generator Building meets the Codes and Standards specified in 
the mechanical and instrumentation and controls design reports, 
previously approved by the 

[[Page 176]]
Commission. Inclusion of components from these systems into the 
Technical Specifications is consistent with Calvert Cliff's current 
licensing basis. The proposed Technical Specifications will 
demonstrate the reliability and capability of No. 1A EDG and the 
upgraded Fairbanks Morse EDGs to perform their accident mitigation 
function. Implementation of the proposed Technical Specifications 
will not reduce the ability of the EDGs to perform their safety 
functions. The increased volume of fuel oil necessary to support 
operation of No. 1A EDG and the upgraded Fairbanks Morse EDGs will 
not adversely impact the ability of any systems to perform their 
safety functions. The auxiliary systems which required modification 
or analysis to support the upgraded ratings of the Fairbanks Morse 
EDGs will not adversely impact operation of any other plant systems 
necessary to mitigate the consequences of an accident. Based on the 
above, the proposed change would not increase the consequences of an 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change adds Surveillance Requirements, Limiting 
Conditions for Operation, and Action Statements to reflect the 
addition of a new EDG to the ESF electrical system, and upgrades the 
electrical capacity of the existing Fairbanks Morse EDGs. This 
change does not add any new equipment, modify any interfaces with 
any existing equipment, or change the equipment's function, or the 
method of operating the equipment to be modified. The system will 
continue to operate in the same manner as before the capacity 
upgrades were implemented. The additional fuel oil required to 
support the capacity upgrades will be stored in the existing Seismic 
Category I fuel oil storage tanks. The modified EDGs will continue 
to serve a function as accident mitigators, and will not become an 
initiator of any accident.
    The NRC has reviewed the design of the new EDG, its attendant 
support systems and the new EDG Building, and concurs with Baltimore 
Gas and Electric Company's determination that the design satisfies 
the design requirements for a safety-related EDG. Number 1A EDG is a 
tandem engine-single generator set, and is physically very different 
from the existing single engine-generator Fairbanks Morse EDGs. 
However, the 4.16 kV three-phase rated electrical output is the same 
as that provided by the Fairbanks Morse EDGs to the other ESF 
busses. The excess capacity of No. 1A EDG will allow the operators 
greater flexibility in choosing post-accident discretionary loads, 
but will not cause any detrimental effects to the ESF busses or the 
equipment served by those busses. Operation of No. 1A EDG in 
accordance with these proposed Technical Specifications will not 
jeopardize the operation of any other plant systems. The design of 
the Fire Protection System in the Diesel Generator Building meets 
the Codes and Standards specified in the mechanical, and 
instrumentation and controls design reports, previously approved by 
the Commission. Inclusion of components from these systems into the 
Technical Specifications is consistent with Calvert Cliffs current 
licensing basis. Furthermore, locating No. 1A EDG and its fuel oil 
supply in a separate Category I building provides additional 
assurance that this equipment will not become an initiator of any 
accident.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The safety function of the EDGs and the ESF electrical system is 
to provide a reliable source of electrical power to the safety-
related busses to operate the necessary accident mitigation 
equipment, should offsite power be lost. The margin of safety 
associated with this safety function is two-fold: (1) a level of 
redundancy must be designed into the EDGs and the ESF electrical 
system such that the single failure criteria is met; and (2) the 
power supplied to the ESF electrical system by the EDGs must be 
sufficient to power the necessary accident mitigation equipment, 
should offsite power be lost.
    The addition of No. 1A EDG provides the plant with an ESF 
electrical system configuration consisting of two EDGs dedicated to 
each unit, thereby eliminating reliance upon a swing diesel capable 
of being aligned to either unit. In the current configuration, the 
facility meets the single failure criteria on a ``per site'' basis. 
However, as a result of the new four-EDG configuration, each unit 
will have redundant diesel generators to supply power to redundant 
safety-related equipment required for safe shutdown or accident 
mitigation. The revised Fuel Oil System configuration and the 
minimum fuel oil volume to be maintained in the fuel oil tanks 
supports the safety function of the EDGs, while maintaining the 
margin of safety associated with this equipment. Altogether, the new 
four-EDG configuration may be considered an increase in the margin 
of safety.
    Inclusion of Surveillances for the Fire Protection System 
components into the Technical Specifications is consistent with 
Calvert Cliffs current licensing basis, and ensures that adequate 
fire detection and suppression capability is available to identify 
and extinguish fires in the Diesel Generator Building, thereby 
reducing the potential for damage to No. 1A EDG and its auxiliaries. 
The Diesel Generator Building and its Fire Protection System is 
designed so that smoke and heat from a fire in that building will 
not impact the redundant safety-related Emergency Diesel Generator 
in the Auxiliary Building.
    At the completion of the modifications to increase the 
capacities of the Unit 2 EDGs and to install the new No. 1A EDG, we 
will have diesel generators with more available margin than 
currently exists. This will provide the operators with more 
flexibility during conditions where the diesel generators are 
providing onsite power. The higher electrical capacities will result 
in an increase in the margin between the EDGs' electrical capacities 
and the electrical power required to operate safety-related 
equipment required for safe shutdown or accident mitigation. 
Therefore, these modifications may be considered an increase in the 
margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendments request: November 30, 1995
    Description of amendments request: The proposed amendments would 
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, 
Technical Specifications (TSs) to allow the installation of tube 
sleeves as an alternative to plugging for repairing steam generator 
(SG) tubes. The proposed changes to TS 3/4.4.5, ``Steam Generators,'' 
and their supporting Bases would permit tube sleeving repair techniques 
developed by Westinghouse Electric Corporation and ABB Combustion 
Engineering, Inc., to be used as a repair method for the SGs at the 
Calvert Cliffs site.
    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 would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The implementation of the proposed steam generator tube sleeving 
has been reviewed for impact on the current CCNPP [Calvert Cliffs 
Nuclear Power Plant] licensing basis.
    Since the sleeve dimensions, materials, and connecting joints to 
the existing tube are designed to the applicable American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, the 
proposed sleeving 

[[Page 177]]
repair acts as an in-kind substitution for the original steam generator 
tubing. The applicable design criteria for the sleeves conform to 
the stress limits and margins of safety of Section III of the ASME 
Code. Safety factors of 3 for normal operation and 1.5 for accident 
conditions were applied to the design. Mechanical testing using the 
ASME Code stress allowables has been performed in support of the 
design. Based on the results of Westinghouse and ABB-Combustion 
Engineering analytical and test programs, the sleeves fulfill their 
intended function as leak tight structural members and meet or 
exceed all design criteria.
    Evaluation of the proposed sleeved tubes indicates no 
detrimental effects on the sleeve or sleeve-tube assembly from 
reactor system flow, primary or secondary coolant chemistries, 
thermal conditions or transients, or pressure conditions or 
transients as may be experienced at CCNPP. Corrosion testing of 
sleeve-tube assemblies indicate no evidence of sleeve or tube 
corrosion considered detrimental under anticipated service 
conditions.
    The installation of the proposed sleeves is controlled via the 
sleeving vendors' proprietary processes and equipment. The ABB 
Combustion Engineering process has been in use since 1984, and has 
been implemented 24 times for the installation of over 4,200 
sleeves. The Westinghouse process has been in use since 1988, and 
approximately 12,000 laser welded sleeves have been installed 
between 1988 and 1994. The CCNPP steam generator design was reviewed 
and found to be compatible with both installation processes and 
equipment.
    The implementation of the proposed sleeves has no significant 
effect on either the configuration of the plant, or the manner in 
which it is operated. The hypothetical consequences of failure of 
the sleeved tube is bounded by the current steam generator tube 
rupture analysis described in Section 14.15 of the Calvert Cliffs 
Updated Final Safety Analysis Report.
    Therefore, BGE [Baltimore Gas and Electric] has concluded that 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) [The proposed amendment] would not create the possibility of 
a new or different kind of accident from any other accident 
previously evaluated.
    As discussed above, the structural integrity, thermal 
characteristics, and material properties of the proposed sleeves are 
consistent with the existing plant steam generators. Therefore, the 
functions of the steam generators will not be significantly affected 
by the installation of the proposed sleeves. In addition, the 
proposed sleeves do not interact with any other plant systems. The 
continued integrity of the installed sleeve is periodically verified 
by the Technical Specification requirements. The implementation of 
the proposed sleeves has no significant effect on either the 
configuration of the plant, or the manner in which it is operated.
    Therefore, BGE concludes that this proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    (3) [The proposed amendment] would not involve a significant 
reduction in a margin of safety.
    The repair of degraded steam generator tubes via the use of the 
proposed sleeves has been confirmed to restore the structural 
integrity of the faulted tube under normal operating and postulated 
accident conditions. The design safety factors utilized for the 
sleeves are consistent with the safety factors in the ASME Boiler 
and Pressure Vessel Code used in the original steam generator 
design. The repair limit for the proposed sleeves is consistent with 
that established for the steam generator tubes. The design of the 
sleeve to tube joints is verified by testing to preclude significant 
leakage during normal and postulated accident conditions. Use of the 
previously identified design criteria and design verification 
testing assures that the margin to safety with respect to the 
implementation of the proposed sleeves is not significantly 
different from the original steam generator tubes.
    Therefore, BGE concludes that the proposed change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendments request: December 7, 1995
    Description of amendments request: The proposed amendments would 
change the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, 
Technical Specifications (TSs) by adding an analysis technique to the 
list of approved core operating limits analytical methods. 
Specifically, these amendments would add the convolution analysis 
technique to the list of approved methodologies in TSs 6.9.1.9.b. The 
convolution analysis technique has already been reviewed and approved 
by the NRC staff and the supporting safety evaluation was provided to 
the licensee by an NRC letter dated May 11, 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:
    The change has been evaluated against the standards in 10 CFR 
50.92 and has been determined to not involve a significant hazards 
consideration in that operation of the facility in accordance with 
the proposed amendment:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change is to add the convolution analysis technique 
previously approved by the NRC to the list of approved methodologies 
in Calvert Cliffs' Unit 1 and 2 Technical Specifications. By letter 
dated November 1, 1994, Baltimore Gas and Electric Company (BGE) 
requested approval to use the ABB/Combustion Engineering (ABB/CE) 
convolution technique for determining the values in the Calvert 
Cliffs Core Operating Limits Report (COLR) related to the pre-trip 
main steam line break event. Approval was given by the NRC in their 
letter dated May 11, 1995. The addition of this technique to the 
list of approved analytical methods in Technical Specification 
6.9.1.9.b is simply intended to identify it as an approved 
methodology. Therefore, the change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change is to add the convolution analysis technique 
previously approved by the NRC to the list of approved methodologies 
in Calvert Cliffs' Unit 1 and 2 Technical Specifications. By letter 
dated November 1, 1994, BGE requested approval to use the ABB/CE 
convolution technique for determining the values in the Calvert 
Cliffs COLR related to the pre-trip main steam line break event. 
Approval was given by the NRC in their letter dated May 11, 1995. 
The addition of this technique to the list of approved analytical 
methods in Technical Specifications 6.9.1.9.b is simply intended to 
identify it as an approved methodology. Therefore, the change would 
not create the possibility of a new or different type of accident 
from any accident previously evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.
    The proposed change is to add the convolution analysis technique 
previously approved by the NRC to the list of approved methodologies 
in Calvert Cliffs' Unit 1 and 2 Technical Specifications. By letter 
dated November 1, 1994, BGE requested approval to use the ABB/CE 
convolution technique for determining the values in the Calvert 
Cliffs COLR related to the pre-trip main steam line break event. 
Approval was given by the NRC in their letter dated May 11, 1995. 
The addition of this technique to the list of approved analytical 
methods in Technical Specification 6.9.1.9.b is simply intended to 
identify it as an approved methodology. Therefore, operation of the 
facility in accordance with the proposed amendment 

[[Page 178]]
does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location:  Calvert County Library, 
Prince Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of amendment request: December 7, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 3.4.5 and 3.4.6.2 and their Bases 
to maintain voltage-based steam generator tube repair criteria for the 
tube support plate elevations beyond the current cycle of operation. 
The proposed amendment would implement a 2.0 volt repair limit to 
replace a 1.0 volt repair limit which was approved on an interim basis 
for only the current fuel cycle by License Amendment No. 184 [issued 
February 3, 1995]. The proposed amendment would also include changes in 
addition to those incorporated by License Amendment No. 184 to reflect 
the guidance provided in NRC Generic Letter (GL) 95-05, ``Voltage-Based 
Repair Criteria for Westinghouse Steam Generator Tubes Affected by 
Outside Diameter Stress Corrosion Cracking.''
    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?
    Tube burst criteria are inherently satisfied during normal 
operating conditions due to the proximity of the tube support plate 
(TSP). Test data indicates that tube burst cannot occur within the 
TSP, even for tubes which have 100% throughwall electric discharge 
machining notches, 0.75 inch long, provided that the TSP is adjacent 
to the notched area. Since tube-to-TSP proximity precludes tube 
burst during normal operating conditions, use of the criteria must 
retain tube integrity characteristics which maintain a margin of 
safety of 1.43 times the bounding faulted condition, main steamline 
break (MSLB) pressure differential. As previously stated, the 
Regulatory Guide (RG) 1.121 criterion requiring maintenance of a 
safety factor of 1.43 times the MSLB pressure differential on tube 
burst is satisfied by 7/8'' diameter tubing with bobbin coil 
indications with signal amplitudes less than 8.82 volts, regardless 
of the indicated depth measurement.
    The upper voltage repair limit (Vurl) will be determined 
prior to each outage using the most recently approved NRC database 
to determine the tube structural limit (Vsl). The structural 
limit is reduced by allowances for nondestructive examination (NDE) 
uncertainty (Vnde) and growth (Vgr) to establish 
Vurl. Using Generic Letter (GL) 95-05 and growth allowances for 
an example, the NDE uncertainty component of 20% and a voltage 
growth allowance of 30% per full power year can be utilized to 
establish a Vurl of 5.9 volts. The 20% NDE uncertainty 
represents a square-root-sum-of-the-squares (SRSS) combination of 
probe wear uncertainty and analyst variability. The degradation 
growth allowance should be an average growth rate or 30% per 
effective full power year, whichever is larger. This growth 
allowance is conservative for BVPS-1 [Beaver Valley Power Station, 
Unit No. 1] as the percent voltage growth rates have decreased for 
each of the last three inspections.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated MSLB 
outside of containment but upstream of the main steam isolation 
valve (MSIV) represents the most limiting radiological condition 
relative to the plugging criteria. In support of implementation of 
the revised plugging limit, analyses will be performed to determine 
whether the distribution of cracking indications at the tube support 
plate intersections during future cycles are projected to be such 
that primary-to-secondary leakage would result in postulated site 
boundary and control room doses exceeding 10 CFR 100, and 10 CFR 50, 
Appendix A, GDC-19 requirements, respectively. A separate 
calculation has determined the maximum allowable MSLB leakage limit 
in a faulted loop. This limit was calculated using the technical 
specification reactor coolant system (RCS) Iodine-131 activity level 
of 1.0 microcuries per gram dose equivalent Iodine-131 and the 
recommended Iodine-131 transient spiking values consistent with 
NUREG-0800. The projected MSLB leakage rate calculation methodology 
prescribed in Section 2.b of GL 95-05 will be used to calculate the 
end-of-cycle (EOC) leakage. Projected EOC voltage distribution will 
be developed using the most recent EOC eddy current results and 
considering an appropriate voltage measurement uncertainty. The log-
logistic probability of leakage correlation will be used to 
establish the MSLB leakrate used for comparison with the faulted 
loop allowable limit. Due to the relatively low voltage levels of 
indications at BVPS-1 and low voltage growth rates, it is expected 
that the calculated leakage values will not exceed this limit. 
Therefore, as implementation of the 2.0 volt voltage-based plugging 
criteria at BVPS-1 does not adversely affect steam generator tube 
integrity and implementation will be shown to result in acceptable 
dose consequences, the proposed amendment does not result in any 
increase in the probability or consequences of an accident 
previously evaluated in the UFSAR [Updated Final Safety Analysis 
Report].
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Implementation of the proposed steam generator tube 2.0 volt 
plugging limit does not introduce any significant changes to the 
plant design basis. Use of the 2.0 volt plugging limit does not 
provide a mechanism which could result in an accident outside of the 
region of the tube support plate elevations as no outside diameter 
stress corrosion cracking (ODSCC) is occurring outside the thickness 
of the tube support plates. Neither a single or multiple tube 
rupture event would be expected in a steam generator in which the 
plugging limit has been applied (during all plant conditions).
    Duquesne Light Company will continue to implement a maximum 
primary-to-secondary leakage rate limit of 150 gpd [gallons per day] 
per steam generator to help preclude the potential for excessive 
leakage during all plant conditions. The RG 1.121 criterion for 
establishing operational leakage rate limits that require plant 
shutdown are based upon leak-before-break considerations to detect a 
free span crack before potential tube rupture during faulted plant 
conditions. The 150 gpd limit provides for leakage detection and 
plant shutdown in the event of the occurrence of an unexpected 
single crack resulting in leakage that is associated with the 
longest permissible crack length. RG 1.121 acceptance criteria for 
establishing operating leakage limits are based on leak-before-break 
considerations such that plant shutdown is initiated if the leakage 
associated with the longest permissible crack is exceeded.
    The single through-wall crack lengths that result in tube burst 
at 1.43 times the MSLB pressure differential and the MSLB pressure 
differential alone are approximately 0.57 inch and 0.84 inch, 
respectively. A leak rate of 150 gpd will provide for detection of 
0.41 inch long cracks at nominal leak rates and 0.62 inch long 
cracks at the lower 95% confidence level leak rates. Since tube 
burst is precluded during normal operation due to the proximity of 
the TSP to the tube and the potential exists for the crevice to 
become uncovered during MSLB conditions, the leakage from the 
maximum permissible crack must preclude tube burst at MSLB 
conditions. Thus, the 150 gpd limit provides for plant shutdown 
prior to reaching critical crack lengths for MSLB conditions using 
the lower 95% leakrate data. Additionally, this leak-before-break 
evaluation assumes that the entire crevice area is uncovered during 
blowdown. Partial uncovery will provide benefit to the burst 
capacity of the intersection. Analyses have shown that only a small 
percentage of the TSPs are deflected greater than the TSP thickness 
during a postulated MSLB.
    As steam generator tube integrity upon implementation of the 2.0 
volt plugging limit 

[[Page 179]]
continues to be maintained through inservice inspection and primary-to-
secondary leakage monitoring, 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 use of the voltage-based bobbin probe tube support plate 
elevation plugging criteria at BVPS-1 maintains steam generator tube 
integrity commensurate with the criteria of RG 1.121. This guide 
describes a method acceptable to the Commission for meeting GDCs 
[General Design Criterion] 14, 15, 30, 31, and 32 by reducing the 
probability or the consequences of steam generator tube rupture. 
This is accomplished by determining the limiting conditions of 
degradation of steam generator tubing, as established by inservice 
inspection, for which tubes with unacceptable cracking should be 
removed from service. Upon implementation of the proposed criteria, 
even under the worst case conditions, the occurrence of ODSCC 
[Outside Diameter Stress Corrosion Cracking] at the tube support 
plate elevations is not expected to lead to a steam generator 
tuberupture event during normal or faulted plant conditions. The EOC 
distribution of crack indications at the tube support plate 
elevations will be confirmed to result in acceptable primary-to-
secondary leakage during all plant conditions and that radiological 
consequences are not adversely impacted.
    In addressing the combined effects of loss-of-coolant-accident 
(LOCA) + safe shutdown earthquake (SEE) on the steam generator 
component (as required by GDC 2), it has been determined that tube 
collapse may occur in the steam generators at some plants. This is 
the case as the tube support plates may become deformed as a result 
of lateral loads at the wedge supports at the periphery of the plate 
due to the combined effects of the LOCA rarefaction wave and SSE 
loadings. Then, the resulting pressure differential on the deformed 
tubes may cause some of the tubes to collapse. There are two issues 
associated with steam generator tube collapse. First, the collapse 
of steam generator tubing reduces the RCS [reactor coolant system] 
flow area through the tubes. The reduction in flow area increases 
the resistance to flow of steam from the core during a LOCA which, 
in turn, may potentially increase peak clad temperature. Second, 
there is a potential that partial through-wall cracks in tubes could 
progress to complete through-wall cracks during tube deformation or 
collapse.
    The results of an analysis using the larger break inputs show 
that the LOCA loads were found to be of insufficient magnitude to 
result in steam generator tube collapse or significant deformation. 
Since the leak-before-break methodology is applicable to BVPS-1 
reactor coolant loop piping, the probability of breaks in the 
primary loop piping is sufficiently low that they need not be 
considered in the structural design of the plant. The limiting LOCA 
event becomes either the accumulator line break or the pressurizer 
surge line break. Analysis results provided in WCAP-14122, dated 
July 1994, demonstrate that no tubes were subject to deformation or 
collapse. No tubes have been excluded from application of the 
subject voltage-based steam generator plugging criteria.
    Addressing RG 1.83 considerations, implementation of the bobbin 
probe voltage-based tube plugging criteria of 2.0 volts is 
supplemented by: enhanced eddy current inspection guidelines to 
provide consistency in voltage normalization, a 100% eddy current 
inspection sample size at the tube support plate elevations, and 
rotating pancake coil inspection requirements for the larger 
indications left inservice to characterize the principal degradation 
as ODSCC.
    As noted previously, implementation of the tube support plate 
intersection voltage-based plugging criteria will decrease the 
number of tubes which must be repaired. The installation of steam 
generator tube plugs reduces the RCS flow margin. Thus, the 
implementation of the 2.0 volt plugging limit will maintain the 
margin of flow that would otherwise be reduced in the event of 
increased tube plugging.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the UFSAR or any 
BASES of the plant 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

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: December 15, 1995
    Description of amendment request: The proposed amendments would (1) 
revise Technical Specifications (TSs) 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, 
3/4.6.1.6, and associated Bases, (2) delete TS 6.9.2.g, and (3) add a 
new TS 6.17. The proposed changes would make the TSs consistent with 
Option B of recently revised Appendix J of 10 CFR Part 50 and the 
implementing guidance of Regulatory Guide 1.163, ``Performance-Based 
Containment Leak Test Program,'' dated September 1995. Option B of 
Appendix J permits licensees to implement a performance based option 
rather than the previous prescriptive requirements now contained in 
Appendix J as Option A. The proposed amendments would remove from the 
TSs the prescriptive requirements of Option A concerning test 
frequencies and test methodology and would also include minor 
administrative and editorial changes to add consistency between the 
Bases and the TSs and to provide additional clarification.
    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?
    Containment leakage is not an accident initiator. The proposed 
amendment does not add or modify any existing plant equipment. 
Therefore there is no increase in the probability of an accident 
previously evaluated.
    The consequences of an accident previously evaluated are not 
significantly increased. The proposed changes do not affect the 
assumptions, parameters or result of any Updated Final Safety 
Analysis (UFSAR) accident analyses. The containment leakage rate 
will continue to be maintained within the limit assumed in the 
accident analysis for a Design Basis Accident (DBA). The proposed 
changes do not modify the response of the containment during a DBA. 
The proposed amendment will continue to ensure that the ability of 
the containment structure, including the containment air locks, to 
limit leakage from a DBA is demonstrated using test methodologies 
and guidance on test frequencies that have been determined to be 
acceptable to meet the requirements of 10 CFR 50, Appendix J, Option 
B.
    The potential increase to overall accident risk due to the 
containment leak tightness decreasing between extended testing 
intervals and the resulting potential increased radioactivity 
release to the environment during a DBA has been determined to be 
minimal based on the findings of NUREG 1493 titled ``Performance-
Based Containment Leak-Test Program.'' In addition, due to the 
performance based nature of 10 CFR 50 Appendix J, Option B, the 
extended test intervals are utilized only when the component(s) have 
demonstrated an acceptable performance history. Therefore, a 
significant decrease in containment leak tightness between extended 
test intervals is not expected as a result of this proposed change.
    Based on the above discussion, it is concluded that this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical changes to the 
plant or changes in 

[[Page 180]]
plant operating configuration. The proposed amendment involves changes 
to plant programs and administrative requirements used in 
determining acceptable containment performance. The performance of 
plant systems, including the containment structure, during plant 
operation remains unchanged.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety is not significantly reduced by this 
proposed change. The acceptance criteria for ``as left'' measured 
containment leakage rates is not being increased as result of this 
proposed amendment. For Beaver Valley Power Station (BVPS) Unit No. 
1 only, the ``as found'' maximum allowable overall Type A leakage 
rate is being slightly increased. However, the slight increase does 
not exceed the value assumed in accident analysis for containment 
leakage during a DBA due to changing the acceptance criteria from 
less than to less than or equal to. The margin between the 
acceptable ``as left'' measured overall Type A containment leakage 
rate and the leakage rate assumed in the accident analysis is not 
being decreased.
    The maximum ``as found'' allowable overall Type A leakage rate 
remains unchanged for BVPS Unit No. 2. The margin between the 
acceptable ``as left'' measured overall Type A containment leakage 
rate and the leakage rate assumed in the accident analysis is also 
not being decreased.
    The maximum allowable measured combined Type B and C leakage 
rate is not being increased above the current limits.
    The maximum peak containment pressure following a DBA remains 
unchanged. The containment depressurization time following a DBA 
remains unchanged. The calculated offsite dose consequences of a DBA 
remains unchanged.
    The proposed amendment continues to ensure reactor containment 
system reliability by periodic testing in compliance with 10 CFR 50, 
Appendix J, Option B. The extension of Type A, B and C test 
frequencies permitted by 10 CFR 50 Appendix J, Option B, is not 
expected to result in a significant decrease in containment leak 
tightness between test intervals. Due to the performance based 
nature of 10 CFR 50 Appendix J, Option B, the extended test 
intervals are utilized only when the component(s) have demonstrated 
an acceptable performance history. Therefore, a significant decrease 
in containment leak tightness between extended test intervals is not 
expected as a result of this proposed change.
    The changes which are either administrative or editorial in 
nature will not reduce the margin of safety because they have no 
impact on any safety analysis assumptions.

    Therefore, based on the above discussion, it can be concluded 
that the proposed change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 19, 1995, as supplemented by letter 
dated December 7, 1995.
    Description of amendment request: May 19, 1995, submittal requested 
to modify Action Statement for Technical Specification (TS) 3.6.4.2 for 
the hydrogen recombiners. It also requested to make the surveillance 
requirements for hydrogen recombiners consistent with NUREG-1432, 
``Standard Technical Specifications Combustion Engineering Plants.'' 
The December 7, 1995, letter withdrew the request to change the Action 
Statement for TS 3.6.4.2.
    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 response is predicated on the following technical bases: (1) 
the current licensing basis of record establishes that only one 
recombiner system is required to maintain hydrogen concentration 
below 4%, (2) the proposed technical specification changes are 
conservative when compared with the recommendations of Regulatory 
Guide 1.7, (3) short term post LOCA hydrogen generation is less than 
1%, (4) long term post LOCA hydrogen generation is less than the 
flame propagation limit, which according to Regulatory Guide 1.7 
would not result in adverse effects to containment systems, and (5) 
a design basis LOCA without long term hydrogen control would produce 
pressures below the containment design pressure.... Therefore, the 
proposed change will not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    The proposed change will not alter the configuration or 
operation of any other plant system or component. The change does 
not involve any change to the operational design or limits of any 
other plant systems or components. Thus, no new failure modes are 
introduced or associated with the proposed change. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change will have no adverse impact on the 
protective boundaries, safety limits, or margin or safety. There are 
no limits or margins of safety being revised for any systems, 
components, or protective boundaries.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 7, 1995
    Description of amendment request: Amendment to Technical 
Specification (TS) 3/4.8.1 ``Electrical Power Systems - AC Sources'' 
and the associated TS BASES. The proposed amendment would implement 
selected changes from NUREG 1432, ``Standard Technical Specifications 
Combustion Engineering Plants,'' Generic Letter (GL) 94-01, ``Removal 
of Accelerated Testing and Special Reporting Requirements for Emergency 
Diesel Generators,'' and GL 93-05, ``Line-Item Technical Specifications 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operation.'' The intent of these changes is to increase Emergency 
Diesel Generator (EDG) reliability by reducing the stresses on the EDGs 
caused by unnecessary testing. This proposed TS amendment will also 
relocate the Surveillance Requirements for maintaining the properties 
of the fuel oil to TS Section 6, ``Administrative Controls.'' These 
requirements will be implemented as part of the Fuel Oil Testing 
Program. In addition, the requirement for cleaning the diesel fuel oil 
storage tanks with a sodium hypochlorite solution or equivalent will be 
changed to also allow an appropriate mechanical method (such as 
pressure washing or manual wiping) to be utilized.
    Basis for proposed no significant hazards consideration 
determination: 

[[Page 181]]
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 Standby Diesel Generators do not initiate any accidents, 
therefore the proposed changes do not increase the probability of an 
accident previously evaluated. The proposed changes to TS 3/4.8.1 
and the associated BASES affect the required actions in response to 
inoperable offsite and onsite AC sources, Surveillance Requirements 
for the EDG, and reporting requirements for EDG failures. The 
majority of the proposed changes are based on the recommendations of 
NUREG 1432, GL 94-01, and GL 93-05. These proposed changes have been 
extensively reviewed by the NRC during the preparation of these 
documents and by Waterford 3 SES during the development of this 
request for TS amendment. The proposed changes are expected to 
result in improvements in EDG performance and reduce EDG aging due 
to excessive testing. The proposed changes will permit the 
elimination of the unnecessary mechanical stress and wear on the 
EDGs while ensuring that the EDGs will perform their design 
function. The elimination of mechanical stress and wear will improve 
reliability and availability of the EDGs which will have a positive 
effect on the ability of the EDGs to perform their design function. 
The proposed changes do not affect the availability or the testing 
requirements of the offsite circuits.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed changes to TS 3/4.8.1 and the associated Bases do 
not introduce any new modes of plant operation or new accident 
precursors, involve any physical alterations to plant 
configurations, or make any changes to system setpoints which could 
initiate a new or different kind of accident. The proposed changes 
do not affect the design or performance characteristics of any EDG 
or its ability to perform its design function. No new failure modes 
have been defined and no new system interactions have been 
introduced for any plant system or component. In addition, there 
have not been any new limiting failures identified as a result of 
the proposed changes. The proposed changes will eliminate 
unnecessary EDG testing and will increase EDG reliability and 
availability. This will have an overall positive affect on plant 
safety. Accidents concerning loss of offsite power and a single 
failure (e.g., loss of an EDG) have previously been evaluated. These 
changes are intended to improve plant safety, decrease equipment 
degradation, and remove an unnecessary burden on personnel resources 
by reducing the amount of testing that the TS requires during power 
operation.
    Relocating the diesel fuel oil testing requirements to the 
Waterford 3 Fuel Oil Testing Program outside of the Technical 
Specifications is an administrative change only and consequently has 
no effect on accident probability, consequences, or margin. Also, 
the proposed cleaning method for the diesel fuel oil storage tanks 
meets the intent of Regulatory Guide 1.137 and will not result in 
the degradation of the fuel oil.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Under the proposed changes to TS 3/4.8.1 and the associated 
Bases, the EDGs will remain capable of performing their safety 
function. The changes do not affect the design or performance of the 
EDGs, but will increase EDG reliability and availability by reducing 
the stresses and the effects of aging on the EDG by eliminating 
unnecessary testing. This will result in an overall increase in 
plant safety. The ability of the EDGs to perform their safety 
function will not be degraded. Relocating the diesel fuel oil 
testing requirements to the Waterford 3 Fuel Oil Testing Program 
outside of the Technical Specifications is an administrative change 
only and consequently has no effect on accident probability, 
consequences, or margin. Also, the proposed cleaning method for the 
diesel fuel oil storage tanks meets the intent of Regulatory Guide 
1.137 and will not result in a reduction in the margin of safety.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: (TS 93-09) December 8, 1995
    Description of amendment request: The proposed change would revise 
the setpoints and time delays for the auxiliary feedwater loss-of-power 
and 6.9-kv shutdown board loss-of-voltage and degraded-voltage 
instrumentation setpoints in Items 6 and 7 of Technical Specification 
Table 3.3-4, respectively.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). 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.
    The proposed revision supports the implementation of design 
logic and setpoint changes to the loss-of-power relaying. This 
relaying is designed to ensure adequate voltage is available to 
safety-related loads in order to enhance their operability and 
support accident mitigation functions and to provide for auxiliary 
feedwater (AFW) pump starts. The design changes alter relay logic 
and delete unnecessary relaying, but do not change the diesel 
generator (D/G) start and load-shedding actuations that result from 
loss-of-power conditions. Therefore, no new actuations or functions 
have been created; and because the existing and proposed functions 
provide for accident mitigation considerations that are not the 
source of an accident, the probability of an accident is not 
increased. The deletion of the 6.9-kilovolt shutdown board normal-
feedwater undervoltage relays actually reduces the potential for 
inadvertent shutdown board blackouts as a result of short-duration 
voltage transients or instrument failures.
    The setpoints and time delays for loss-of-power functions have 
been modified based on the guidelines developed by the Electrical 
Distribution System Clearinghouse as evaluated and determined 
through detailed analysis by TVA. This design is documented in TVA 
Calculations SQN-EEB-MS-T106-0008, 27DAT, and DS-1-2 and is 
available for NRC review at the SQN site. The assigned values are 
conservative settings that will ensure adequate voltage is supplied 
to safety-related loads for accident mitigation and safety functions 
under normal, degraded, and loss-of-offsite power voltage conditions 
with appropriate time delays to prevent damage to electrical loads 
and minimize premature or unnecessary actuations. The identification 
of loss-of-voltage conditions is enhanced by the design changes to 
ensure the timely sequencing of loads onto the D/G and the 
initiation of AFW pump starts for accident mitigation. Because there 
are no reductions in safety functions resulting from the design 
logic, setpoint and time-delay changes to the loss-of-power 
instrumentation and offsite dose levels for postulated accidents 
will not be increased, the consequences of an accident are not 
increased.
    The applicable mode addition, TS 3.0.4 exclusion deletion, and 
response time measurement clarification incorporated in the proposed 
change do not affect plant functions. These changes reflect the 
requirements that SQN has been maintaining and serve to clarify the 
requirements to provide consistency of application and easier 
understanding. The AFW footnote addition and bases revision only 
clarify operability conditions that are consistent with the plant 
design for the AFW pump and loss-of-power instrumentation. Because 
there are no changes to plant functions or operations, these 
revisions have no impact on accident probabilities or consequences.

[[Page 182]]

    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    As described above, the loss-of-power instrumentation ensures 
adequate voltage to safety-related loads by initiating D/G starts 
and load shedding and provides for AFW pump starting, but is not 
considered to be the source of an accident. Although the design 
logic, setpoint, and time-delay actuation criteria have changed, the 
output functions to various plant systems that actuate for load 
shedding and D/G starts remain the same. Therefore, actuation 
criteria have been affected, but not safety functions, and the TVA 
evaluation has confirmed that the new design enhances the ability to 
maintain adequate voltage to support safety functions. Since safety 
functions have not changed and the new loss-of-power instrumentation 
design continues to support operability of safety-related equipment, 
no new or different accident is created.
    The applicable mode addition, TS 3.0.4 exclusion deletion, and 
response time measurement clarification, as well as the AFW 
operability clarifications, do not affect plant functions and will 
not create a new accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed loss-of-power TS changes support design logic, 
setpoint, and time-delay requirements that have been verified by TVA 
analysis to provide acceptable voltage levels for safety-related 
components. In determining the acceptability of these voltage 
levels, the minimum voltage for operation as well as detrimental 
component heating resulting from sustained degraded-voltage 
conditions were considered. This design ensures that safety-related 
loads will be available and operable for normal and accident plant 
conditions. The applicable mode addition, TS 3.0.4 exclusion 
deletion, response time measurement clarification, and AFW 
operability clarifications provide enhancements to TS requirements 
and do not affect plant functions. Therefore, no safety functions 
are reduced by these changes and 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 llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: (TS 95-20) December 8, 1995
    Description of amendment request: The proposed change would revise 
Surveillance Requirements 4.6.2.1.1.d and 4.6.2.1.2.b to extend the 
containment spray nozzle air or smoke flow tests from the present 5-
year interval to a 10-year interval, in accordance with 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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). 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.
    The TS change is consistent with the guidance provided in 
Generic Letter 93-05. Containment spray (CS) systems' header piping 
is stainless steel; therefore, corrosion will be negligible during 
the extended surveillance interval. Since the CS systems' headers 
are maintained dry, there is no mechanism that could cause blockage 
of the spray nozzles. Therefore, the nozzles in the CS systems will 
remain operable, during the 10-year surveillance interval, to 
mitigate the consequence of an accident previously evaluated. 
Additionally, clogging or blockage has not been observed during the 
5-year surveillance tests that have been performed in the past at 
SQN. Testing the CS systems' nozzles at the proposed reduced 
frequency will not increase the probability of occurrence of a 
postulated accident or the consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed reduced frequency testing of the CS systems' 
nozzles does not change the manner in which these systems are 
operated. The reduced testing frequency of the spray nozzles does 
not generate any new accident precursors. Therefore, the possibility 
of a new or different kind of accident previously evaluated is not 
created by the proposed changes in surveillance frequency of the CS 
system's nozzles.
    3. Involve a significant reduction in a margin of safety.
    Reduced testing of the CS systems' nozzles does not change the 
way the systems are operated or the systems' operability 
requirements. In this application, any additional corrosion of 
stainless steel piping will be negligible during the extended 
surveillance interval. Since the CS systems are maintained dry, 
there is no additional mechanism that could cause blockage of the 
nozzles. Therefore, the proposed reduced testing frequency is 
adequate to ensure spray nozzle operability. The surveillance 
requirements do not affect the margin of safety since the 
operability requirements of both the CS systems remains unchanged. 
The existing safety analysis remains bounding. 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

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

    Date of amendment request: December 8, 1995 (TS 95-24)
    Description of amendment request: The proposed change would modify 
various Technical Specification requirements in order to implement the 
recent rule change to 10 CFR Part 50, Appendix J. The new Appendix J 
rule (Option B) provides a voluntary performance based testing option 
for containment leakage rate testing (CLRT). Option B CLRT requirements 
are based on system and component performance in lieu of compliance 
with the current prescriptive requirements. Option B allows extension 
of the integrated leakage rate test (Type A test) frequency based on an 
acceptable past history. For Type B and Type C local leak rate test, 
Option B allows extension of the test frequency based on plant-specific 
experience history of each component and establishes controls to ensure 
continued performance during extended testing intervals.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria 

[[Page 183]]
established in 10 CFR 50.92(c). 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.
    The proposed amendment to SQN TSs is in accordance with Option B 
to 10 CFR 50, Appendix J. The proposed amendment adds a voluntary 
performance based option for containment leak rate testing. The 
changes being proposed do not affect the precursor for any accident 
or transient analyzed in Chapter 15 of SQN Updated Final Safety 
Analysis Report. The proposed change does not increase the total 
allowable primary containment leakage rate. The proposed change does 
not reflect a revision to the physical design and/or operation of 
the plant. Therefore, operation of the facility, in accordance with 
the proposed change, does not significantly affect the probability 
or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed amendment to SQN TSs is in accordance with the new 
performance-based option (Option B) to 10 CFR 50, Appendix J. The 
changes being proposed will not change the physical plant or the 
modes of operation defined in the facility license. The proposed 
changes do not increase the total allowable primary containment 
leakage rate. The changes do not involve the addition or 
modification of equipment, nor do they alter the design or operation 
of plant systems. Therefore, operation of the facility in accordance 
with the proposed change does 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 change to SQN TSs is in accordance with the new 
option to 10 CFR 50, Appendix J. The proposed option is formulated 
to adopt performance-based approaches. This option removes the 
current prescriptive details from the TS. The proposed changes do 
not affect plant safety analyses or change the physical design or 
operation of the plant. The proposed change does not increase the 
total allowable primary contaiment leakage rate. Therefore, 
operation of the facility, in accordance with the proposed change, 
does not involve a significant 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

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

    Date of amendment request: December 12, 1995 (TS 95-23)
    Description of amendment request: The proposed change would 
incorporate new requirements associated with steam generator tube 
inspections and repair. The new requirements would establish alternate 
steam generator tube plugging criteria at the tube support plate 
intersections.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). 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.
    Testing of model boiler specimens for free-span tubing (no tube 
support plate restraint) at room temperature conditions shows burst 
pressures in excess of 5,000 pounds per square inch (psi) for 
indications of outer diameter stress corrosion cracking with voltage 
measurements as high as 19 volts. Burst testing performed on 
intersections pulled from SQN with up to a 1.9-volt indication shows 
measured burst pressure in excess of 6,600 psi at room temperature. 
Burst testing performed on pulled tubes from other plants with up to 
7.5-volt indications shows burst pressures in excess of 5,200 psi at 
room temperatures. Correcting for the effects of temperature on 
material properties and minimum strength levels (as the burst 
testing was done at room temperature), tube burst capability 
significantly exceeds the safety-factor requirements of NRC 
Regulatory Guide (RG) 1.121.
    Tube burst criteria are inherently satisfied during normal 
operating conditions because of the proximity of the tube support 
plate (TSP). Since tube-to-tube support plate proximity precludes 
tube burst during normal operating conditions, use of the criteria 
must retain tube integrity characteristics that maintain a margin of 
safety of 1.43 times the bounding faulted condition steam line break 
(SLB) pressure differential. During a postulated SLB, the TSP has 
the potential to deflect during blowdown following a main SLB, 
thereby uncovering the TSP intersections.
    Based on the existing database, the RG 1.121 criterion requiring 
maintenance of a safety factor of 1.43 times the SLB pressure 
differential on tube burst is satisfied by 7/8-inch-diameter tubing 
with bobbin coil indications with signal amplitudes less than 8.82 
volts (WCAP-13990), regardless of the indicated depth measurement. A 
2.0-volt plugging criterion (resulting in a projected end-of-cycle 
[EOC] voltage) compares favorably with the 8.82-volt structural 
limit considering the extremely slow apparent voltage growth rates 
and few numbers of indications at SQN. Using the established 
methodology of RG 1.121, the structural limit is reduced by 
allowances for uncertainty and growth to develop a beginning of 
cycle (BOC) repair limit that would preclude indications at EOC 
conditions that exceed the structural limit. The nondestructive 
examination (NDE) uncertainty component is 20.5 percent, and is 
based on the Electric Power Research Institute (EPRI) alternate 
repair criteria (ARC).
    Test data indicates that tube burst cannot occur within the TSP, 
even for tubes that have 100 percent throughwall electro-discharge 
machining notches, 0.75 inch long, provided that the TSP is adjacent 
to the notched area. Because of the few number of indications at 
SQN, the EPRI methodology of applying a growth component of 35 
percent per effective full power year (EFPY) will be used. Near-term 
operating cycles at SQN are expected to be bounded by 1.23 years, 
therefore, a 43 percent growth component is appropriate. When these 
allowances are added to the BOC alternate plugging criteria (APC) of 
2.0 volts in a deterministic bounding EOC voltage of approximately 
3.26 volts for Cycle 7, operation can be established. A 5.56-volt 
deterministic safety margin exists (8.82 structural limit - 3.26-
volt EOC equal 5.56-volt margin).
    For the voltage/burst correlation, the EOC structural limit is 
supported by a voltage of 8.82 volts. Using this structural limit of 
8.82 volts, a BOC maximum allowable repair limit can be established 
using the guidance of RG 1.121. The BOC maximum allowable repair 
limit should not permit the existence of EOC indications that exceed 
the 8.82-volt structural limit. By adding NDE uncertainty allowances 
and an allowance for crack growth to the repair limit, the 
structural limit can be validated. Therefore, the maximum allowable 
BOC repair limit (RL) based on the structural limit of 8.82 volts 
can be represented by the expressions:
    RL + (0.205 x RL) + (0.43 x RL) = 8.82 volts, or,
    the maximum allowable BOC repair limit can be expressed as,
    RL = 8.82-volt structural limit/1.64 = 5.4 volts.
    This RL (5.4 volts) is the appropriate limit for APC 
implementation to repair bobbin indications greater than 2.0 volts 
independent of rotating pancake coil (RPC) confirmation of the 
indication. This 5.4-volt upper limit for non-confirmed RPC calls is 
consistent with other recently approved APC programs (Farley Nuclear 
Plant, Unit 2).
    The conservatism of the growth allowance used to develop the 
repair limit is shown by the most recent SQN eddy current data. Only 
seven tubes in Unit 2 required repair because of outside diameter 
stress corrosion cracking (ODSCC) at the TSP intersections.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated main 

[[Page 184]]
SLB outside of containment, but upstream of the main steam isolation 
valve (MSIV), represents the most limiting radiological condition 
relative to the APC. Implementation of the APC will determine 
whether the distribution of cracking indications at the TSP 
intersections is projected to be such that primary-to-secondary 
leakage would result in site boundary doses within a small fraction 
of the 10 CFR 100 guidelines. A separate analysis has determined 
this allowable SLB leakage limit to be 3.7 gallons per minute (gpm) 
in the faulted loop. This limit uses the TS reactor coolant system 
(RCS) Iodine-131 activity level of 1.0 microcuries per gram dose 
equivalent Iodine-131 and the recommended Iodine-131 transient 
spiking values consistent with NUREG-0800. The analysis method is 
WCAP-14277, which is consistent with the guidance of the NRC generic 
letter (GL) [95-05] and will be used to calculate EOC leakage. 
Because of the relatively low number of indications at SQN, it is 
expected that the actual leakage values will be far less than this 
limit. Additionally, the current Iodine-131 levels at SQN range from 
about 25 to 100 times less than the TS limit.
    Application of the criteria requires the projection of 
postulated SLB leakage, based on the projected EOC voltage 
distribution for Cycle 8 operation. Projected EOC voltage 
distribution is developed using the most recent EOC eddy current 
results and a voltage measurement uncertainty. Data indicates that a 
threshold voltage of 2.8 volts would result in throughwall cracks 
long enough to leak at SLB condition. The GL requires that all 
indications to which the APC are applied must be included in the 
leakage projection. Tube pull results from another plant with 7/8-
inch tubing with a substantial voltage growth database have shown 
that tube wall degradation of greater than 40 percent throughwall 
was readily detectable either by the bobbin or RPC probe. The tube 
with maximum throughwall penetration of 56 percent (42 average) had 
a voltage of 2.02 volts. The SQN Unit 1 pulled tube had a 1.93-volt 
indication with a maximum depth of 91 percent and did not leak at 
SLB condition. Based on the SQN pulled tube and industry pulled tube 
data supporting a lower threshold for SLB leakage of 2.8 volts, 
inclusion of all APC intersections in the leakage model is quite 
conservative. The ODSCC occurring at SQN is in its earliest stages 
of development. The conservative bounding growth estimations to be 
applied to the expected small number of indications for the upcoming 
inspection should result in very small levels of predicted SLB 
leakage. Historically, SQN has not identified ODSCC as a contributor 
to operational leakage.
    In order to assess the sensitivity of an indication's BOC 
voltage to EOC leakage potential, a Monte Carlo simulation was 
performed for a 2.0-volt BOC indication.
    The maximum EOC voltage (at 99.8 percent cumulative probability) 
was found to be 4.8 volts. The leakage component from an indication 
of this magnitude, using the EPRI leakage model, is 0.028 gpm.
    Therefore, as implementation of the 2.0-volt APC does not 
adversely affect steam generator (S/G) tube integrity and 
implementation will be shown to result in acceptable dose 
consequences, the proposed amendment does not result in significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Implementation of the proposed S/G tube APC does not introduce 
any significant changes to the plant design basis. Use of the 
criteria does not provide a mechanism that could result in an 
accident outside of the region of the TSP elevations; no ODSCC is 
occurring outside the thickness of the TSP. Neither a single or 
multiple tube rupture event would be expected in a S/G in which the 
plugging criteria is applied (during all plant conditions).
    TVA will implement a maximum leakage rate limit of 150 gallon 
per day per S/G to help preclude the potential for excessive leakage 
during all plant conditions. The SQN TS limits on primary-to-
secondary leakage at operating conditions include a maximum of 0.42 
gpm (600 gallons per day [gpd]) for all S/Gs, or, a maximum of 150 
gpd for any one S/G. The RG 1.121 criterion for establishing 
operational leakage rate limits that require plant shutdown is based 
upon leak-before-break considerations to detect a free-span crack 
before potential tube rupture during faulted plant conditions. The 
150-gpd limit should provide for leakage detection and plant 
shutdown in the event of the occurrence of an unexpected single 
crack resulting in leakage that is associated with the longest 
permissible crack length. RG 1.121 acceptance criteria for 
establishing operating leakage limits are based on leak-before-break 
considerations such that plant shutdown is initiated if the leakage 
associated with the longest permissible crack is exceeded. The 
longest permissible crack is the length that provides a factor of 
safety of 1.43 against bursting at faulted conditions maximum 
pressure differential. A voltage amplitude of 8.82 volts for typical 
ODSCC corresponds to meeting this tube burst requirement at a lower 
95 percent prediction limit on the burst correlation coupled with 
95/95 lower tolerance limit material properties. Alternate crack 
morphologies can correspond to 8.82 volts so that a unique crack 
length is not defined by the burst pressure versus voltage 
correlation. Consequently, typical burst pressure versus through-
wall crack length correlations are used below to define the 
``longest permissible crack'' for evaluating operating leakage 
limits.
    The single through-wall crack lengths that result in tube burst 
at 1.43 times the SLB pressure differential and the SLB pressure 
differential alone are approximately 0.57 inch and 0.84 inch, 
respectively. A leak rate of 150 gpd will provide for detection of 
0.4-inch-long cracks at nominal leak rates and 0.6-inch-long cracks 
at the lower 95 percent confidence level leak rates. Since tube 
burst is precluded during normal operation because of the proximity 
of the TSP to the tube and the potential exists for the crevice to 
become uncovered during SLB conditions, the leakage from the maximum 
permissible crack must preclude tube burst at SLB conditions. Thus, 
the 150-gpd limit provides for plant shutdown before reaching 
critical crack lengths for SL-conditions. Additionally, this leak-
before-break evaluation assumes that the entire crevice area is 
uncovered during blowdown. Partial uncover will provide benefit to 
the burst capacity of the intersection.
    As S/G tube integrity upon implementation of the 2.0-volt APC 
continues to be maintained through in-service inspection and 
primary-to-secondary leakage monitoring, the possibility of a new or 
different kind of accident from any accident previously evaluated is 
not created.
    3. Involve a significant reduction in a margin of safety.
    The use of the voltage based APC at SQN is demonstrated to 
maintain S/G tube integrity commensurate with the criteria of RG 
1.121. RG 1.121 describes a method acceptable to the NRC Staff for 
meeting General Design Criteria (GDC) 14, 15, 31, and 32 by reducing 
the probability or the consequences of S/G tube rupture. This is 
accomplished by determining the limiting conditions of degradation 
of S/G tubing, as established by in-service inspection, for which 
tubes with unacceptable cracking should be removed from service. 
Upon implementation of the criteria, even under the worst-case 
conditions, the occurrence of ODSCC at the TSP elevations is not 
expected to lead to a S/G tube rupture event during normal or 
faulted plant conditions. The EOC distribution of crack indications 
at the TSP elevations will be confirmed to result in acceptable 
primary-to-secondary leakage during all plant conditions and 
radiological consequences are not adversely impacted.
    In addressing the combined effects of loss-of-coolant accident 
(LOCA), plus safe shutdown earthquake (SSE) on the S/G component (as 
required by GDC 2), it has been determined that tube collapse may 
occur in the S/Gs at some plants. This is the case as the TSP may 
become deformed as a result of lateral loads at the wedge supports 
at the periphery of the plate because of the combined effects of the 
LOCA rarefaction wave and SSE loadings. Then, the resulting pressure 
differential on the deformed tubes may cause some of the tubes to 
collapse.
    There are two issues associated with S/G tube collapse. First, 
the collapse of S/G tubing reduces the RCS flow area through the 
tubes. The reduction in flow area increases the resistance to flow 
of steam from the core during a LOCA, which in turn, may potentially 
increase peak clad temperature (PCT). Second, there is a potential 
that partial through-wall cracks in tubes could progress to through-
wall cracks during tube deformation or collapse.
    Consequently, since the leak-before-break methodology is 
applicable to the SQN reactor coolant loop piping, the probability 
of breaks in the primary loop piping is sufficiently low that they 
need not be considered in the structural design of the plant. The 
limiting LOCA event becomes either the accumulator line break or the 
pressurizer surge line break. LOCA loads for the primary pipe breaks 
were used to bound the conditions at SQN for smaller breaks. The 
results of the analysis 

[[Page 185]]
using the larger break inputs show that the LOCA loads were found to be 
of insufficient magnitude to result in S/G tube collapse or 
significant deformation. The LOCA, plus SSE tube collapse evaluation 
performed for another plant with Series 51 S/Gs using bounding input 
conditions (large-break loadings), is applicable to SQN. Therefore, 
at SQN, no tubes will be excluded from using the voltage repair 
criteria due to deformation of collapse of S/G tubes following a 
LOCA plus an SSE. Additional supporting information relative to NRC 
review of J.M. Farley Nuclear Plant was provided in Enclosure 5, 
Item 3 of TVA's submittal dated September 7, 1995 (TAC No. M92961).
    Addressing RG 1.83 considerations, implementation of the bobbin 
probe voltage based interim tube plugging criteria of 2.0 volt is 
supplemented by: (1) enhanced eddy current inspection guidelines to 
provide consistency in voltage normalization, (2) a 100 percent eddy 
current inspection sample size at the TSP elevations, and (3) RPC 
inspection requirements for the larger indications left in service 
to characterize the principal degradation as ODSCC.
    As noted previously, implementation of the TSP elevation 
plugging criteria will decrease the number of tubes that must be 
repaired. The installation of S/G tube plugs reduces the RCS flow 
margin. Thus, implementation of the alternate plugging criteria will 
maintain the margin of flow that would otherwise be reduced in the 
event of increased tube plugging.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
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

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

    Date of amendment request: November 21, 1995
    Brief description of amendments: The proposed amendments would 
modify the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 
Technical Specifications (TS) to allow the containment personnel 
airlock (PAL) doors to remain open during movement of irradiated fuel 
and during core alterations.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change allows the PAL doors for containment to 
remain open during the movement of irradiated fuel and core 
alterations. Whether or not the PAL doors are open does not effect 
the movement of fuel, the strict compliance with the procedures 
governing refueling operations, or the integrity of fuel assemblies. 
The position of the airlock doors cannot, in itself, be the 
initiating event in any accident. The probability of a fuel handling 
accident is not changed.
    The consequences of leaving the airlock doors open during this 
accident are bounded by the existing analysis, provided the fuel 
handling accident assumptions are maintained (e.g. 100 hours after 
reactor shutdown and the water level remains 23 feet above the 
fuel). The existing analysis postulates the limiting fuel handling 
accident to occur in the Fuel Building with no credit taken for 
barrier or filtration. This accident analysis envelopes the proposed 
change for a fuel handling accident occurring in the Containment 
Building.
    Were a fuel handling accident to occur with the PAL doors open, 
the impact would be minimal. Pressure is expected to be essentially 
equalized across the door with little air flow either into or out of 
containment. Based on transport time from the location of the 
accident to the PAL, little, if any, radioactive material is 
expected to escape containment via the PAL. The amount that might 
escape would not necessarily be anymore than might escape as the 
door is cycled to evacuate personnel. What does escape will be 
filtered by the Primary Plant Ventilation System, the same as if the 
accident were to occur in the fuel building. In summary, not only is 
the accident clearly bounded by the existing analysis, the actual 
increase in release of radioactive material outside the plant will 
be insignificant if there is any measurable increase at all.
    Based on the above, allowing the PAL doors to remain open during 
movement of irradiated fuel and core alterations, has no significant 
effect on the probability or consequences of an accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different type of accident from any accident previously evaluated?
    The change does not add new hardware. The only change in the 
operation of the plant is that the PAL doors will remain open during 
movement of irradiated fuel and core alterations. Because the 
current fuel handling accident analysis considers fuel handling 
accidents in either the Fuel Building or the Containment Building, 
the current fuel handling accident analysis remains bounding for the 
proposed change. Therefore, the proposed change does not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The assumptions used to calculate the offsite dose resulting 
from a fuel handling accident in [the] Containment Building are 
equivalent to assuming that the PAL remains open for the entire 
accident and that no filtration occurs. Since no credit was taken 
for any containment barrier or ventilation system filtration, the 
dose to the public as calculated in the analysis is not affected by 
this change. Therefore, the proposed change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 21, 1995
    Brief description of amendments: The proposed amendment would 
revise the core safety limit curves and revised N-16 Overtemperature 
reactor trip setpoints as a result of the reload analyses for CPSES 
Unit 2, Cycle 3. In addition, the minimum required Reactor Coolant 
System (RCS) flow is increased and an administrative enhancement is 
included in the footnotes of the RCS flow - low reactor trip function 
setpoint for both Units.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    A. Increase in Unit 2 minimum required flow
    This revision increases the Unit 2 minimum required RCS flow 
rate assumed in 

[[Page 186]]
the safety analyses by 3.6%. The actual core flow is unchanged and is 
approximately 6.6% higher than the value assumed in previous 
accident analyses. The remaining 3.0% flow is sufficient to account 
for all uncertainties associated with the core flow measurement.
    Since this change only involves analysis methodology and does 
not affect the actual core flow, it does not increase the actual 
probability or consequences of any postulated accident.
    When considered separately, increasing the minimum required RCS 
flow is a conservative change. Although there is no impact on the 
initiation of any postulated accidents, the potential severity of 
the affected accidents is typically less when flow is increased. In 
general, the increased ability to remove heat from the fuel will 
reduce the peak temperature seen by the fuel and reduce the 
potential for undesirable boiling conditions. Thus, the increase in 
the assumed RCS flow will not increase the probability or 
consequences of an accident previously analyzed.
    B. Revision to the Unit 2 Core Safety Limits
    Analyses of reactor core safety limits are required as part of 
reload calculations for each cycle. TU Electric has performed in-
house analyses of the Unit 2, Cycle 3 core to determine the reactor 
core safety limits. The newer methodologies and safety analysis 
values result in new operating curves which, in general, permit 
plant operation over a similar range of acceptable conditions. This 
change means that if a transient were to occur with the plant 
operating at the limits of the new curve, a higher temperature and 
power level might be attained than if the plant were operating 
within the bounds of the old curves. However, since the new curves 
were developed using approved methodologies which are wholly 
consistent with and do not represent a change in the Technical 
Specification bases for safety limits, all applicable postulated 
transients will continue to be properly mitigated. As a result, 
there will be no significant increase in the consequences, as 
determined by accident analyses, of any accident previously 
evaluated.
    C. Revision to Unit 2 Overtemperature N-16 Reactor Trip 
Setpoints, Parameters and Coefficients
    As a result of changes discussed, the Overtemperature N-16 
reactor trip setpoint has been recalculated. These trip setpoints 
help ensure that the core safety limits are maintained and that all 
applicable limits of the safety analysis are met.
    Based on the calculations performed, the safety analysis value 
for Overtemperature N-16 reactor trip setpoint has changed. This 
essentially means if a transient were to occur, the actual 
temperature and power level could be slightly higher. However, the 
analyses performed show that, using the TU Electric methodologies, 
all reactor core safety limits are met and all applicable limits of 
the safety analysis are met. This parameter has a setpoint which 
allows the mitigation of postulated accidents and has no impact on 
accident initiation. Therefore, the changes in safety analysis 
values do not involve an increase in the probability of an accident 
and, based on satisfying the core safety limits and all applicable 
safety analysis limits, there is no significant increase in the 
consequences of any accident previously evaluated.
    In addition, the changes result in setpoint values which 
potentially offer safety benefits. The risk of turbine runbacks or 
reactor trips due to upper plenum flow anomalies will be minimized 
with a higher overtemperature setpoint, thus reducing potential 
challenges to the plant safety systems. A final benefit is that the 
new methods for considering N-16 setpoints and values will be 
consistent with Unit 1, which reduces the potential for personnel 
error due to unit differences.
    Considering both the safety analysis impact and the benefits 
described above, the changes in N-16 setpoints and parameters will 
result in slight reduction in the probability of an accident and do 
not significantly increase the consequences of an accident 
previously evaluated.
    D. Deletion of footnotes associated with the RCS flow - low 
reactor trip setpoint
    In lieu of revising the footnotes to support the Unit 2 Cycle 3 
operation, the deletion of the footnote is proposed. Further, for 
consistency with Unit 2, the same change is proposed for Unit 1. 
This change will not affect current plant practice; however, it will 
impose a more restrictive RCS flow - low setpoint than is currently 
required. The RCS flow - low reactor trip setpoint is currently 
specified in Technical Specification Table 2.2-1, Functional Unit 
12.b, to be 90% of the minimum measured RCS flow. The proposed 
change would require the setpoint to be 90% of the instrument span 
where 100% of instrument span approximately corresponds to the 
actual RCS flow. The actual RCS flow is verified to be greater than 
the RCS flow assumed in the accident analysis through compliance 
with Technical Specification 3.2.5. Thus, through deletion of the 
footnotes, the RCS volumetric flow corresponding to the reactor trip 
setpoint will be greater than or equal to the volumetric flow 
allowed by the current specifications.
    In summary, the proposed deletion of the footnotes will have no 
impact on current plant operations. A possible relaxation of the RCS 
flow - low setpoint which is currently allowed by the Technical 
Specifications will be removed without creating the potential for 
unnecessary plant trips.
    The RCS flow - low reactor trip setpoint can have no effect on 
the probability of an accident. Because the reactor will be tripped 
at or prior to the conditions assumed in the accident analyses, 
there will be no effect on the consequences of an accident 
previously identified.
    SUMMARY
    The changes in the amendment request applies new NRC approved 
methodologies, changes in safety analysis values, new core safety 
limits and new N-16 setpoint and parameter values to assure that all 
applicable safety analysis limits have been met. The potential for 
an operational transient to occur has been reduced and there has 
been no significant impact on the consequences of any accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve the use of revised safety analysis 
values and the calculation of new reactor core safety limits and 
reactor trip setpoints. As such, the changes play an important role 
in the analysis of postulated accidents but none of the changes 
effect plant hardware or the operation of plant systems in a way 
that could initiate an accident. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    In reviewing and approving the methods used for safety analyses 
and calculations, the NRC has approved the safety analysis limits 
which establish the margin of safety to be maintained. While the 
actual impact on safety is discussed in response to question 1, the 
impact on margin of safety is discussed below.
    A. Increase in the Unit 2 minimum required flow
    In performing the DNB-related analyses, the Reactor Coolant 
System flow rate assumed in these analyses is increased by 3.6 
percent to insure that all applicable limits of the safety analysis 
are met. The Technical Specification 3/4.2.5 limit for this 
parameter will be changed to insure that it is maintained within the 
normal steady-state envelope of operation assumed in the transient 
and accident safety analyses (i.e., ensuring that the RCS flow rate 
assumed in the safety analyses remains valid). The Technical 
Specification limits are consistent with the initial safety analysis 
assumption (plus uncertainties) and have been analytically 
demonstrated to be adequate to maintain a minimum DNBR at or above 
the safety analysis DNBR limit throughout each analyzed transient. 
Because the 95/95 DNBR acceptance criteria is met with the proposed 
change and assumptions of the safety analyses are maintained valid 
by the Technical Specification limits, there is no change in a 
margin of safety.
    B. Revision to the Unit 2 Reactor Core Safety Limits
    The TU Electric reload analysis methods have been used to 
determine new reactor core safety limits. All applicable safety 
analysis limits have been met. The methods used are wholly 
consistent with Technical Specification BASES 2.1 which is the bases 
for the safety limits. In particular, the curves assure that for 
Unit 2, Cycle 3, the calculated DNBR is no less than the safety 
analysis limit and the average enthalpy at the vessel exit is less 
than the enthalpy of saturated liquid.
    In conjunction with the reactor core safety limit methodology, 
the NRC approved TUE-1 DNB correlation is used for performing DNB-
related analyses. This correlation will be applied to the core 
configuration of CPSES Unit 2, Cycle 3 and future core 
configurations. The TUE-1 correlation DNBR limit is established such 
that there is a 95 percent probability with 95 percent confidence 
level that DNB will not occur when the minimum DNBR for the limiting 
fuel is greater than or equal to the TUE-1 correlation DNBR limit. 
This 95/95 criteria defines the ``margin of safety'' for the DNB-

[[Page 187]]
related analysis and remains valid even though the DNB correlation and 
associated correlation limit are changed. Margin is provided in the 
DNB-related analysis for known and potential effects such as 
hydraulic differences between the two co-resident fuel assembly 
designs and the presence of the Reactor Coolant System lower plenum 
flow anomaly. The TUE-1 correlation DNBR limit plus margin 
constitutes the safety analysis DNBR limit. The accident analyses 
are performed to ensure that the safety analysis DNBR limit 
acceptance criteria are satisfied. Because the 95/95 DNBR acceptance 
criteria remains valid and continues to be satisfied, no change in a 
margin of safety occurs.
    C. Revision to Unit 2 Overtemperature N-16 Reactor Trip 
Setpoints, Parameters and Coefficients
    Because the reactor core safety limits for CPSES Unit 2, Cycle 3 
are recalculated, the Reactor Trip System instrumentation setpoint 
values for the Overtemperature N-16 reactor trip setpoint which 
protect the reactor core safety limits must also be recalculated. 
The Overtemperature N-16 reactor trip setpoint helps prevent the 
core and Reactor Coolant System from exceeding their safety limits 
during normal operation and design basis anticipated operational 
occurrences. The most relevant design basis analysis in Chapter 15 
of the CPSES Final Safety Analysis Report (FSAR) which is affected 
by the change in the safety analysis value for the CPSES Unit 2 
Overtemperature N-16 reactor trip setpoint is the Uncontrolled Rod 
Cluster Control Assembly Bank Withdrawal at Power (FSAR Section 
15.4.2). This event has been re-analyzed with the revised safety 
analysis value for the Overtemperature N-16 reactor trip setpoint to 
demonstrate compliance with event specific acceptance criteria. 
Because all event acceptance criteria are satisfied, there is no 
degradation in a margin of safety.
    The nominal Reactor Trip System instrumentation setpoints values 
for the Overtemperature N-16 reactor trip setpoint (Technical 
Specification Table 2.2-1) are determined based on a statistical 
combination of all of the uncertainties in the channels to arrive at 
a total uncertainty. The total uncertainty plus additional margin is 
applied in a conservative direction to the safety analysis trip 
setpoint value to arrive at the nominal and allowable values 
presented in Technical Specification Table 2.2-1. Meeting the 
requirements of Technical Specification Table 2.2-1 assures that the 
Overtemperature N-16 reactor trip setpoint assumed in the safety 
analyses remains valid. The CPSES Unit 2, Cycle 3 Overtemperature N-
16 reactor trip setpoint is different from previous cycles which 
provides more operational flexibility to withstand mild transients 
without initiating automatic protective actions. Although the 
setpoint is different, the Reactor Trip System instrumentation 
setpoint values for the Overtemperature N-16 reactor trip setpoint 
are consistent with the safety analysis assumption which has been 
analytically demonstrated to be adequate to meet the applicable 
event acceptance criteria. Thus, there is no reduction in a margin 
of safety.
    D. Deletion of footnotes associated with the RCS flow - low 
reactor trip function
    The deletion of the footnotes, and the potential relaxation of 
the RCS flow - low setpoint which could be used, will provide 
further assurance that, in the event of a partial loss of forced RCS 
flow or locked rotor transient, a reactor trip signal would be 
initiated prior to the conditions assumed in the accident analyses. 
Thus, the accident analyses are unaffected, and there is no 
reduction in a margin of safety.
    SUMMARY
    The proposed changes to the CPSES Technical Specifications 
involve using NRC-approved licensing analysis methods developed by 
TU Electric to determine the Technical Specification reactor core 
safety limits and perform DNB-related analysis for CPSES Unit 2, 
Cycle 3. The DNB-related analyses are performed by TU Electric using 
a qualified, state-of-the-art departure from nucleate boiling (DNB) 
correlation, TUE-1, which has also been approved by the NRC for the 
CPSES Unit 2, Cycle 3 core configuration. In performing these 
analyses, the minimum required Reactor Coolant System flow rate is 
increased by 3.6 percent. Because the core safety limits for CPSES 
Unit 2, Cycle 3 are recalculated, the Reactor Trip System 
instrumentation setpoints values for the Overtemperature N-16 
reactor trip setpoint which protect the core safety limits are also 
recalculated.
    Using the NRC approved TU Electric methods, the reactor core 
safety limits are determined such that all applicable limits of the 
safety analyses are met, particularly the 95/95 DNBR limit. The 
Technical Specification 3/4.2.5 limits for the DNB Parameters insure 
the assumptions in the safety analyses remain valid. Because the 
applicable event acceptance criteriacontinue to be met, there is no 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: October 17, 1995
    Description of amendment request: The proposed amendment would 
modify the North Anna Power Station, Units 1 and 2 Technical 
Specifications (TS) to allow both of the containment personnel airlock 
doors to remain open during refueling operations, delete the license 
condition referencing the analyses for limiting doses to the control 
room operators, and modify the TS Bases to clarify the emergency power 
system requirements relative to mitigation of the consequences of a 
Fuel Handling Accident.
    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:
    There is no significant change in the probability or 
consequences of an accident previously evaluated. There are no 
system changes which would increase the probability of an accident 
occurring. Allowing both personnel airlock doors to remain open 
during core alterations or fuel movement inside containment will not 
have any impact on the probability of a Fuel Handling Accident 
either in containment or in the fuel building. The consequences of a 
Fuel Handling Accident have been investigated by performing a 
reanalysis with no credit for isolation or filtration by the Fuel 
Building or containment ventilation systems. The Exclusion Area 
Boundary [EAB] and Low Population Zone [LPZ] doses for a Fuel 
Handling Accident without credit for iodine filtration remain well 
within (<25%) of the NRC regulatory limits of 10 CFR [Part] 100. The 
predicted control room operator doses remain bounded by the limiting 
case for control room doses and within the regulatory limits of 
General Design Criterion [GDC] 19. In addition, the action to 
clarify the responses to NRC question 6.72 [of the original Final 
Safety Analysis Report] will not increase the probability or 
consequences of the Fuel Handling Accident.
    No new accident types or equipment malfunction scenarios are 
introduced as a result of the clarification to the Virginia Power 
response to [NRC question] 6.72 or as a result of these changes in 
analysis methods or the proposed Technical Specifications changes to 
allow both personnel airlock doors to remain open during core 
alterations or fuel movement inside containment. Therefore, there is 
no possibility of an accident of a different type than any 
previously evaluated in the North Anna USFAR [Updated Final Safety 
Analysis Report].
    There is no significnt reduction in the margin of safety. An 
evaluation of the Fuel Handling Accident doses at the EAB, the LPZ 
and to control room operators has been performedand it has been 
concluded that the acceptance criteria defined by GDC-19, 10 CFR 
100, and the NRC Standard Review Plan will continue to be met.
    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 

[[Page 188]]
involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23212.
    NRC Project Director: David B. Matthews

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 application for amendments: September 13, 1995, as amended 
November 27, 1995
    Brief description of amendments: The proposed amendments would 
permit the licensee to implement the performance-based option provided 
by 10 CFR Part 50, Appendix J, which allows leakage testing intervals 
to be based on system and component testing performance.
    Date of publication of individual notice in Federal Register: 
December 12, 1995 (60 FR 63739)
    Expiration date of individual notice: January 11, 1996
    Local Public Document Room location: The University of North 
Carolina at Wilmington, William Madison Randall Library, 601 S. College 
Road, Wilmington, North Carolina 28403-3297

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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

    Date of application for amendments: October 3, 1995
    Brief description of amendments: The amendments delete Sections 
2.B.(7)(a) and (b) of
    Facility Operating License No. NPF-51 (Unit 2) and Sections 
2.b.(6)(a) and (b) of
    Facility Operating License No. NPF-74 (Unit 3) relating to certain 
previous sale and leaseback transactions that were added by Amendment 
No. 3 for NPF-51 and Amendment No. 1 for NPF-74.
    Date of issuance: December 8, 1995
    Effective date: December 8, 1995
    Amendment Nos.: Unit 2 - Amendment No. 91; Unit 3 - Amendment No. 
74
    Facility Operating License Nos. NPF-51 and NPF-74: The amendments 
revised the license.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56363) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 8, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

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

    Date of application for amendments: October 23, 1995
    Brief description of amendments: The amendments revised the 
Technical Specifications to delete the applicability of the primary 
coolant water chemistry limits when the primary system is being 
chemically decontaminated and the reactor vessel is defueled.
    Date of issuance: December 13, 1995
    Effective date: December 13, 1995
    Amendment Nos.: 180 and 211
    Facility Operating License Nos. DPR-71 and DPR-62.
    Date of initial notice in Federal Register: The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated December 13, 1995. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Commonwealth Edison Company, Docket Nos. 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 application for amendments: September 14, 1995, as 
supplemented November 8, 1995.
    Brief description of amendments: The amendments allow the use of an 
alternate zirconium based fuel cladding, 

[[Page 189]]

ZIRLO, and permit limited substitution of fuel rods with ZIRLO filler 
rods. In addition, a clarification and an editorial change have been 
included.
    Date of issuance: December 19, 1995
    Effective date: December 19, 1995
    Amendment Nos.: 78 and 70
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54716) The November 8, 1995 letter, provided clarifying information 
that did not change the scope of the September 14, 1995, 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 December 19, 1995No significant 
hazards consideration comments received: No
    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.

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

    Date of application for amendments: September 15, 1995.
    Brief description of amendments: The amendments upgrade the current 
custom Technical Specifications (TS) for Dresden and Quad Cities to the 
Standard Technical Specifications contained in NUREG-0123, ``Standard 
Technical Specification General Electric Plants BWR/4.'' The 
application dated September 15, 1995, contains some of the TSUP open 
items from previous Dresden and Quad Cities TS amendments issued by the 
NRC.
    Date of issuance: December 19, 1995Effective date: Immediately, to 
be implemented no later than June 30, 1996.
    Amendment Nos.:  145, 139, 167 and 163
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 5, 1995 (60 FR 
52220) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 19, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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

    Date of amendment request: December 8, 1995
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.1.A.5 to revise the wording to allow a single 
train of Power-Operated Relief Valves (PORVs)/Block Valves to be closed 
and deenergized indefinitely. The proposed change is administrative and 
is intended to correct inconsistencies between the intended operation 
of the PORVs/Block Valves and the language of the TSs.
    Date of issuance: December 8, 1995
    Effective date: As of the date of issuance to be implemented 
immediately.
    Amendment No.: 185
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: NoThe Commission's related 
evaluation of the amendment, emergency circumstances and consultation 
with the State, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated December 8, 
1995.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: Ledyard B. Marsh

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: October 31, 1994
    Brief description of amendments: The amendments remove the stroke 
times for the steam generator power operated relief valves from 
Technical Specification Tables 3.6-2a and 3.6-2b.
    Date of issuance: December 18, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.:  Unit 1 - 139 - Unit 2 - 133
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8745) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 18, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas

    Date of amendment request: September 4, 1993, as supplemented by 
letters dated February 16, 1994, and August 4, 1995
    Brief description of amendments: The license amendments revised the 
Arkansas Nuclear One Industrial Security Plan.
    Date of issuance: December 19, 1995
    Effective date: December 19, 1995
    Amendment Nos.: 183 and 172
    Facility Operating License Nos. DPR-51 and NPF-6. Amendments 
revised the licenses.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56368) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 19, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of amendment request: December 6, 1993, as supplemented by 
letters dated May 12, August 9, and September 18, 1995.
    Brief description of amendment: The amendment changes the Appendix 
A TSs to allow installation of steam generator tube repair sleeves at 
the Waterford Steam Electric Station, Unit 3. The sleeves are designed 
and manufactured by Combustion Engineering Incorporated.
    Date of issuance: December 14, 1995
    Effective date: December 14, 1995, to be implemented within 60 days
    Amendment No.: 117
    
[[Page 190]]

    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2868) The May 12, August 9, and September 18, 1995, letters provided 
additional information that did not change the initial proposed no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated December 14, 1995.No 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: September 7, 1993, as supplemented by 
letters dated February 8, 1994, and August 9, 1995.
    Brief description of amendment: The amendment revised the license 
condition on physical security and approves the revision to Physical 
Security Plan for the Waterford Steam Electric Station, Unit 3.
    Date of issuance: December 19, 1995
    Effective date: December 19, 1995
    Amendment No.: 118
    Facility Operating License No. NPF-38. Amendment revised the 
license. The additional information contained in the supplemented 
letter dated August 9, 1995, was clarifying in nature and thus, within 
the scope of the initial notice and did not affect the staff's proposed 
no significant hazards consideration determination.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14887) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 19, 1995.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, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: July 26, 1995
    Brief description of amendments: The amendments consist of changes 
to the Technical Specifications relating to nuclear instrumentation 
system adjustments based on calorimetric measurements at reduced power 
levels.Date of issuance: December 12, 1995
    Effective date: December 12, 1995
    Amendment Nos. 180 and 174Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47617) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 12, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendment: January 26, 1995, as 
supplemented March 9 and May 24, 1995
    Brief description of amendment: This amendment increases the 
allowable U-235 enrichment of fuel to be stored in the new and spent 
fuel storage facilities.
    Date of issuance: December 15, 1995
    Effective date: December 15, 1995
    Amendment No.: 151
    Facility Operating License No. DPR-72. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20517) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 15, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629

Gulf States Utilities Company, Cajun Electric Power Cooperative, 
and Entergy Operations, Inc., Docket No. 50-458, River Bend 
Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: October 24, 1995
    Brief description of amendment: The amendment revised the Technical 
Specifications to reflect the approval for the River Bend Station to 
use 10 CFR Part 50, Appendix J, Option B for the containment leak rate 
testing.
    Date of issuance: December 19, 1995
    Effective date: December 19, 1995
    Amendment No.: 84
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56368) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 19, 1995.No significant 
hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: May 25, 1995 (AEP:NRC:1200B)
    Brief description of amendments: The amendments change the 
surveillance frequency for the manual actuation function for main steam 
line isolation from monthly to quarterly and delete obsolete footnotes 
associated with previous surveillance interval extensions from Unit 2 
Table 4.3-2.
    Date of issuance: December 13, 1995
    Effective date: December 13, 1995, with full implementation within 
45 days
    Amendment Nos.: 204 and 189
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35081)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 13, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of application for amendment: August 8, 1995
    Brief description of amendment: This amendment modifies the 
definitions of Transthermal (Condition 4), Hot Shutdown (Condition 5), 
and Hot Standby (Condition 6) reactor operating conditions. The 
Transthermal and Hot Shutdown Conditions are modified to establish an 
applicable range of subcriticality and be consistent with other 
Definitions. The wording of Hot Standby is modified to remove reference 
to control rod position, consistent with NUREG-1432, Standard Technical 
Specifications for Combustion Engineering Plants, Revision 1, dated 
April 1995.

[[Page 191]]

    Date of issuance: December 15, 1995
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 154
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52931) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 15, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

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

    Date of amendment request: June 27, 1995
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 2.2 on chemical and volume control system (CVCS) to 
reformat and clarify the requirements and make them more consistent 
with the requirements of the Combustion Engineering Standard Technical 
Specifications (STS), as presented in NUREG-0212, Revision 2.
    Date of issuance: December 12, 1995
    Effective date: December 12, 1995
    Amendment No.: 171
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39447) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 12, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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: February 10, 1995, as 
supplemented by letter dated November 10, 1995.
    Brief description of amendments: These amendments (1) modify the 
Susquehanna Steam Electric Station, Unit 1 and 2 Technical 
Specifications to extend the allowable out-of-service times (AOTs) for 
maintenance and repair and the surveillance test intervals (STIs) 
between channel functional tests for the following groups of 
instruments: reactor protection systems instrumentation (TS 3.3.1), 
isolation actuation instrumentation (TS 3.3.2), emergency core cooling 
system actuation instrumentation (TS 3.3.3), ATWS (anticipated 
transient without scram) recirculation pump trip system instrumentation 
(TS 3.3.4.1), end-of-cycle recirculation pump trip system 
instrumentation (TS 3.3.4.2), reactor core isolation cooling system 
(RCIC) actuation instrumentation (TS 3.3.5), control rod block 
instrumentation (TS 3.3.6), radiation monitoring instrumentation (TS 
3.3.7.1), and feedwater/main turbine trip system actuation 
instrumentation (TS 3.3.90); (2) change the required actions and AOTs 
for the instruments listed above to make requirements consistent with 
supporting analysis in General Electric topical reports and change 
additional actions required to prevent extended AOTs from resulting in 
extended loss of instrument function; (3) change the required actions 
and AOTs for the instruments listed above for instrumentation 
associated with the ADS (automatic depressurization system), 
recirculation pump trip, and pump suction lineup for HPCI (high 
pressure core injection) and RCIC; (4) change applicability 
requirements and required actions for the reactor vessel water level-
low, level 3 function that isolates the RHR (residual heat removal) 
system shutdown cooling system so that the function is required to be 
operable in operational conditions 3,4, and 5 to prevent inadvertent 
loss of reactor coolant via the RHR shutdown cooling system; (5) remove 
notes in Table 3.3.2-1, 3.3.2-2, and 4.3.1-1 related to maintenance on 
leak detection temperature detectors and remove the note toTS 3.3.6 for 
Unit 1 related to a previous relief from TS 3.0.4; and (6) reformat, 
renumber, and/or reword existing requirements to incorporate the 
changes listed above.
    Date of issuance: December 18, 1995
    Effective date: As of date of issuance and to be implemented within 
30 days.
    Amendment Nos.: 155 and 126
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16194) The supplemental letter provided corrected TSs and did not 
change the original proposed no significant hazards consideration nor 
the Federal Register notice.The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated December 18, 
1995No significant hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

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: September 26, 1995
    Brief description of amendments: The amendments change the 
containment air lock door seal leakage rate from ``no detectable seal 
leakage'' to ``less than or equal to 0.01 La'' when the gap 
between the door seals is pressurized to greater than or equal to 10 
psig for a period of not less than 15 minutes.
    Date of issuance: December 8, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 118 and 109
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56370) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 8, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

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

    Date of application for amendments:  August 7, 1995 (TS 95-17)
    Brief description of amendments: The changes relocate the heat flux 
hot channel factor penalty from Surveillance Requirement 4.2.2.2.e.1 to 
the Core Operating Limits Report and replace the methodology (WCAP-
10216-P-A) listed in Technical Specification 6.9.1.14.a.2 with WCAP-
10216-P-A, Revision 1A.
    Date of issuance: December 11, 1995
    Effective date: December 11, 1995
    Amendment Nos.: 216 and 206
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45186) The Commission's related evaluation of the amendment is 
contained in a Safety 

[[Page 192]]
Evaluation dated December 11, 1995.No significant hazards consideration 
comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402No significant 
hazards consideration comments received: None

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

    Date of amendment request: August 15, 1995 (TXX-95215)
    Brief description of amendments: These changes relocated the 
Shutdown Margin limits from the Technical Specifications (TSs) to the 
Core Operating Limits Report (COLR). The changes were consistent with 
the intent of Generic Letter 88-16 which provides guidelines for the 
removal of cycle-specific parameter limits from the TSs.
    Date of issuance: December 15, 1995
    Effective date: December 15, 1995
    Amendment Nos.: Unit 1 - Amendment No. 44; Unit 2 - Amendment No. 
30
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52935) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 15, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room Location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019

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

    Date of amendment request: April 26, 1995
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3/4.7.6 to reduce the upper limit on the flow rate 
through the control room filtration subsystem and adopts ASTM D-3803-
1989 as the laboratory testing standard for control room filtration and 
control building pressurization charcoal adsorber. The amendment also 
revises the Bases for TS 3/4.7.6 to reflect the changes.
    Date of issuance: December 20, 1995
    Effective date: December 20, 1995, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 106
    Facility Operating License No. NPF-30. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27345) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 20, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of amendment request: June 14, 1995, as supplemented by 
letters dated July 13, 1995, and August 22, 1995.I11Brief description 
of amendment: The amendment revises Technical Specification (TS) 3.2.3, 
``Nuclear Enthalpy Rise Hot Channel Factor,'' TS 6.9.1.9, ``Core 
Operating Limits Report,'' and the associated Bases sections. The 
revisions incorporate changes associated with the planned 
implementation of advanced nuclear and core thermal-hydraulic design 
methodologies licensed from Westinghouse Electric Corporation for core 
reload design, starting with Cycle 9.
    Date of issuance:  December 8, 1995
    Effective date: December 8, 1995, to be implemented prior to 
restart from the eighth refueling outage, which is scheduled to begin 
in March 1996.
    Amendment No.: 92
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39456) The August 22, 1995, supplemental letter forwarded the 
nonproprietary version of Wolf Creek Nuclear Operating Corporation's 
safety evaluation and analysis provided in the June 14, 1995, submittal 
and did not change the staff's original no significant hazards 
determination.The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 8, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

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

    Date of amendment request: August 22, 1995
    Brief description of amendment: The amendment revises the 
requirements of Technical Specification (TS) 3.3.1 and TS 3.3.2 and 
relocate Tables 3.3-2 and 3.3-5 and applicable Bases, which provide the 
response time limits for the reactor trip system (RTS) and the 
engineered safety features actuation system (ESFAS) instruments, from 
the TS to the Updated Safety Analysis Report (USAR). The licensee has 
stated that the next USAR change request will include these changes.
    Date of issuance: December 12, 1995
    Effective date: December 12, 1995, to be implemented within 60 days 
of issuance.
    Amendment No.: 93
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49950) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 12, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Dated at Rockville, Maryland, this 21st Day of December 1995.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor 
Regulation
[Doc. 96-1 Filed 1-2-96; 8:45 am]
BILLING CODE 7590-O1-F