Biweekly Notice

Federal RegisterJun 19, 1996

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from May 24, 1996, through June 7, 1996. The last

biweekly notice was published on June 5, 1996 (61 FR 28604).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By July 19, 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

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petition for leave to intervene shall be filed in accordance with the

Commission's ``Rules of Practice for Domestic Licensing Proceedings''

in 10 CFR Part 2. Interested persons should consult a current copy of

10 CFR 2.714 which is available at the Commission's Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC and at

the local public document room for the particular facility involved. If

a request for a hearing or petition for leave to intervene is filed by

the above date, the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or by the Chairman of the Atomic Safety

and Licensing Board Panel, will rule on the request and/or petition;

and the Secretary or the designated Atomic Safety and Licensing Board

will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Docketing and

Services Branch, or may be delivered to the Commission's Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,

by the above date. Where petitions are filed during the last 10 days of

the notice period, it is requested that the petitioner promptly so

inform the Commission by a toll-free telephone call to Western Union at

1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: April 25, 1996

Description of amendment request: The proposed amendment would

change the definition of Operable-Operability, revise Technical

Specifications (TSs) and associated Bases Section for TSs 3.5.F.1,

``Core and Containment Cooling systems,'' TSs 3.9.B.1, 3.9.B.2,

3.9.B.3, 3.9.b.4, ``Auxiliary Electrical System,'' and TSs 3.7.B.1.a,

c, and e, and 3.7.b.2.a, c, and e, ``Standby Gas Treatment System and

Control Room High Efficiency Air Filtration System,'' and delete TSs

4.5.F.1, ``Core and Containment Cooling Systems,'' and 3.7.B.1.f,

``Standby Gas Treatment System and Control Room High Efficiency Air

Filtration System.''

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Operation of PNPS [Pilgrim Nuclear Power Station] in accordance

with the proposed license amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated because of the following:

Definition of ``Operable-Operability''

Definitions perform a supporting function for other sections of

the TS. The definition of ``Operable-Operability'' affects the

manner

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in which the requirements for a Limiting Condition for Operation

(LCO) and its associated remedial actions are applied when a support

system is inoperable. This definition re-affirms the principle that

a system is operable when it is capable of performing its specified

function and when all necessary support systems are also capable of

performing their related support functions. The corollary is that a

system is inoperable when it is not capable of performing its

specified function or when a necessary support system is not capable

of performing its related support function.

No changes are being made to the plant design, system

configuration, or method of operation. The proposed change does not

affect the ability of the AC power sources to perform their required

safety functions nor affect the ability of the features they support

to perform their respective safety functions. Therefore, the

proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

EDG [Emergency Diesel Generator]

An Individual Plant Examination (IPE) for Internal Events was

submitted to the NRC in response to Generic Letter 88-20 in

September 1992. The IPE was used to quantify the overall impact of

the proposed 14 day allowed outage time on core damage frequency.

Part III provides the results of a comprehensive Probabilistic

Safety Assessment (PSA) of the impact of the proposed AOTs [allowed

outage times] for the EDGs and Startup and Shutdown transformers. As

shown in Part III, there is not a significant increase in risk due

to the proposed change. Thus the proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The existing specification 3.9.B.1 is being separated into two

segments (a and b) because of the proposed and different AOTs for

the Startup and Shutdown transformers. As a result of the PSA, the

AOT for the Startup transformer (a) is reduced from 7 days to 72

hours, while the AOT for the Shutdown transformer (b) remains at 7

days. The reduction of the AOT from 7 days to 3 days is based on the

relative risk importance of the Startup transformers support to the

balance of plant systems. Similarly, an additional reduction from 72

hours to 48 hours is proposed in the AOT for a simultaneous loss of

both the Startup transformer and an EDG (TS 3.9.B.4.b) based upon

the Startup transformer's contribution to risk in relation to the

EDG 14-day AOT risk assessment analysis and that two power sources

have been removed from the associated bus. The AOT reductions

represent a measurable decrease in risk as assessed in the PSA.

Thus, the probability or consequences of an accident previously

evaluated are not significantly increased.

The current technical specifications allow one EDG to be out of

service for three days based on the availability of the SUT [startup

transformer] and SDT [shutdown transformer] and the fact that each

EDG carries sufficient engineered safeguards equipment to cover all

design basis accidents. With one EDG out of service and a Loss of

Offsite Power (LOOP) condition, the capability to power vital and

auxiliary system components remains available via the other EDG, and

for one train of ESF equipment via the SDT for all operating,

transient and accident conditions. Increasing the EDG AOT to 14 days

provides flexibility in the maintenance and repair of the EDGs. The

EDG unavailability will be monitored and trended in accordance with

the Maintenance Rule. The PSA analyses supports the change to a 14

day AOT for the EDGs based on an insignificant increase in overall

risk. Implementation of the proposed change is expected to result in

less than a one percent increase in the baseline core damage

frequency (2.84E-05/yr), which is considered to be insignificant

relative to the underlying uncertainties involved with probabilistic

safety assessments. Additional conditions are added to the Standby

Liquid Control, Standby Gas Treatment, and Control Room High

Efficiency Air Filtration systems requiring the EDG associated with

these systems to remain operable while in the 14 day EDG AOT. Thus,

the 14 day EDG AOT does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Eliminating the 4.5.F.1 requirement for daily testing of the

operable diesel generator when the redundant diesel generator

becomes inoperable is consistent with the guidance provided in

Generic Letter 93-05. The change does not affect the ability of the

emergency diesel generator to perform on demand, and by actually

lowering the number of demands to demonstrate operability, reduces

the probability of equipment failure. The redundant EDG will remain

in service during the entire period of inoperability of the out-of-

service EDG. If a common cause failure cannot be ruled out, the

redundant EDG will be tested to assure operability. The proposed

revisions do not involve a significant change to the plant design or

operation, only to the manner in which remaining equipment is

confirmed to be operable, which is consistent with NRC guidance.

Thus operation of PNPS in accordance with the proposed license

amendment will not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The 3.9.B.1 and 2 requirements to demonstrate both EDGs and

associated emergency buses operable are deleted. This change is

based on the NRC guidance provided in item 10.1 of Generic Letter

93-05, ``Line-Item Technical Specification Improvements to Reduce

Surveillance Requirements for Testing During Power Operation.''

Revising the methods for verifying EDG and emergency bus operability

does not physically alter the plant or have an affect on the

probability or consequences of an accident previously evaluated.

Deleting the testing requirements for an EDG when the other EDG is

inoperable does not increase the probability or consequences of an

accident previously evaluated because the reliability program and

routinely performed TS surveillances continue to provide the added

assurance sought by the testing. The elimination of this testing

will serve to improve the overall reliability of the EDGs. Since the

proposed change does not affect the design or negatively affect the

performance of the EDGs, the change will not result in a significant

increase in the consequences or probability of an accident

previously analyzed.

SGT [Standby Gas Treatment] and CRHEAF [Control Room High

Efficiency Air Filtration]

During normal plant operation, with one SGT or CRHEAF subsystem

inoperable, the inoperable subsystem must be restored to operable

status in 7 days. In this condition, the remaining operable SGT or

CRHEAF subsystem is adequate to perform the required radioactivity

release control function. However, the overall system reliability is

reduced because a single failure in the operable subsystem could

result in the radioactivity release control function not being

adequately performed. The 7 day completion time is based on

consideration of such factors as the availability of the operable

redundant SGT subsystem and the low probability of a DBA [design

basis accident] occurring during this period.

If the SGT or CRHEAF subsystem cannot be restored to operable

status within 7 days when in the Run, Startup, or Hot Shutdown MODE,

the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least Hot

Shutdown within 12 hours and to Cold Shutdown within 36 hours. The

allowed completion times are reasonable, based on operating

experience, to reach the required plant conditions from full power

conditions in an orderly manner and without challenging plant

systems.

Current TS governing refueling operations restrict fuel movement

if one train of SGTS or one train of CRHEAF are inoperable. In this

condition the remaining operable SGT and CRHEAF trains are adequate

to perform the required radioactivity release control functions.

However, the overall system reliability is reduced because a single

failure in the operable train could result in the radioactivity

release control function of the systems not being adequately

performed. New requirements are added that require if one train of

SGT or CRHEAF is inoperable, the redundant train of SGT or CRHEAF

must be demonstrated to be operable within 2 hours. This

substantiates the availability of the operable trains. Fuel handling

is limited only to the following 7 days and if the inoperable train

is not returned to an operable condition within that time frame, the

operable SGT train is placed in operation or fuel handling

activities are suspended. For CRHEAF, after 7 days, the operable

subsystem is demonstrated operable in accordance with existing

surveillances on a daily basis. The proposed changes do not modify

system design, use, or configuration in a manner different from

their original design and therefore do not involve a significant

increase in the consequences or probability of an accident

previously analyzed.

The revisions to make the SGT and CRHEAF TS sections similar in

wording are made to enhance usability and alleviate possible

confusion. These changes are strictly editorial, have no impact, and

do not alter

[[Page 31174]]

technical content or meaning of the specifications. These editorial

changes do not involve a significant increase in the probability or

consequences of an accident previously analyzed.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The operation of PNPS in accordance with the proposed license

amendment will not create the possibility of a new or different kind

of accident from any accident previously evaluated because of the

following:

Definition of ``Operable-Operability''

The revised definition redefines the AC power needs to allow

either onsite or offsite power available for systems/subsystems to

be considered operable. This does not compromise the level of safety

already afforded to such systems/subsystems because the functional

operability requirements continue to be assured through the

technical specifications applicable to such systems/subsystems. AC

power availability continues to be assured through existing and

proposed surveillances and action statements applicable to AC power

systems. Reducing the need for both onsite and offsite power sources

in order to consider operable, the systems/subsystems powered by

these AC power sources, provides additional operational flexibility

by allowing redundant systems/subsystems to still be considered

``operable'' within the requirements of their functional operability

requirements. No new change or modes of plant operation are

involved. Therefore, operation in accordance with the revised

definition does not introduce any new or different kind of accident

from any accident previously evaluated.

EDG

The proposed amendment will extend the action completion/allowed

outage time for an inoperable emergency diesel generator from 72

hours to 14 days. The EDGs are designed as backup AC power sources

for essential safety systems in the event of loss of offsite power.

The proposed AOT does not change the conditions, operating

configurations or minimum amount of operating equipment assumed in

the safety analysis for accident mitigation. The EDGs and AC

equipment are not accident initiators. No change is being made in

the manner in which the EDG's provide plant protection. No new modes

of plant operation are involved. An extended AOT for one EDG does

not increase the probability of occurrence of a new or different

kind of accident previously evaluated. The PSA results concluded

that the risk contribution of the EDG AOT extension is

insignificant.

The current Pilgrim Technical Specifications requiring immediate

and daily testing of the redundant operable EDG is based on the

assumption that the increased testing provides additional assurance

that the equipment is available should it be needed. Industry

experience indicates that repetitive testing can place demands and

wear on the EDG without necessarily providing additional confidence

of availability. Also, the new surveillance requires verification

that offsite power is available and that a common cause failure is

not present. These actions provide assurance that the required

emergency buses can be energized with no loss of functions to

mitigate accident or transient conditions. In addition, Pilgrim has

implemented an EDG reliability program to maintain reliability of

EDGs. The proposed change does not introduce any new mode of plant

operation or new accident precursors, involve any physical

alterations to plant configurations, or make changes to system set

points that could initiate a new or different kind of accident.

Therefore, operation in accordance with the proposed change does not

create the possibility of a new or different kind of accident from

any previously evaluated.

The AOT for an inoperable Startup Transformer is reduced from 7

days to 72 hours based upon the PSA that was performed to

quantitatively assess the risk impact of the proposed amendment. The

proposed reduction in AOT improves overall AC power source

availability because the SUT will potentially be inoperable for

shorter time periods. Therefore, reducing the AOT does not create

the possibility of a new or different kind of accident from any

accident previously evaluated.

SGT and CRHEAF

The SGT system is designed to filter radioactive materials from

the secondary containment following a postulated DBA or fuel

handling accident prior to release to the environment to ensure

compliance with 10 CFR 100 limits.

The CRHEAF is designed to filter intake air for the control room

atmosphere during conditions when normal intake air may be

contaminated.

The proposed revisions do not affect the ability of the SGTS or

CRHEAF to perform their intended function, do not create the

possibility of a new or different kind of accident from the loss of

coolant or fuel handling accidents previously analyzed, and do not

modify system configuration, use, or design. Therefore, operating

Pilgrim in accordance with this change will not create the

possibility of a new or different kind of accident from any accident

previously analyzed.

The revisions to make the SGT and CRHEAF TS sections similar in

wording are made to enhance usability and alleviate possible

confusion. These changes are strictly editorial, have no impact, and

do not alter technical content or meaning of the specifications.

These editorial changes do not create the possibility of a new or

different kind of accident from any previously analyzed.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The operation of PNPS in accordance with the proposed license

amendment will not involve a significant reduction in a margin of

safety because of the following:

Definition of ``Operable-Operability''

The implementation of the ``Operability'' definition clarifies

the relationship between AC power supplies and the operability

status of the equipment requiring AC power. No change is being made

in which the plant systems relied upon in the safety analyses

provide plant protection. Plant safety margins are maintained

through the limitations established in the TS LCOs. Since there will

be no significant reduction to the physical design or operation of

the plant there will be no significant reduction to any of these

margins.

EDG

Operation of PNPS in accordance with the proposed license

amendment will not involve a significant reduction in a margin of

safety. As shown in Part III [of the application dated April 25,

1996], incorporation of the proposed change involves an

insignificant reduction in the margin of safety.

The proposed changes do not significantly reduce the basis for

any technical specification related to the establishment of, or the

maintenance of, a safety margin nor do they require physical

modifications to the plant. Additional conditions are added to the

Standby Liquid Control, Standby Gas Treatment, and Control Room High

Efficiency Air Filtration systems requiring the diesel generator

associated with the redundant operable trains of these systems to

remain operable while in the 14 day EDG AOT. Moreover, the PSA

results showed that the risk contribution of extending the AOT for

an inoperable EDG is insignificant. The reduction in the AOT for the

SUT could improve availability, therefore, reducing overall risk.

Likewise the proposed changes in the deletion of testing have no

impact on the safety margin.

As previously stated, implementation of the proposed changes is

expected to result in an insignificant increase in: (1) power

unavailability to the emergency buses (given that a loss of offsite

power has occurred), and (2) core damage frequency. Implementation

of the proposed changes does not increase the consequences of a

previously analyzed accident nor significantly reduce a margin of

safety. Functioning of the EDGs and the manner in which limiting

conditions of operation are established are unaffected.

SGT and CRHEAF

SGT and CRHEAF contribute to the margin of safety by supporting

the secondary containment system during fuel handling by mitigating

the consequences of a fuel handling event. Allowing fuel movement to

continue as established in the LCOs does not involve a significant

reduction in the margin of safety because the first line of defense,

the other SGT and CRHEAF trains will be operable. The proposed

change will allow placing the Operable SGT subsystem in operation,

or in the case of CRHEAF, conducting daily testing, as an

alternative to suspending movement of irradiated fuel. This

alternative is less restrictive than the existing requirement,

however, the proposed requirements ensure that the remaining

subsystem is operable, that no failures that could prevent actuation

have occurred, and that any failure would be readily detected. The

proposed change does not result in a significant reduction in a

margin of safety because it allows operations which have the

potential for releasing radioactive material to the secondary

containment to continue only if the system designed to mitigate the

[[Page 31175]]

consequences of this release is functioning. Proper operation of

only one SGT or one CRHEAF subsystem is sufficient to mitigate the

consequences of any analyzed accident. Therefore, this change does

not change any of the assumptions in the accident analysis and does

not involve a significant reduction in a margin of safety.

The revisions to make the SGT and CRHEAF TS sections similar in

wording are made to enhance usability and alleviate possible

confusion. These changes are strictly editorial, have no impact, and

do not alter technical content or meaning of the specifications.

These editorial changes do not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Jocelyn A. Mitchell, Acting

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County, Connecticut

Date of amendment request: April 22, 1996

Description of amendment request: The licensee is proposing to

change the technical specifications to reflect a revision to the

overload cutoff limit on the manipulator crane inside the containment

at the Haddam Neck Plant. Due to a change in fuel design and supplier,

the heaviest fuel assembly design starting in Cycle 20 will be the

Westinghouse-supplied LOPAR design. Therefore, the heaviest combination

beginning in Cycle 20 will be the Westinghouse LOPAR fuel assembly with

a full-length rod cluster control assembly (RCCA) inserted. It will now

be used as the standard for the overload cutoff limit on the

manipulator crane.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. [The proposed change does not] involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will revise the method of determining the

overload cutoff limit for the manipulator crane. The actual absolute

value of the cutoff limit will not be increased and will not affect

the [probability] of any plant accidents.

Since there is no actual increase in the absolute overload

cutoff limit, there will be no adverse effects to the crane, cables,

or associated hardware. Therefore, there is no impact on the crane's

ability to perform its intended function. Even though the net

lifting forces on an individual assembly have increased 25 pounds,

the limit is within the recommended Westinghouse guidelines with

respect to fuel handling and will not result in potential damage to

assembly grids during fuel handling activities.

As such, CYAPCO [Connecticut Yankee Atomic Power Company] has

concluded that these changes do not involve an increase in the

probability or consequences of an accident previously evaluated.

2. [The proposed change does not] create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The changes conservatively revise the method of determining the

overload cutoff limit for the manipulator crane. There is no impact

on the basic functioning of plant systems or equipment. Therefore,

the change does not create a malfunction that is different from

those previously evaluated.

As such, the proposed changes described above do not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. [The proposed change does not] involve a significant

reduction in a margin of safety.

The proposed revisions in the methodology for determining the

overload cutoff limit for the manipulator crane is conservative and

in accordance with vendor standards. The changes do not adversely

affect any equipment credited in the safety analysis. Also, the

changes do not adversely affect the probability or consequences of

any plant accident, including the fuel handling accident or offsite

doses associated with those accidents.

As such, the proposed changes have no significant impact on 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: Russell Library, 123 Broad

Street, Middletown, CT 06457

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270

NRC Project Director: Phillip F. McKee

Duke Power Company, Docket Nos. 50-413 and 50-414, Catawba Nuclear

Station, Units 1 and 2, York County, South Carolina

Date of amendment request: December 14, 1995, as supplemented by

letter dated May 16, 1996

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) to improve the TS Action

Statements and Surveillance Requirements for diesel generators in

accordance with the recommendations and guidance in Generic Letter 93-

05, Generic Letter 94-01, NUREG-1366, and NUREG-1431. The proposed

amendments would also incorporate technical and administrative changes.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

Operation of the facilities in accordance with the requested

amendments will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Improvements to the LCOs [limiting condition for operation] and

surveillance requirements for the emergency diesel generators do not

affect their capability to provide emergency power to plant vital

instruments and safety related equipment. In fact, these

improvements make the diesel generators more reliable since they

significantly reduce the amount of wear and stress due to excessive

and unnecessary testing. The proposed monthly testing of the diesel

generator continues to ensure that the system is ready for service

when needed. The fast starts and fast loadings continue to ensure

that the timing and loading requirements for engineered safety

features actuation are met. The proposed changes do not affect any

of the design basis accident analyses previously evaluated.

Therefore, these proposed changes do not involve any increase in the

probability or consequences of any accident previously evaluated.

The proposed changes are fully consistent with the recommendations

and guidance contained in GL [Generic Letter] 93-05, GL 94-01,

NUREG-1366, NUREG-1431, and are compatible with plant operating

experience.

Criterion 2

Operation of the facilities in accordance with the requested

amendments will not create the possibility of a new or different

kind of accident from any accident previously evaluated. The

proposed changes in fact improve the reliability of the diesel

generators by eliminating unnecessary wear and stress. Improved

reliability decreases the failure probability which also decreases

the probability of an accident not previously evaluated. None of the

requested amendments increase the common mode failure probability

thus would not increase the chance of both EDG's [emergency diesel

[[Page 31176]]

generators] for a particular nuclear unit being out of service

simultaneously. The proposed changes are fully consistent with the

recommendations and guidance contained in GL 93-05, GL 94-01, NUREG-

1366, NUREG-1431, and are compatible with plant operating

experience.

Criterion 3

Operation of the facilities in accordance with the requested

amendments will not involve a significant reduction in a margin of

safety. The proposed monthly testing of the diesel generators

continues to ensure that the system is ready for service when

needed. The fast starts and fast loadings continue to ensure that

the timing and loading requirements for engineered safety features

actuation are met. The proposed changes improve the reliability of

the diesel generators. Implementation of the Maintenance Rule also

ensures continued reliability of the diesel generators. No margin of

safety is decreased as a result of these TS changes.

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: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi and Docket

No. 40-458, River Bend Station, Unit 1, West Feliciana Parish,

Louisiana

Date of amendment request: April 18, 1996, as supplemented by

letter dated June 4, 1996

Description of amendment request: The licensee has proposed to (1)

amend Limiting Condition for Operation (LCO) 3.10.6 and Surveillance

Requirement 3.10.6.3, and (2) add a Surveillance Requirement 3.10.6.4

of the Technical Specifications (TSs) for the Grand Gulf Nuclear

Station, Unit 1, and the River Bend Station, Unit 1, to allow another

method of fuel movement and loading in the core when control rods are

removed or withdrawn from defueled core cells. Currently, LCO 3.10.6

allows only fuel loading as part of the approved spiral reloading

sequence to prevent fuel loading into core cells in which the control

rod has been removed or withdrawn. This amendment request does not

withdraw this approved method, revise the frequency of performing the

surveillance during fuel loading, or alter the method of verifying the

fuel is being loaded in compliance with the approved method. Grand Gulf

Unit 1 and River Bend Unit 1 are both General Electric (GE) Boiling

Water Reactor (BWR)-6 plants, the latest version of the GE design

series.

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:

Entergy Operations, Inc. [(EOI)] propose[d] to change the

current Grand Gulf Nuclear Station (GGNS) and River Bend Station

(RBS) Technical Specifications [(TSs)]. The specific proposed change

is to add an additional method of performing fuel loading into LCO

3.10.6, ``Multiple Control Rod Withdrawal - Refueling''. The

proposed change would allow fuel loading [in the core] if a positive

means of assuring fuel assemblies cannot be loaded into a core cell

with a withdrawn or removed control rod is in effect. [Currently,

the TSs for both plants allow fuel assembles to be loaded in

compliance with an approved spiral reload sequence which is used to

ensure the reactivity additions are minimized. Spiral loadings

encompass reloading a core cell on the edge of a continuous fueled

region.]

The Commission has provided standards for determining whether a

no significant hazards consideration exists as stated in 10 CFR

50.92(c). A proposed amendment to an operating license involves no

significant hazards consideration if 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.

Entergy Operations, Inc. [EOI] has evaluated the no significant

hazards consideration in its request for this license amendment and

determined that no significant hazards consideration results from

this change. In accordance with 10 CFR 50.91(a), Entergy Operations,

Inc. [EOI] is providing the analysis of the proposed amendment

against the three standards in 10 CFR 50.92(c). A description of the

no significant hazards consideration determination follows:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

The refueling interlocks (i.e., the refueling equipment and one-

rod-out interlocks) allowed to be bypassed by Technical

Specification [TS] LCO 3.10.6 are explicitly assumed in the analysis

of the control rod removal error or fuel loading error during

refueling. This analysis evaluates the consequences of control rod

withdrawal during refueling. Criticality and, therefore, subsequent

prompt reactivity excursions are prevented during the insertion of

fuel, provided all control rods are fully inserted during the fuel

insertion. The refueling interlocks accomplish this by preventing

loading fuel into the core with any control rod withdrawn, or by

preventing withdrawal of a rod from the core during fuel loading.

LCO 3.10.6 allows multiple control rod withdrawals, control rod

removals, associated control rod drive (CRD) removal, or any

combination of these, and the ``full in'' position indication input

to the refueling interlocks is allowed to be bypassed for each

withdrawn control rod if all fuel has been removed from the cell.

This supports the GGNS Updated Final Safety Analyses Report (UFSAR)

and RBS Updated Safety Analyses Report (USAR) analyses since, with

no fuel assemblies in the core cell, the associated control rod has

no reactivity control function and does not need to remain inserted.

Prior to reloading fuel into the cell, however, the associated

control rod must be inserted to ensure that an inadvertent

criticality does not occur, as evaluated in the analysis.

The Technical Specification [TS] requirements prohibiting fuel

loading was placed in the Technical Specifications [TSs] for GGNS

and RBS as part of the originally enforced Technical Specification

[TS] requirements to resolve NRC concerns identified in IE

Information Notice No. 83-35, ``Fuel Movement with Control Rods

Withdrawn at BWRs,'' (IEN 83-35). IEN 83-35 details instances where

fuel assemblies were loaded into core cells while the control rod

was withdrawn and discusses that the General Electric Company (GE)

had issued Service Information Letter (SIL) No. 372.

SIL No. 372 discusses a potential event where 8 fuel assemblies

are loaded into 2 [two] adjacent core cells where the control rods

are withdrawn and no action is taken to recover from the errors. In

this SIL GE identified that the probability of such an event

occurring was extremely low but potentially slightly higher than

10-6 probability of the event even further to where it need not

be considered credible (i.e., below 10-6 per reactor year), GE

recommended that the additional administrative control of

prohibiting loading fuel with withdrawn rods be enforced.

The proposed change will only provide an additional way to meet

the intent of the original GE recommendation. [The currently

approved method is listed in LCO 3.10.6 and Surveillance Requirement

3.10.6.3.]. The proposed change will provide the additional

allowance to perform fuel loading only if an additional positive

means of assuring fuel assemblies cannot be loaded into a core cell

with a withdrawn or removed control rod is in effect. The positive

means will entail a physical barrier such that, even if refueling

procedures were violated and an attempt was made to load a fuel

assembly into a core cell with a withdrawn or removed control rod,

the action would be prevented. This requirement provides sufficient

additional restrictions to meet the intent of the GE recommendation

to add additional administrative controls to prevent the postulated

event from occurring.

The probability of an inadvertent criticality occurring will

continue to be precluded by

[[Page 31177]]

the same number of layers of administrative controls [as the

currently approved method]; therefore, the proposed change does not

significantly increase the probability or consequences of an

accident previously evaluated.

II. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The administrative changes in the Technical Specification [TS]

requirements do not involve a change in the design of the plant. The

proposed requirements will continue to ensure that fuel is not

loaded into a core cell that is associated with a removed or

withdrawn control rod.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

The margin of safety associated with criticality events during

fuel handling is provided by the event being a non credible event.

The proposed change will only provide an additional means to meet

the same intent of ensuring that the event is of such low

probability as to be considered non credible. The proposed change

will provide the additional allowance to perform fuel loading only

if an additional positive means of assuring fuel assemblies cannot

be loaded into a core cell with a withdrawn or removed control rod

is in effect. The positive means will entail a physical barrier such

that even if refueling procedures were violated and an attempt was

made to load a fuel assembly into a core cell with a withdrawn or

removed control rod the action would be prevented. This requirement

provides sufficient additional restrictions to ensure that the event

is of such low probability as to be considered non credible.

The probability of an inadvertent criticality occurring will

continue to be precluded by the same number of layers of

administrative controls [as the currently approved method];

therefore, this change does not reduce the level of safety imposed

by the current Technical Specification [TS] requirements.

Therefore, the proposed changes do not cause 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: (1) Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120, for Grand Gulf

Nuclear Station and (2) Government Documents Department, Louisiana

State University, Baton Rouge, LA 70803, for River Bend Station.

Attorney for licensee: (1) Nicholas S. Reynolds, Esquire, Winston

and Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502,

for Grand Gulf Nuclear Station and (2) Mark Wetterhahn, Esq., Winston &

Strawn, 1400 L Street, N.W., Washington, DC 20005, for River Bend

Station.

NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: May 9, 1996

Description of amendment request: The amendment request would allow

allow the licensee to perform the surveillance of the relief mode of

operation of each of the 20 safety/relief valves (S/RVs) on the 4 main

steam lines without physically lifting the disk off the seat at power.

The proposed changes are to Surveillance Requirements (SRs) 3.4.4.3,

Safety/Relief Valves, 3.5.1.7, Automatic Depressurization System

Valves, and 3.6.1.6.1, Low-Low Set Valves, of the Technical

Specifications, and the changes would state that the required operation

of the valve to verify is that the relief-mode actuator strokes when

the valve is manually actuated. Each S/RV is a Dikkers, 8 X 10, direct-

acting, spring loaded, safety valve with attached pneumatic actuator

for relief-mode operation. Eight of the S/RVs use the relief mode to

perform the Automatic Depressurization System (ADS) function. Also, six

S/RVs, two of which are also ADS S/RVs, use the relief mode to perform

the Low-Low Set valve function. The licensee also proposed changes to

the Bases of the Technical Specifications that are associated with the

above proposed changes.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below: The Dikkers S/RV provides

pressure relief based on the principle of vertically moving the stem

that attaches directly to the valve disk. The force that provides the

stem movement is provided by one of two sources; the vessel pressure

directly against the force of the stem spring (safety mode), or the

pneumatic actuator arm against the force of the stem spring (relief

mode). ASME Boiler and Pressure Vessel Code requires testing the safety

mode of operation once every five year operating cycle. Once a safety

valve is installed, the safety mode is never tested while the S/RV is

installed in the plant. The testing of the relief mode of operation for

a direct-acting S/RV provides verification that the control functions

of electrical and pneumatic connections have been properly reconnected,

and that the actuator arm will provide the necessary force to operate

the S/RV.

This proposed change provides verification of proper control

connections by requiring the pneumatic and electrical controls to

cycle the actuator arm on each S/RV after installation in the

drywell. The test population of S/RVs removed each outage for safety

setpoint testing will be tested in the relief mode. This testing

will demonstrate that the installed S/RVs will function properly in

the relief mode. The remaining installed S/RVs will continue to be

tested for proper system function. As presently required by GGNS

Technical Specifications and administrative procedures, proper

operation of the solenoid control block will be demonstrated by

providing an open signal to each S/RV, with a check to verify that

each solenoid valve repositions. Verification of proper solenoid

valve operation, in addition to the proper relief-mode operation of

the test population, provides assurance that the S/RV will perform

as expected when control air pressure is applied to the solenoid

valve control block.

Entergy Operations, Inc. is proposing that the Grand Gulf

Nuclear Station Operating License be amended to perform the

surveillance of each safety relief valve (S/RV) relief mode of

operation without physically lifting the disk off the seat at power.

During the refueling outage, a sample population of the S/RVs

will be removed for safety-mode setpoint testing in accordance with

the GGNS IST program, using ASME Boiler and Pressure Vessel Code,

Section XI. Each of these removed S/RVs will be tested in the relief

mode to verify that the pneumatic actuator functions correctly, and

this test sample will be used to provide assurance that the

installed S/RV pneumatic actuators will function properly. After the

test sample of S/RVs has been replaced with recertified spares, and

S/RV controls have been connected, the upper stem nut that couples

the valve stem to each newly- installed S/RV's pneumatic actuator

will be moved up the stem to allow an uncoupled actuation of the

relief-mode actuator. Control air pressure to each actuator will be

reduced from normal system pressure to prevent damaging the

pneumatic relief-mode actuator. The actuator will be remotely

operated from the control room, as required by current test methods,

and visual verification will be performed for proper actuator

response and range of motion. After proper actuator operation has

been verified, the upper stem nut will be returned to its operating

stem location. Verification of proper system logic controls and

function for every installed S/RV will continue to be performed, as

required by Technical Specifications.

The commission has provided standards for determining whether a

no significant hazards consideration exists as stated in 10 CFR

50.92(c). A proposed amendment to an operating license involves no

significant hazards if the operation of the facility in accordance

with the proposed amendment would not: (1) involve a significant

increase

[[Page 31178]]

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.

Entergy Operations has evaluated the no significant hazards

considerations in its request for a license amendment. In accordance

with 10 CFR 50.91(a), Entergy Operations, Inc. is providing the

following analysis of the proposed amendment against the three

standards in 10 CFR 50.92:

a. No significant increase in the probability or consequences of

an accident previously evaluated results from this change.

Each refueling outage, a test sample of the population of S/RVs

is removed from the plant to perform testing as required by ASME

Boiler and Pressure Vessel Code, Section XI. These S/RVs will be

stroked in the relief mode during as-found testing, and are

therefore verified to operate properly when each S/RV stem is raised

by the relief-mode pneumatic actuator. This proposed surveillance

verifies proper S/RV relief-mode operation of all installed S/RVs

based upon this test sample. This testing, in conjunction with

replacement of each S/RV prior to the end of its expected service

life, provides reasonable assurance that the installed S/RVs will

perform as well as the test population of S/RVs.

After the S/RVs have been replaced in the plant, and after all

controls are reconnected, the relief-mode actuator on each newly-

installed S/RV will be uncoupled from the S/RV stem, and stroked.

This actuator stroke will verify that no damage has occurred to the

relief-mode actuator during S/RV transportation from its storage

location to its operating location. The direct coupling of the valve

stem to disk provides assurance that proper relief actuation will

occur when the actuator is operated. The safety-mode components are

completely encased within the valve body and bonnet, which provides

a rugged structure to prevent damage to these components. The

remaining installed S/RVs will continue to be tested for proper

control system function as previously required by Technical

Specifications. The direct coupling of the S/RV stem to disk

provides assurance that proper relief-mode actuation will occur when

the actuator is operated. The safety mode of the GGNS S/RVs is not

affected by a malfunction of the relief-mode components.

Blockage of each S/RV discharge line will be prevented by the

same Foreign Material Exclusion (FME) controls that exist for other

reactor vessel and support systems. These FME controls, combined

with the horizontal orientation of the S/RV discharge piping mating

surfaces, provide reasonable assurance that discharge line blockage

will not occur.

Therefore, no significant increase in the probability or

consequences of an accident previously evaluated results from this

proposed change.

b. This change would not create the possibility of a new or

different kind of accident from any previously analyzed.

The proposed change demonstrates that each S/RV will perform its

intended relief-mode function, which is the intent of the present

surveillance. The relief mode of S/RV operation is demonstrated to

be operable based upon successful performance of a test population,

S/RV component service life, and existing Technical Specification

surveillances. No new failure mechanisms to the relief- mode of

operation are introduced, as the proposed surveillance verifies

relief actuator operability. Plant FME controls, combined with the

horizontal orientation of the S/RV discharge piping mating flange,

provides reasonable assurance that discharge line blockage will not

occur. This proposed change does not add any new systems,

structures, or components, nor does it introduce new S/RV operating

modes.

Therefore, this change would not create the possibility of a new

or different kind of accident from any previously analyzed.

c. This change would not involve a significant reduction in the

margin of safety.

This proposed change will verify that the relief mode of all

installed S/RVs will operate properly based upon demonstrated relief

mode performance of a sample of S/RVs. The failure mode of the S/RV

relief function would require a failure of either the pneumatic

actuator, lifting linkage, or solenoid block. Each of these items

has been verified to have a service life exceeding the replacement

cycle of each S/RV. Therefore, proper operation of a sample

population of S/RVs provides reasonable assurance that the remaining

S/RVs would perform identically, within the original margin of

expected S/RV operability. In addition, each S/RVFEs solenoid block

and control functions will continue to be tested and cycled each

refueling outage. The removal of the valve stroke surveillance for

all S/RVs does not increase the possibility of valve malfunction,

since valve stroke is verified during the as-found testing of the

sample population of S/RVs. This proposed surveillance test reduces

the number of S/RV actuations, and therefore, reduces challenges to

the system both mechanically and thermally. Also, the proposed

alternative method of testing reduces the possibility of a stuck-

open S/RV, since this proposed method will not stroke the S/RVs with

the reactor pressurized during reactor power operations.

Therefore, this change would not involve a significant reduction

in the margin of safety.

Based on the above evaluation, Entergy Operations, Inc. has

concluded that operation in accordance with the proposed amendment

involves no significant hazards considerations.

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: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502

NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: May 31, 1996

Description of amendment request: The amendment would provide an

alternative method to compensate for inoperable refueling equipment

interlocks. The alternative method would be to insert a control rod

withdrawal block and verify that all control rods are fully inserted;

however, the control rods required to be inserted would not apply to

those control rods withdrawn in accordance with LCO 3.10.6, ``Multiple

Control Rod Withdrawal -Refueling.'' The amendment would add an

additional Required Action for Limiting Condition for Operation (LCO)

3.9.1, ``Refueling Equipment Interlocks,'' of the Technical

Specifications (TSs) for Grand Gulf Nuclear Station, Unit 1 (GGNS). The

alternative method then could be used to respond to inoperable

interlocks instead of only the current method of halting in-vessel fuel

movement with equipment associated with the inoperable interlock.

The proposed change does not remove the current Required Action

method for LCO 3.9.1 and does not change the surveillance requirements

on the refueling equipment. The licensee has also provided changes to

the Bases of the TSs for the proposed amendment.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration. The licensee has proposed the amendment for the TSs for

both GGNS and River Bend Station (RBS). References made to the RBS TSs

and to RBS in the licensee's analysis of no significant hazards

consideration have been removed and replaced by [...]. The licensee's

analysis is presented below:

Entergy Operations, Inc. proposes to change the current Grand

Gulf Nuclear Station (GGNS) [...] Technical Specifications. The

specific proposed change adds additional acceptable Required Actions

to the Actions of LCO 3.9.1, ``Refueling Equipment Interlocks,''

[for inoperable interlocks]. The additional Required Actions will

add an alternative [method] to [the current method of] suspending

fuel movement in the reactor vessel when the refueling interlocks

are inoperable. The requested alternative is to insert a control rod

withdrawal block

[[Page 31179]]

immediately and verify all control rods required to be inserted are

fully inserted. [The control rods required to be inserted would not

apply to control rods withdrawn in accordance with LCO 3.10.6,

``Multiple Control Rod Withdrawal--Refueling.'']