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

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.

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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

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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 (a'' 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]

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