Biweekly Notice; 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 January 30, 1999, through February 11, 1999.

The last biweekly notice was published on February 10, 1999.

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administration 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 March 26, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

[[Page 9184]]

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: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: January 28, 1999.

Description of amendment request: The H. B. Robinson, Unit No. 2,

Technical Specifications (TSs) are proposed to be changed to replace

and add analytical methodologies used to determine acceptable core

designs and provide inputs to methodologies that develop the core

operating limits in the Core Operating Limits Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes in a methodology have been previously

generically reviewed and approved for use by the NRC for determining

core neutronics design and gadolinimum oxide thermal conductivity.

Analyzed events are assumed to be initiated by the failure of plant

structures, systems, or components. The fuel design parameters

developed in accordance with the new methodologies are bounded by

the limitations in the NRC acceptance in its safety evaluations of

the new methodologies. The topical reports associated with the new

methodologies demonstrate that the integrity of the fuel will be

maintained during normal operations and that design requirements

preclude fuel rods containing gadolinium oxide from being limiting

in accident and related safety analyses. The proposed change does

not have a detrimental impact on the integrity of any plant

structure, system, or component. The proposed change will not alter

the operation of any plant equipment, or otherwise increase its

failure probability. Therefore, the probability of occurrence for a

previously analyzed accident is not significantly increased.

The consequences of a previously analyzed accident are dependent

on the initial conditions assumed for the analysis, the behavior of

the fuel during the analyzed accident, the availability and

successful functioning of the equipment assumed to operate in

response to the analyzed event, and the setpoints at which these

actions are initiated. The proposed changes to methodology continues

to meet applicable design and safety analyses acceptance criteria

for neutronics design analysis and gadolinimum oxide thermal

conductivity. The topical reports associated with the new

methodologies demonstrate that the integrity of the fuel will be

maintained as is assumed or is bounded initially in accident

analyses and that design requirements preclude fuel rods containing

gadolinimum oxide from being limiting in accident and related safety

analyses. The proposed change does not affect the performance of any

equipment used to mitigate the consequences of an analyzed accident.

As a result, no analyses assumptions are violated and there are no

adverse effects on the factors that contribute to offsite or onsite

dose as the result of an accident. The proposed change does not

affect setpoints that initiate protective or mitigative actions. The

proposed change ensures that plant structures, systems, or

components are maintained consistent with the safety analysis and

licensing bases. Based on this evaluation, there is no significant

increase in the consequences of a previously analyzed event.

Therefore, the proposed change does not involve any 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 alteration of

plant systems, structures, or components. The proposed changes in

methodology continue to meet applicable criteria for neutronics

design analysis and assure that design requirements preclude fuel

rods containing gadolinimum oxide from being limiting. The proposed

change does not involve a physical alteration of the plant other

than allowing for fuel design in accordance with NRC approved

methodologies. No new or different equipment is being installed. No

installed equipment is being operated in a different manner. There

is no alteration to the parameters within which the plant is

normally operated or in the setpoints that initiate protective or

mitigative actions. As a result no new failure modes are being

introduced. There are no changes in the methods governing normal

plant operation, nor are the methods utilized to respond to plant

transients altered. Therefore, the proposed change does not create

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

accident previously evaluated.

3. Does this change involve a significant reduction in a margin

of safety?

The margin of safety is established through the design of the

plant structures, systems, and components, through the parameters

within which the plant is operated, through the establishment of the

setpoints for the actuation of equipment relied upon to respond to

an event, and through margins contained within the safety analyses.

The proposed change is to methodologies that continue to meet

applicable criteria for neutronics design analysis and continues to

assure that design requirements preclude fuel rods containing

gadolinimum oxide from being limiting. The proposed change does not

impact the condition or performance of structures, systems,

setpoints, and components relied upon for accident mitigation. The

proposed change does not significantly impact any safety analysis

assumptions or results. Therefore, the proposed change does not

result in a significant reduction in the margin of safety.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

[[Page 9185]]

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Cecil B. Thomas.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: November 25, 1998.

Description of amendment request: The proposed amendments would

revise Improved Technical Specifications 3.8.4 and 3.8.9 to support on-

line replacement of the Braidwood 125 Volt DC AT&T batteries with new

Charter Systems Inc. batteries.

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:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

During the replacement of the existing batteries, a temporary

battery bank will provide the same function as the AT&T batteries

being removed. Even though this temporary battery will not be

seismically mounted, due to its location in the Turbine Building, it

is the safety related AT&T battery which was previously qualified

and used to perform this function on Unit 1.

While the temporary battery is being connected, the DC bus will

be supplied by the existing crosstie with Unit 1. Similar crosstie

conditions are allowed under the present Improved Technical

Specifications.

The DC system is normally supplied by the AC system through the

ESF [Engineered Safety Feature] battery charger. The essential

function of the DC system battery is to supply control power

necessary to start and load the Diesel Generators. Once the Diesel

Generators are on line, the DC system will be supplied via the

battery charger. However, the ESF batteries have been sized for one

hour to provide additional assurance that the critical DC loads are

available in the event of a loss of a battery charger.

During the 10 day Completion Time when the temporary battery and

the ESF charger are supporting the bus, the ability of that DC

Division to mitigate an event/accident is unchanged except for its

ability to cope with a seismic event. However, the probability of a

seismic event concurrent with the 10 day Completion Time is

extremely small. During a seismic event, one DC division may be

compromised, however, the unit has adequate DC power available in

the form of the other division to mitigate all Design Basis

accidents. This loss of one DC division is bounded by the loss of an

entire AC division, a condition which the plant is currently

evaluated to withstand.

During the 8 hour Completion Time to connect and disconnect the

temporary battery, there is no adverse impact on Unit 1. The

compensatory measures to manually open the crosstie will ensure the

Unit 1 DC battery can supply its required loads for the entire one

hour duty cycle. The Unit 2 DC bus, which is crosstied, will be de-

energized in the event of a Unit 2 accident based on the

compensatory measures. This action would only be required if the

associated Diesel Generator were to fail to re-energize its

associated charger. This condition is consistent with the other

crosstie scenarios currently permitted by the Technical

Specifications. Thus, the 8 hour Completion Time is consistent with

the two hour Completion Time with respect to the ability to safely

shutdown the Unit. Only the duration of the Completion Time is

different.

Based on the above, the overall design, function, and operation

of the DC system and equipment has not been significantly modified

by these changes. The proposed changes do not affect any accident

initiators or precursors and do not alter the design assumptions for

the systems or components used to mitigate the consequences of an

accident as analyzed in UFSAR Chapter 15.

Therefore, this proposed amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

During the replacement of the existing batteries, a temporary

battery bank will provide the same function as the batteries being

removed. Even though this temporary battery is not seismically

mounted, it is the safety related AT&T battery which was previously

qualified and utilized to perform this function on Unit 1. Because

this temporary battery is identical to the battery that is currently

installed, and will be connected and used in the same way, no new

electrical or functional failure modes are created.

The temporary battery will be located in the turbine building,

which is non-seismic. The temporary battery will not be seismically

mounted. Thus, a seismic failure of the batteries is possible. Since

the temporary battery is located in the turbine building the

potential for battery failure to initiate an accident is not

present, and failure of the battery cannot create a different

response from any previously postulated accident.

Due to the location of the main generator in relationship to the

temporary batteries, a turbine blade failure would not hit the

battery unless it penetrated the turbine casing and ricocheted in

the direction of the battery, which is an unlikely scenario due to

the orientation of the temporary battery. Likewise, an unmitigated

Outside Containment Steam Line Break of either unit would be

interrupted by the successful closure of all MSIVs [Main Steam

Isolation Valves] thereby leaving the battery and the DC bus intact

and available. Also any affects of a postulated storm on the turbine

building have been previously addressed and would not change as a

result of the batteries being temporary located there.

While the temporary battery is being connected, the DC bus will

be supplied by the existing crosstie with Unit 1. To prevent any

occurrence on Unit 2 from adversely affecting Unit 1, this crosstie

will be manually disconnected based on specific criteria that may be

indicative of a Unit 2 accident (specifically a Unit 2 LOOP). Once

the crosstie is opened, the Unit 2 bus will be de-energized and the

other Unit 2 division will be required to mitigate the accident.

This loss of one DC division is bounded by the loss of one division

(AC or DC), a condition which the plant is currently evaluated to

withstand.

The DC system and its equipment will continue to perform the

same function and be operated in the same fashion. The proposed

changes do not introduce any new accident initiators or precursors,

or any new design assumptions for the systems or components used to

mitigate the consequences of an accident. Therefore, the possibility

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

evaluated has not been created.

Therefore, this proposed amendment does not create the

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

previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

During the replacement of the existing batteries, a temporary

safety related battery bank will perform the same function as the

batteries being removed. Even though this temporary battery is not

seismically mounted, it is the safety related battery which was

previously qualified and used to perform this function on Unit 1 and

is identical to the safety related battery that is currently

installed. Therefore, it has the same capacity, margin and

capability to fulfill the requirements of the Unit 2 DC bus as the

existing qualified battery. The proposed replacement activity will

not prevent the plant from responding to either a seismic event or

design basis accident. In both cases, the design mitigation

capability will be maintained. Due to the limited duration of the

activity and the planned contingency actions, a significant

reduction in the margin of safety will not result.

While the temporary battery is being connected, the DC bus will

be supplied by the existing crosstie with Unit 1. This condition is

currently allowed for a limited time by the Improved Technical

Specifications.

The inherent design conservatism of the DC system and its

equipment has not been altered. The DC system and its equipment will

continue to be operated with the same degree of conservatism.

Accordingly, 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

[[Page 9186]]

satisfied. Therefore, the NRC staff proposes to determine that the

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Wilmington Public Library, 201

S. Kankakee Street, Wilmington, Illinois 60481.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Project Director: Stuart A. Richards.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will

County, Illinois

Date of amendment request: December 29, 1998.

Description of amendment request: The proposed amendments would

revise the Technical Specification Tables 3.3.1-1 and 3.3.2-1, to

revise twelve Reactor Trip System and Engineered Safety Feature

Actuation System Allowable Values.

Basis for proposed no significant hazards consideration

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

provided its analysis of the 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?

These changes to the twelve AVs [Allowable Values] do not

involve an increase in the probability of an accident previously

evaluated. The AVs provide the basis for determining instrument

channel operability and do not change the system function, or

channel operation or calibration. Operation within the AV ensures

the instrument channel's ability to provide the required reactor

trip or engineered safety feature actuation signal during plant

operation. In all cases, the proposed changes only make the twelve

AVs more restrictive with respect to the current AVs, and do not

effect the response characteristics of the instrumentation because

actual trip setpoints are unchanged. There is no change being made

to the approved design, nor is there any operational change being

made which would increase the probability of occurrence of an

accident previously evaluated. The RTS [Reactor Trip System] and

ESFAS [Engineered Safety Feature Actuation System] systems which are

actuated by the corresponding instrumentation setpoints will operate

in the same manner as before and within their design limits.

These changes to the twelve AVs do not involve an increase in

the consequences of an accident previously evaluated. These changes

have no effect on plant operation. There is no physical or

operational change being made which would alter the sequence of

events, plant response, or assumptions or conclusions of the

affected analyses. The use of the AVs as a basis for determining

instrument or channel operability does not change system operation

or channel function. The proposed changes do not change the

established trip setpoints for these functions. No design analyses

have changed or will be affected. The twelve revised AVs are more

restrictive than the current AVs and continue to ensure that the

safety limits are not violated during anticipated transients, and

that the consequences of design basis accidents remain acceptable.

The change to the AVs does not degrade or prevent any actions from

taking place in response to an accident. The use of NRC approved or

endorsed methodology in developing the proposed AVs ensures that the

present analytical limits for all accidents will be maintained.

These proposed changes to the AVs for RTS and ESFAS instrumentation

will continue to ensure that the associated RTS trip or ESFAS

actuation signals will be generated when required within the bounds

of the plant safety analyses. There is no change in the type or

amount of any effluents released, and no change in either the onsite

or offsite dose consequences as a result of this change.

Therefore, based on this evaluation, this proposed amendment

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?

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

different kind of accident from any accident previously evaluated.

The proposed changes to the twelve AVs for RTS and ESFAS

instrumentation will not affect the trip setpoints at which a

reactor trip or engineered safety feature actuation is initiated.

The trip setpoints contained in the Technical Requirements Manual

are not being changed and will continue to be maintained. The only

changes being made are to the AVs used as a basis for determining

instrument channel operability. Because the trip setpoints are

unchanged, RTS or ESFAS setpoint actuation is not affected by the

revised AVs.

An RTS trip or ESFAS actuation signal that may initiate between

its trip setpoint and the associated AV is acceptable because an

allowance has been made in the affected instrument uncertainty

calculation to accommodate this deviation. It allows for potential

drift while ensuring plant operation in a safe manner. Using this

methodology provides plant operational flexibility and yet remains

within the allowances accounted for in the various accident

analyses. No new equipment is being installed, and no installed

equipment is being operated in a new or different manner with these

twelve AV changes. The revised AVs do not alter the intended design

or operation of systems or instrument channels.

As no physical plant equipment changes are being made, no new

equipment failure modes are being introduced as a result of these

proposed changes. There is no change in plant operation that affects

previously evaluated failure modes and no change in plant response

to a transient condition. These changes do not represent a new

failure mode over what has been previously evaluated.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

There is no significant reduction in the margin of safety from

these proposed changes. These proposed changes move twelve AVs

closer to the trip setpoints compared to the existing AVs, which

increases the margin of safety. An RTS trip or ESFAS actuation

signal that may initiate between its trip setpoint and the

associated AV is acceptable because an allowance has been made in

the affected instrument uncertainty calculation to accommodate this

deviation. The revised AVs have been calculated using NRC approved

or endorsed methodology, which is consistent with existing safety

analyses that define the margin of safety. Safety analyses

assumptions and results are not affected.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

requested amendments involve no significant hazards consideration.

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Project Director: Stuart A. Richards.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: January 21, 1999.

Description of amendment request: This amendment request proposes

to relocate Technical Specification (TS) Section 3/4.6.I to the Updated

Final Safety Analysis Report (UFSAR) and plant procedures. TS Section

3/4.6.I contains reactor coolant chemistry limiting conditions for

operation (LCO) and surveillance requirements (SR) for conductivity,

chloride concentration and pH.

[[Page 9187]]

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:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes simplify the TS, meet regulatory

requirements for relocated TS's, and implement the recommendations

of the Commission's Final Policy Statement on TS improvements. The

Chemistry requirements will be relocated to the Updated Final Safety

Analysis Report (UFSAR) and to applicable station procedures. Future

changes to these requirements will be controlled by 10 CFR 50.59.

The proposed changes are administrative in nature and do not involve

any modification to any plant equipment or affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of any previously

evaluated accident.

Consequently, this proposed amendment does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes are administrative in nature, do not

involve any physical alterations to any plant equipment, and cause

no change in the method by which any safety related system performs

its function. Therefore, this proposed TS amendment will not create

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

accident previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed amendment represents the relocation of current

requirements which are based on generic guidance or previously

approved provisions for other stations. The proposed changes are

administrative in nature and do not adversely affect existing plant

safety margins or the reliability of the equipment assumed to

operate in the safety analysis. The proposed changes have been

evaluated and found to be acceptable for use at Quad Cities Nuclear

Power Station. Since the proposed changes are administrative in

nature, and are based on NRC accepted provisions which have been

adopted at other nuclear facilities, and maintain the necessary

levels of system reliability, the proposed changes do not involve a

significant reduction in the margin of safety.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Project Director: Stuart A. Richards.

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,

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

Date of amendment request: January 28, 1999.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) to correct Surveillance

Requirement (SR) 3.7.13.4 and the associated Bases. This SR currently

is incorrect and does not reflect the Fuel Handling Ventilation Exhaust

System (FHVES) as designed. Specifically, the FHVES flow rate

requirement has been inadvertently stated at half the design value

(18,221 instead of 36,443 cfm [cubic feet per minute]). The proposed

amendments would only revise the SR to the correct design value; no

physical change to the FHVES design is involved.

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:

First Standard

Implementation of this amendment would not involve a significant

increase in the probability or consequences of an accident

previously evaluated. Approval of this amendment will have no effect

on accident probabilities or consequences. The FHVES is not an

accident initiating system; therefore, there will be no impact on

any accident probabilities by the approval of this amendment. The

design of the system is not being modified by this proposed

amendment. The amendment merely aligns TS requirements with the

existing design and function of the system. Therefore, there will be

no impact on any accident consequences.

Second Standard

Implementation of this amendment would not create the

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

previously evaluated. No new accident causal mechanisms are created

as a result of NRC approval of this amendment request. No changes

are being made to the plant which will introduce any new accident

causal mechanisms. This amendment request does not impact any plant

systems that are accident initiators; neither does it impact any

accident mitigating systems.

Third Standard

Implementation of this amendment would not involve a significant

reduction in a margin of safety. Margin of safety is related to the

confidence in the ability of the fission product barriers to perform

their design functions during and following an accident situation.

These barriers include the fuel cladding, the reactor coolant

system, and the containment system. The performance of these fission

product barriers will not be impacted by implementation of this

proposed amendment. The FHVES is already capable of performing as

designed. No safety margins will be impacted.

Based upon the preceding analysis, Duke Energy has concluded

that the proposed amendment does not involve a significant hazards

consideration.

The staff reviewed the licensee's analysis, and agrees that the

three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC

staff proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R. Newton, Legal Department

(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,

North Carolina.

NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: December 16, 1998, supplemented January

25, 1999.

Description of amendment request: The proposed amendments would

completely replace the High Pressure Injection (HPI) section of the

Improved Technical Specifications that were issued on December 16,

1998. The proposed changes would: (1) expand the applicability for the

requirements regarding the third HPI pump, discharge crossover valves,

and the HPI suction headers; (2) specify the HPI conditions and allowed

times that require the discharge headers be cross-connected or

separated; (3) incorporate limiting conditions for operation when

specified equipment was inoperable during specified plant conditions;

(4) specify changes in HPI system discharge path valve lineup when

certain equipment is inoperable; (5) change the requirement to reduce

reactor power when an HPI system is inoperable from 60 percent power to

75 percent power and specify the length of time operation may continue

at this power level; (6) address the failure to cross-connect the HPI

[[Page 9188]]

discharge headers as an independent condition; (7) add a requirement to

verify by administrative means that the Atmospheric Dump Valve flow

path for each steam generator is operable every 12 hours under certain

conditions; (8) add a requirement that the HPI pump and crossover

valves be restored to operable status within 30 days; (9) delete the

requirement to restore the capability to automatically actuate the HPI

within 24 hours; (10) add a Required Action to reduce reactor power to

less than or equal to 75 percent power within 3 hours in the event an

HPI train cannot be actuated by automatic or manual means; (11) expand

the Completion Time for restoring an inoperable HPI train to 72 hours;

(12) require that Limiting Condition for Operation 3.0.3 be entered

immediately if two HPI trains or two HPI (low pressure injection) -LPI

flow paths are inoperable; (13) change the surveillance requirement to

manually cycle open each LPI-HPI flow path discharge valve every 18

months to require that the HPI discharge crossover valves be cycled

every 18 months; and (14) add or modify various administrative and

Bases changes that support the proposed changes. The licensee supplied

data resulting from risk-informed analyses that were performed in

accordance with Regulatory Guides 1.174 and 1.177 to support the

evaluation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated:

No. The proposed change do not involve a physical alteration of

the plant. No new or different equipment is being installed, and no

installed equipment is being operated in a new or different manner.

No set points for parameters which initiate protective or mitigative

action are being changed.

The proposed changes do not have any impact upon the ability of

the HPI [High Pressure Injection] System to add soluble poison to

the Reactor Coolant System. The remaining potential impact is upon

the ability to mitigate the consequences of a small break LOCA

[Loss-of-Coolant Accident], which is addressed below. The small

break LOCA is the limiting design basis accident with respect to HPI

System operability requirements.

The Technical Specification requirements for the HPI System are

supported by a spectrum of small break LOCA analyses based on the

approved Evaluation Model described in FTI [Framatome Technologies

Incorporated] topical report BAW-10192PA. These small break LOCA

analyses demonstrate that the acceptance criteria of 10 CFR 50.46

are satisfied.

The requirements of LCO [Limiting Condition for Operation] 3.5.2

assure that flow can be provided via two HPI trains (i.e., one HPI

train responds automatically upon an ESPS [Engineered Safeguards

Protective System] signal, and the second HPI train is aligned

within 10 minutes via operator actions in the Control Room)

following a small break LOCA and a single active failure. The full

power small break LOCA analyses supporting this proposed license

amendment have been performed in accordance with the approved

Evaluation Model described in FTI topical report BAW-10192P.

If enhanced steam generator cooling is not credited in the

accident analysis, two HPI trains are required to mitigate specific

small break LOCAs with Thermal Power [less than or equal to] 75% RTP

[Reactor Thermal Power]. However, if equipment not qualified as QA-1

(i.e., an ADV [Atmosphic Dump Valve] flow path for one steam

generator) is credited for enhanced steam generator cooling, the

safety analyses have determined that the capacity of one HPI train

is sufficient to mitigate a small break LOCA on the discharge of the

reactor coolant pumps if Thermal Power [less than or equal to] 75%

RTP. An ADV flow path for each steam generator is credited as a

compensatory measure in Actions B and C of LCO 3.5.2 to permit

operation to continue with THERMAL POWER [less than or equal to] 75%

RTP: a) for 30 days with an HPI pump of one or more HPI discharge

crossover valve(s) inoperable; and b) for 72 hours with one HPI

train inoperable. This provides additional defense-in-depth, because

the ADV flow path for each steam generator is required to be

operable while only one is needed to perform the function.

Additionally, a risk-informed assessment (provided as Attachment 7

to Duke's license amendment request dated December 18, 1998)

concluded that operating the plant in accordance with the Required

Actions was acceptable.

The proposed changes involve crediting an additional operator

action (i.e., steaming that steam generator through an ADV flow

path) that has not previously been reviewed and approved by the

staff for licensing basis small break LOCA analyses. Additionally,

while the EFW System has been credited in past SBLOCA [small break

LOCA] analyses as described in responses to NUREG-0565, actions to

raise steam generator levels to the loss of subcooled margin

setpoint were only assumed in the smaller SBLOCAs. These operator

actions have been included in the Emergency Operating Procedure

(i.e., AP/1, 2, or 3/A/1800/001) for many years.

The times for completing these operator actions (i.e., feeding a

steam generator via EFW [Emergency Feedwater] and steaming that

steam generator through an ADV flow path) are new to the small break

LOCA analysis and the licensing basis, and are considered

reasonable. Crediting the performance of these operator actions

within the specified time frames in the SBLOCA analyses does not

result in any substantive change to the operator's response to [an]

SBLOCA.

In summary, the technical analyses described in this license

amendment justify the adequacy of this specification and assure that

operability of the HPI System is maintained in a manner consistent

with the requirements of the design basis accidents. Therefore, it

is concluded that this amendment request will not significantly

increase the probability or consequences of an accident previously

evaluated.

(2) Create the possibility of a new or different kind of

accident from any kind of accident previously evaluated:

No. The proposed changes do not involve a physical alteration of

the plant. No new or different equipment is being installed, and no

installed equipment is being operated in a new or different manner.

No set points for parameters which initiate protective or mitigative

action are being changed. As a result, no new failure modes are

being introduced.

The requirements of ITS [Improved Technical Specification] 3.5.2

continue to assure that operability of the HPI System is maintained

in a manner consistent with the requirements of the design basis

accidents. The requirements are supported by small break LOCA

analyses which demonstrate that the acceptance criteria of 10 CFR

50.46 are satisfied.

The proposed change involve crediting an additional operator

action (i.e., steaming that steam generator through an ADV flow

path) that has not previously been reviewed and approved by the

staff for licensing basis small break LOCA analyses. Additionally,

while the EFW System has been credited in past SBLOCA analyses as

described in responses to NUREG-0565, actions to raise steam

generator levels to the loss of subcooled margin setpoint were only

assumed in the smaller SBLOCAs. These operator actions have been

included in the Emergency Operating Procedure (i.e., AP/1, 2, or 3/

A/1800/001) for many years.

The times for completing these operator actions (i.e., feeding a

steam generator via EFW and steaming that steam generator through an

ADV flow path) are new to the small break LOCA analysis and the

licensing basis, and are considered reasonable. Crediting the

performance of these operator actions within the specified time

frames in the SBLOCA analyses does not result in any substantive

change to the operator's response to [an] SBLOCA.

Therefore, this proposed amendment will not create the

possibility of any new or different kind of accident.

(3) Involve a significant reduction in a margin of safety.

No. The requirements of ITS 3.5.2 continue to assure that

operability of the HPI System is maintained in a manner consistent

with the requirements of the design basis accidents. The

requirements are supported by small break LOCA analyses which

demonstrate that the acceptance criteria of 10 CFR 50.46 are

satisfied. These analyses were performed in accordance with the

Evaluation Model described in FTI topical report BAW-10192P.

Therefore, it is concluded that the proposed amendment request

will not result in a significant decrease in the margin of safety.

[[Page 9189]]

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC.

NRC Project Director: Herbert N. Berkow.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: January 18, 1999

Description of amendment request: The proposed amendments would:

(1) delete license condition 2.C.(3) from the Beaver Valley Power

Station, Unit No. 1 (BVPS-1) operating license and delete some

references to two-loop operation from BVPS-1 Technical Specifications

(TSs); (2) revise BVPS-1 and Beaver Valley Power Station, Unit No. 2

(BVPS-2) TS 2.2.1, 3.3.2.1, associated tables 2.2-1 and 3.3.4, and

associated bases, to use consistent format and wording between units;

(3) revise BVPS-1 and BVPS-2 TS 2.2.1, 3.3.2.1, associated tables 2.2-1

and 3.3.4, and associated bases, to include revised nominal trip

setpoints and allowable values which are more conservative than those

currently listed; (4) delete or revise TS to reflect the current

configuration of Unit 1 plant hardware; and (5) make miscellaneous

editorial changes to BVPS-1 and BVPS-2 TS and associated Bases to

define terms, revise formatting, modify titles, and add license numbers

to pages.

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 [as modified by the NRC staff

based upon information provided elsewhere in the licensee's submittal].

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

This proposed amendment includes changes to nominal Reactor Trip

System (RTS) and Engineered Safety Feature Actuation System (ESFAS)

trip setpoints and allowable values that have been determined with

the use of an approved methodology. The new values ensure that all

automatic protective actions will be initiated at or before the

condition assumed in the safety analysis. This change, which

includes modification of the requirements stated in Limiting Safety

System Setting (LSSS) 2.2.1 and Limiting Condition for Operation

(LCO) 3.3.2.1, will allow the nominal trip setpoints to be adjusted

within the calibration tolerance band allowed by the setpoint

methodology. There will be no adverse effect on the ability of the

channels to perform their safety functions as assumed in the safety

analyses. Since there will be no adverse effect on the trip

setpoints or the instrumentation associated with the trip setpoints,

there will be no significant increase in the probability of any

accident previously evaluated.

Other changes in trip system function, content and format are

proposed based on the current configuration of the trip system

hardware at Beaver Valley Power Station (BVPS) Unit No. 1.

Similarly, since the ability of the instrumentation to perform its

safety function is not adversely affected, there will be no

significant increase in the consequences of any accident previously

evaluated.

Since the safety analysis is unaffected by this change there is

no change in the consequences of any previously evaluated accident.

The editorial changes do not affect plant safety. The

administrative change, for BVPS Unit 1 only, pertaining to two loop

operation and Reactor Coolant System isolation valve position, does

not affect plant safety. The Technical Specification requirements in

LCOs 3.4.1.1 and 3.4.1.4.1 will continue to [prohibit two-loop

operation and] ensure safe plant operation by properly controlling

the operation and position of the reactor coolant loops and Reactor

Coolant System isolation valves.

[The administrative change to delete line item 7.d, pertaining to

Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus

Undervoltage, from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not

affect plant safety because this function is not directly initiated

by bus undervoltage. Rather, the automatic start of the motor-driven

AFW pumps is accomplished by the combination of 1) Emergency Bus

feed breaker opening 2) valid start signal from ESFAS, and 3)

Emergency Diesel Generator (EDG) sequencer actuation. Requirements

for these items are included in the ESFAS related TS, Table 3.3-3

and 3.3-4 items 7.a, 7.c, 7.e, and EDG related TS 4.8.1.1.2.b.3 (b).

Therefore, since there is no change made to the plant hardware or

its operation and requirements related to the AFW pump auto-start

function are maintained elsewhere in the BVPS-1 TS, deleting line

item 7.d from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not

change the probability or consequences of any accident previously

evaluated.]

Therefore, this change does not involve any significant increase

in the probability of occurrence of any 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 amendment includes changes to the format and

magnitudes of nominal trip setpoints and allowable values that

preserve all safety analysis assumptions related to accident

mitigation. The protection system will continue to initiate the

protective actions as assumed in the safety analysis. The proposed

changes to LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that

the trip setpoints are maintained consistent with the setpoint

methodology and the plant safety analysis. This proposed amendment

does not involve additional hardware changes. Plant operation will

not be changed.

Other proposed changes are made so that the Technical

Specifications more accurately reflect the plant-specific trip

system hardware in BVPS Unit No. 1.

Furthermore, the proposed changes do not alter the functioning

of the RTS and ESFAS. Therefore, the proposed change does not create

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

previously evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes do not alter the manner in which safety

limits, limiting safety system settings or limiting conditions for

operation are determined. The proposed RTS and ESFAS trip setpoints

are calculated with an approved methodology. The proposed changes to

LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that the trip

setpoints are maintained consistent with the setpoint methodology

and the plant safety analysis. Therefore, the response of the RTS

and ESFAS to accident transients reported in the Updated Final

Safety Analysis Report is unaffected by this change. No additional

hardware changes are involved. Therefore, accident analysis

acceptance criteria are not affected. Other proposed changes are

made so that the protection system Technical Specifications more

accurately reflect the plant-specific trip system hardware in BVPS

Unit No. 1.

The editorial changes do not affect plant safety. The

administrative change, for BVPS Unit 1 only, pertaining to two loop

operation and Reactor Coolant System isolation valve position, does

not affect plant safety. The Technical Specification requirements in

LCOs 3.4.1.1 and 3.4.1.4.1 will continue to [prohibit two-loop

operation and] ensure safe plant operation by properly controlling

the operation and position of the reactor coolant loops and Reactor

Coolant System isolation valve.

[The administrative change to delete line item 7.d, pertaining to

Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus

Undervoltage, from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not

affect plant safety because this function is not directly initiated

by bus undervoltage. Rather, the automatic start of the motor-driven

AFW pumps is accomplished by the combination of (1) Emergency Bus

feed breaker opening, (2) valid start signal from ESFAS, and (3) EDG

sequencer actuation. Requirements for these items are included in

the ESFAS related TS, Table 3.3-3 and 3.3-4 items 7.a, 7.c, 7.e, and

EDG related TS 4.8.1.1.2.b.3 (b). Therefore, since there is no

change made to the plant hardware or its operation and requirements

related to the AFW pump auto-start function are maintained elsewhere

in the BVPS-1 TS,

[[Page 9190]]

deleting line item 7.d from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2

will not involve a significant reduction in a margin of safety.]

Therefore, operation of the facility in accordance with the

proposed amendment will not involve a significant reduction in a

margin of safety.

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

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: S. Singh Bajwa.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: December 16, 1998.

Description of amendment request: The licensee has proposed an

amendment of Facility Operating License No. NPF-47, Appendix A--

Technical Specifications, Section 2.1.1.2, entitled ``Reactor Core

[Safety Limits].'' The proposed amendment will change the two

recirculation loop Minimum Critical Power Ratio (MCPR) limit from 1.13

to 1.12 and the single recirculation loop MCPR limit from 1.14 to 1.13.

The revised limits are necessary to address the operation of Cycle 9

following the refueling outage which is scheduled to begin April 1999.

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 request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The plant/cycle specific SLMCPRs have been calculated using

methods identical to those used by General Electric (GE) to assess

the SLMCPR for other Boiling Water Reactors (BWRs). Similar methods

were used to determine the value of the SLMCPR for the previous

cycle. These methods are within the existing design and licensing

basis and cannot increase the probability or severity of an

accident. The basis of the SLMCPR calculation is to ensure that

greater than 99.9% of all fuel rods in the core avoid transition

boiling and fuel damage in the event of the occurrence of

Anticipated Operational Occurrences (AOO) or a postulated accident.

The SLMCPR is used to establish the Operating Limit Minimum

Critical Power Ratio (OLMCPR). Neither the SLMCPR nor the OLMCPR are

initiators or affect initiators of an accident previously evaluated

and therefore changes to the SLMCPR do not increase the probability

of any accident previously evaluated. The proposed changes involve

the use of an accepted methodology in calculating the SLMCPR and,

since there is no change in the definition of the SLMCPR, these

changes will not affect the consequences of any accident previously

evaluated. In addition, the proposed changes do not involve any

change in the way the plant is operated. Existing procedures will

ensure that the SLMCPR is not violated. Therefore, these changes

have no effect on the consequences of an accident.

On these bases, there will be no increase in the probability or

consequences of an accident previously analyzed as a result the

proposed changes.

2. The request does not create the possibility of occurrence of

a new or different kind of accident from any accident previously

evaluated.

The proposed changes consist of SLMCPR calculated from an

accepted method of analysis that has been used by many BWRs. These

changes do not involve any alteration of the plant and do not affect

the plant operation. Neither the SLMCPR nor the OLMCPR can initiate

an event, therefore a change to the SLMCPR does not create the

possibility of occurrence of a new or different kind of accident

from any accident previously evaluated.

3. The request does not involve a significant reduction in the

margin of safety.

The SLMCPR is a Technical Specification numerical value to

ensure that 99.9% of all fuel rods in the core will avoid transition

boiling if the limit is not violated. The proposed SLMCPR change

results from SLMCPR analysis using the accepted methods as

identified in the Attachment.

The margin of safety resides between the SLMCPR and the point at

which fuel fails. Maintaining the MCPR above the proposed SLMCPR

will maintain the margin of safety associated with GE's SLMCPR

methodology. Existing plant procedures will continue to ensure that

the SLMCPR is not violated.

Therefore, this request does not involve a 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: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005.

NRC Project Director: John N. Hannon.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 23, 1998.

Description of amendment request: The proposed changes will modify

the Limiting Condition for Operation for Technical Specifications

3.3.3.7.1 for the chlorine detection system at Waterford Steam Electric

Station, Unit 3. A change in the alarm/trip setpoint from 3 parts per

million (ppm) to 2 ppm is requested. Additionally, the proposed request

corrects a typographical error in Table 3.3-4.

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. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: The chlorine detection system has no effect on the

accidents analyzed in Chapter 15 of the Final Safety Analysis

Report. Its only effect is on habitability of the control room,

which will be enhanced by specifying a more conservative setpoint in

the Technical Specifications (TS). Analysis using more conservative

assumptions show that a setpoint of 2 parts per million (ppm)

chlorine is acceptable.

Correcting the typographical error on TS page 3/4 3-19 has no

effect on the probability or consequences of an accident previously

evaluated.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different type of

accident from any accident previously evaluated?

Response: The proposed Technical Specification change in itself

does not change the design or configuration of the plant. Using a

more conservative setpoint performs the same function as the old

setpoint, but it accomplishes this function with increased

conservatism.

Correcting the typographical error on TS page 3/4 3-19 will not

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

any accident previously evaluated.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change

[[Page 9191]]

involve a significant reduction in a margin of safety?

Response: The chlorine detection system has no effect on a

margin of safety as defined by Section 2 of the Technical

Specifications. Its only effect is on habitability of the control

room, which will be enhanced by a more conservative setpoint

provided by this change to the Technical Specifications.

Correcting the typographical error on TS page 3/4 3-19 does not

involve a significant reduction in a 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, LA 70122.

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400

L Street N.W., Washington, D.C. 20005-3502.

NRC Project Director: John N. Hannon.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: January 25, 1999.

Description of amendment request: The proposed change request will

modify Technical Specification (TS) 3.5.1 to allow up to 72 hours to

restore safety injection tank (SIT) to operable status if one SIT is

inoperable due to boron concentration not within the limits or the

inability to verify level and pressure. The proposed change would also

allow up to 24 hours to restore SIT to operable status if one SIT is

inoperable due to other reasons when Reactor Coolant System pressure is

greater than or equal to 1750 psia. The ACTIONS for an inoperable SIT

are being subdivided based on pressurizer pressure to be consistent

with the current Waterford 3 requirements and applicability.

Additionally, the Surveillance requirement to sample the SIT after a 1%

volume increase is being changed to not be required if the source of

the makeup is the refueling water storage pool. This amendment request

is a collaborative effort of participating Combustion Engineering

Owners Group members based on a review of plant operations,

deterministic and design basis considerations, and plant risk, as well

as previous generic studies and conclusions drawn by the NRC Staff and

contained within NUREG-1366, ``Improvements to Technical Specifications

Surveillance Requirements,'' and NUREG-1432, Revision 1, ``Standard

Technical Specifications for Combustion Engineering (CE) Plants.'' TS

Bases 3/4.5.1 will be revised to support above 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:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: The Safety Injection Tanks (SITs) are passive

components in the Emergency Core Cooling System. The SITs are not an

accident initiator in any accident previously evaluated. Therefore,

this change does not involve an increase in the probability of an

accident previously evaluated.

The SITs were designed to mitigate the consequences of Loss of

Coolant Accidents (LOCA). These proposed changes do not affect any

of the assumptions used in deterministic LOCA analyses. Hence the

consequences of accidents previously evaluated do not change.

In order to fully evaluate the affect of the SIT Allowed Outage

Time (AOT) extension from 1 hour to 24 hours when one SIT is

inoperable for reasons other than boron concentration or inability

to measure level or pressure, probabilistic safety analysis (PSA)

methods were utilized. The results of these analyses show no

significant increase in the core damage frequency. As a result,

there would be no significant increase in the consequences of an

accident previously evaluated. These analyses are detailed in CE

NPSD-994, Combustion Engineering Owners Group ``Joint Applications

Report for Safety Injection Tank AOT/STI Extension.''

The proposed change to extend the AOT from 1 hour to 72 hours

when unable to measure level or pressure is acceptable because SIT

operability is not based on instrumentation availability. Therefore,

this does not involve a significant increase in the consequences of

an accident as evaluated and are endorsed by the Nuclear Regulatory

Commission (NRC) in NUREG-1366, ``Improvements to Technical

Specifications Surveillance Requirements.'' The inability to measure

level or pressure is acceptable because the SIT instrumentation

provides no safety actuation.

The AOT extension from 1 hour to 72 hours, based upon boron

concentration outside the prescribed limits does not involve a

significant increase in the consequences of an accident as evaluated

and approved by the NRC in NUREG-1432, ``Standard Technical

Specifications for Combustion Engineering Plants.'' These changes

are acceptable because the reduced concentration effects on core

subcriticality during reflood are minor.

The change in sampling requirements to not require sampling if

the makeup source is of the same concentration limit as the SIT is

acceptable as the concentration will remain within the TS limits.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different kind of

accident from any accident previously evaluated?

Response: The proposed change does not alter the design or

configuration of the plant. It also does not alter the mitigation

capabilities of any safety system or components. This change

increases the AOTs for the condition of SIT inoperability. The boron

concentration is maintained by make-up from a source of water with

the required concentration of the SITs.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: The proposed changes do not affect the limiting

conditions for operation or their bases that are used in the

deterministic analyses to establish the margin of safety. PSA and

deterministic evaluations were used to evaluate these changes. The

PSA evaluations demonstrated that the applicable changes are either

risk neutral or risk beneficial. These evaluations are detailed in

CE NPSD-994. The deterministic evaluations show that the SITs would

be able to perform their safety function. These changes are

consistent with NUREG-1366 and NUREG-1432. The margin of safety is

not significantly affected by makeup from a source of the same

concentration limit as the SIT or increase in the AOT for boron

concentration of one SIT not within limits.

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: John N. Hannon.

[[Page 9192]]

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: January 25, 1999.

Description of amendment request: The proposed changes modify

Technical Specifications Section 6.0 to remove certain administrative

controls and instead rely on the change controls of 10 CFR 50.54(a)(3)

and to add a requirement to Section 6.0 concerning the responsibilities

of the General Manager Plant Operations. The requested changes are

consistent with the Improved Standard Technical Specifications for

Combustion Engineering plants, NUREG-1432.

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. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: The requested changes are purely administrative in

nature. The proposed changes do not affect the operation of any

structures, systems, or components or the assumptions of any

accident analyses. The requested changes only affect Section 6.0 of

the Waterford 3 Technical Specifications which describe the

administrative controls to be implemented at the site. The requested

changes either add an additional administrative requirement or

remove quality assurance program details from the Technical

Specifications. The details are being removed from the Technical

Specifications and instead rely on the change controls of 10 CFR

50.54(a)(3). This submittal makes no changes to the regulatory

controls governing changes. The requested changes are purely

administrative in nature.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different kind of

accident from any accident previously evaluated?

Response: The proposed changes to the Technical Specification

requirements are purely administrative in nature and do not involve

a change in plant design or affect the configuration or operation of

any structure, system, or component.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: The proposed changes do not affect the operation of

any structures, systems, or components or the assumptions of any

accident analyses. The requested changes are purely administrative

in nature.

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: John N. Hannon.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: January 22, 1999.

Description of amendment request: The proposed amendment would

revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)

Section 4.3, ``Fuel Storage,'' by updating the criticality requirements

(k-infinity and U-235 enrichment limits) for storage of fuel assemblies

in the spent fuel racks. This change would allow for storage of nuclear

fuel assemblies with new designs, including GE-12 with a 10X10 pin

array.

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:

After reviewing this proposed amendment, we have concluded:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The probability of occurrence of the accident/abnormal

conditions evaluated in UFSAR Section 9.1.2.3 is not significantly

increased by this change because no modification in fuel handling

equipment, fuel pool cooling equipment, fuel storage racks, or fuel

handling practices is taking place. Only the k-infinity and

enrichment limits for the stored fuel are being changed.

The postulated accident/abnormal conditions evaluated in UFSAR

Section 9.1.2.3 have been re-evaluated for the proposed changes in

k-infinity and enrichment limits. The results demonstrate that the

consequences are negligible. The analyses performed show that the

requirement to maintain K-eff less than 0.95 (substantially

subcritical) is satisfied for normal and postulated abnormal

conditions using methods and assumptions that are consistent with

the existing UFSAR. Seismic adequacy and structural integrity of the

pool and racks are not affected by the introduction of GE-12 fuel.

Local and bulk pool temperatures remain bounded by the current UFSAR

analysis for fuel exposures with GE-12 fuel expected through two

cycles of operation (i.e., through Cycle 18 operation). Based upon a

scoping study comparing the hydraulic diameters of GE-10 and GE-12

fuel, large margins to pool boiling conditions at the final

discharge exposures of GE-12 fuel will be maintained. Therefore, the

consequences of the accident are not significantly increased by this

change.

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

new or different kind of accident from any accident previously

evaluated.

No new types of accidents are being introduced because no

modification in fuel handling equipment, fuel pool cooling

equipment, fuel storage racks or fuel handling procedures is being

made. The design basis function of the spent fuel racks is to

maintain the fuel configuration substantially subcritical and within

allowable temperatures under both normal and postulated abnormal

conditions. This design basis function will be maintained with the

proposed k-infinity and enrichment limits.

3. The proposed amendment will not involve a significant

reduction in a margin of safety.

The margin of safety is not significantly reduced. This margin

is based on the requirement to limit the K-eff of fuel in the spent

fuel racks to less than 0.95. The proposed changes in k-infinity and

enrichment limits have been shown to meet this requirement, using

methods and assumptions that are consistent with the existing UFSAR.

Seismic adequacy and structural integrity of the pool and racks are

not affected by the introduction of GE-12 fuel. Local and bulk pool

temperatures remain bounded by the current UFSAR analysis for fuel

exposures with GE-12 fuel expected through two cycles of operation

(i.e., through Cycle 18 operation). Based upon a scoping study

comparing the hydraulic diameters of GE-10 and GE-12 fuel, large

margins to pool boiling conditions at the final discharge exposures

of GE-12 fuel will be maintained.

Based upon the above, we have determined that the proposed

amendment will not involve a significant hazards consideration.

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: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

[[Page 9193]]

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: Cynthia A. Carpenter.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: October 15, 1998, as supplemented on

December 21, 1998.

Description of amendment request: The proposed amendment would

revise the Duane Arnold Energy Center (DAEC) Technical Specifications

(TS) by adding a new TS 3.7.9, ``Control Building/Standby Gas Treatment

System (CB/SBGT) Instrument Air System.'' The proposed amendment would

also revise (TS) 3.6.1.3, ``Primary Containment Isolation Valves

(PCIVs),'' Condition E, by adding a time limit for plant operation if a

penetration flow path is isolated by a single purge valve with

resilient seal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The amendment is adding new requirements for the CB/SBGT

Instrument Air System that are commensurate with the safety

functions it supports and consistent with other support systems in

the Technical Specifications. These requirements provide appropriate

actions and time limits for plant operation with one or both CB/SBGT

Instrument Air subsystems inoperable. The probability of an event

while in this condition is low, and the consequences are bounded by

the failure of the supported systems. The CB/SBGT Instrument Air

System is not assumed to be an initiator of an analyzed event.

The amendment is also adding a time limit for plant operation if

a purge valve with resilient seal is used to satisfy TS 3.6.1.3

Required Action E.1 (isolate the affected penetration flow path).

While primary containment integrity is provided by the purge valve,

it is prudent to limit operation in this condition due to the

potential for increased leakage from a single active failure.

These additions will provide assurance that affected systems

will be OPERABLE when required and as assumed in the design basis.

This change will not physically alter the plant (no new or

different type of equipment will be installed). This change will not

alter the operation of process variables, structures, systems, or

components as described in the safety analysis. This change will not

alter assumptions relative to the mitigation of an accident or

transient event. This change will not increase the probability of

initiating, or the consequences of an analyzed event.

(2) The proposed amendment will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The amendment adds new requirements for the CB/SBGT Instrument

Air System and adds a time limit for plant operation if a purge

valve with resilient seal is used to satisfy TS 3.6.1.3 Required

Action E.1.

This change will not physically alter the plant (no new or

different type of equipment will be installed). This change will not

alter the operation of process variables, structures, systems, or

components as described in the safety analysis. Thus, a new or

different kind of accident will not be created.

(3) The proposed amendment will not involve a significant

reduction in a margin of safety.

The amendment is adding new requirements for the CB/SBGT

Instrument Air System to provide appropriate actions and time limits

for plant operation with one or both CB/SBGT Instrument Air

subsystems inoperable.

The amendment is also adding a time limit for plant operation if

a purge valve with resilient seal is used to satisfy TS 3.6.1.3

Required Action E.1 (isolate the affected penetration flow path).

While primary containment integrity is provided by the purge valve,

it is prudent to limit operation in this condition due to the

potential for increased leakage from a single active failure in the

remaining OPERABLE components.

This change will not physically alter the plant (no new or

different type of equipment will be installed). This change will not

alter the operation of process variables, structures, systems, or

components as described in the safety analysis. This change will not

alter assumptions relative to the primary success path for

mitigation of an accident or transient event.

These additions will provide assurance that the accident

mitigation functions will perform as assumed in the safety analysis.

Thus, the margin of safety will not be reduced.

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: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: Cynthia A. Carpenter.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,

Unit No. 1, Washington County, Nebraska

Date of amendment request: January 29, 1999.

Description of amendment request: The amendment would revise the

technical specifications (TS) to relocate three cycle-specific

parameter limits; shutdown margin with Tcold>210 deg.F,

moderator temperature coefficient, and minimum boric acid storage tank

level versus concentration, to the Core Operating Limits Report (COLR).

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 safety analysis most impacted by a change to the negative

Moderator Temperature Coefficient (MTC) limit is the Main Steam Line

Break (MSLB) event. The Steam Line Break Cooldown curves for an MTC

are calculated and then input to the cycle-specific MSLB analysis

(if necessary) during the reload analysis process, using an NRC-

approved methodology. The required/acceptable Shutdown Margin (SDM)

is dependent upon the core loading pattern used (i.e., cycle-

specific core physics parameters) and is largely dependent on the

cycle-specific MTC and available scram worth. The SDM is determined

based on the analysis of the Hot Zero Power (HZP) MSLB event in

which the return-to-critical and return-to-power conditions are

evaluated to provide acceptable results. With the ongoing changes in

MTC as a result of core loadings for FCS and higher U-235

enrichments, the end-of-cycle MTC is becoming more negative than the

present Technical Specifications limit. Since the MTC is fuel cycle

specific and influences the required SDM, it is appropriate to move

both of these values to the COLR, consistent with Generic Letter 88-

16. Note that no change to the SDM for Tcold

210 deg.F is being proposed.

The cycle-specific reload analysis is performed for every

operating cycle and the results, as incorporated into the COLR

pursuant to the 10 CFR 50.59 process, are transmitted to the NRC.

FCS will continue to provide COLR updates to the NRC. The relocation

of the negative MTC and the ``BAST level versus BAST Concentration''

curves into the COLR, consistent with the NRC recommendations of

Generic Letter 88-16, will not modify the methodology used in

generating the limits, nor the manner in which they are implemented.

These limits will continue to be determined by analyzing the same

postulated events as previously analyzed. FCS will continue to

operate within the limits specified in the COLR and will take the

same corrective actions when or if these limits are exceeded as

required by

[[Page 9194]]

current Technical Specifications. The potential increase of the

absolute magnitude of the negative MTC with Shutdown Margin decrease

is evaluated during the COLR reload analysis process in accordance

with OPPD's NRC-approved topical report. Therefore, this proposed

amendment is administrative in nature and has been concluded not to

increase the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes to FCS Technical Specifications were the

result of a recommendation from a Generic Letter. Future changes to

the parameters being relocated to the COLR can only be performed

with approved Reload Analyses. No new or different kind of accident

is created by this administrative change because the actual

operation of FCS remains unchanged. Therefore the possibility of an

accident or malfunction of a different type than previously

evaluated in the safety analysis report would not be created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

As indicated above, the implementation of this proposed COLR

change, consistent with the guidance of Generic Letter 88-16, makes

use of the existing safety analysis methodologies and the resulting

limits and setpoints for plant operation. Additionally, the safety

analysis acceptance criteria for operation with this proposed

amendment have not changed from the criteria used in the current

reload analysis. Therefore, the margin of safety as defined in the

bases of Technical Specifications is not reduced.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102.

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502.

NRC Project Director: William H. Bateman.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating

Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: January 12, 1999.

Description of amendment request: The proposed change involves

revising Technical Specification (TS) Section 3/4.4.2, ``Safety/Relief

Valves,'' and TS Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2, to

increase the allowable as-found main steam Safety Relief Valve (SRV)

code safety function lift setpoint tolerance from plus or minus 1% to

plus or minus 3%. This change will also require the as-left SRV code

safety function lift setting to be set within plus or minus 1% of the

specified nominal lift setpoint prior to reinstallation in the plant.

In support of this proposed TS change, the required number of OPERABLE

SRVs in Operational Conditions (OPCONs) 1, 2, and 3 will be changed

from 11 to 12. The number of SRVs in each lift pressure grouping will

remain the same. This proposed TS change does not alter the SRV nominal

lift setpoints or the SRV lift setpoint test frequency currently

specified by TS Section 3/4.4.2. The proposed change does not change

the SRV testing commitment specified in LGS Updated Final Safety

Analysis Report (UFSAR) Chapter 5.2.2.10, ``Inspection and Testing.''

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 Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS changes allow for an increase in the as-found

main steam Safety Relief Valve (SRV) setpoint tolerance from plus or

minus 1% to plus or minus 3%. The proposed changes also reduce the

allowable number of SRVs to be out-of-service from three (3) to two

(2). The proposed changes do not alter the SRV nominal lift

setpoints or SRV lift setpoint test frequency. The actuation of an

SRV is the precursor to the inadvertent opening of a SRV transient,

as discussed in Updated Final Safety Analysis Report (UFSAR) Chapter

15.1.4. Increasing the allowable as-found SRV code safety function

lift setpoint tolerance from plus or minus 1% to plus or minus 3%

does have the potential for the minimum SRV simmer margin to be

reduced from 113.3 psig to 89.9 psig. A reduction in simmer margin

will not directly result in an increase of the probability on an

inadvertent self actuation of an SRV. A reduction in simmer margin

will reduce the seating force which may initiate leakage. However,

this leakage is monitored and corrective actions can be implemented

prior to progressing to the point of the potential of an inadvertent

actuation. This reduction in SRV simmer margin has been evaluated by

the SRV manufacturer and determined to be acceptable; therefore, the

probability of an inadvertent SRV actuation remains unchanged.

Actuation of an SRV is not a precursor for any other event evaluated

in the Safety Analysis Report (SAR).

The proposed TS changes have been evaluated on both a generic

and plant specific basis. The NRC has approved the general approach

of this change; however, implementation is contingent on several

plant specific evaluations. The required plant specific analyses and

evaluations included transient analysis of the anticipated

operational transients (AOTs); analysis of the design basis

overpressurization event; evaluation of the performance of high

pressure systems, motor operated valves, and vessel instrumentation

and associated piping; and evaluation of the containment response

during Loss-of-Coolant Accident (LOCA) and hydrodynamic loads on the

SRV discharge lines and containment. In addition to the plant

specific analyses and evaluations required by the NRC, the following

items were also considered: ECCS/LOCA [Emergency Core Cooling

System] performance, SRV simmer margin, high pressure--low pressure

interfaces, i.e., High Energy Line Break (HELB), Station Blackout

(SBO), and Fire Safe Shutdown (FSSD), and the short term

pressurization phase of an ATWS [anticipated transient without

scram] event. These analyses and evaluations show that there is

adequate margin to the design core thermal limits and reactor vessel

pressure limits using the plus or minus 3% SRV code safety function

lift setpoint tolerance and two (2) SRVs out-of-service. The

analyses and evaluations also show that the operation of the high

pressure injection systems will not be adversely affected, that SRV

discharge piping stresses will not be exceeded, and that the

containment response during a LOCA will be acceptable.

Evaluations of the impact of the proposed change on the

Equipment Important to Safety have been performed and no adverse

conditions were identified. The reactor pressure vessel and attached

systems and piping have been evaluated for the impact of this

proposed TS change. A plant specific analysis has been performed

which indicates that neither the American Society of Mechanical

Engineers (ASME) Code upset limits or the TS Safety Limits for the

reactor pressure vessel will be exceeded for the limiting event,

i.e., Main Steam Isolation Valve (MSIV) closure with flux Scram. The

reactor pressure vessel and attached piping design values will not

be exceeded. The current high pressure--low pressure interface

evaluation utilized nominal SRV setpoints, and therefore, is

unaffected. Therefore, the probability of a malfunction of the

reactor pressure vessel and attached systems and piping is not

increased.

The nuclear fuel has been evaluated for the impact of the

proposed change. Plant specific analyses were performed which

indicate that for all abnormal operational transients adequate

margin to the limiting thermal limit parameter, i.e., Minimum

Critical Power Ratio (MCPR), is maintained. Emergency Core Cooling

System (ECCS)/LOCA performance is maintained adequate to meet the

requirements of 10CFR50.46. Therefore, the probability of the

malfunction of the nuclear fuel is not increased.

The SRVs have been evaluated for the impact of the proposed TS

changes. No physical changes to the SRVs will be made as a result of

the proposed TS changes. Adequate simmer margin will be maintained

with the increased tolerance to ensure that an inadvertent lifting

of a SRV does not occur.

[[Page 9195]]

The increase in SRV discharge flow and reactor vessel pressure due

to the potential for higher SRV lift setpoints are bounded by the

SRV steam flows and reactor vessel pressure currently used in the

evaluation of SRV discharge piping, quencher, quencher support, and

hydrodynamic loads on the suppression pool and submerged structures;

therefore, the probability of a malfunction of a SRV or associated

components and structures is not increased.

The Containment response during a LOCA has been evaluated for

the impact of the proposed change. The major factor in the

Containment response to a LOCA is the rate of reactor vessel water

inventory loss. The rate of reactor vessel water inventory loss is

mainly dependent on reactor decay heat which is not affected by the

proposed change. Therefore, the probability of the malfunction of

the Containment is not increased.

The High Pressure Coolant Injection (HPCI) system has been

evaluated for the impact of the proposed TS changes. The analysis

determined that the HPCI system would not be capable of developing

its design flowrate of 5600 gpm at a reactor pressure of 1205 psig

(lowest SRV nominal setpoint +3% tolerance) unless the HPCI turbine/

pump maximum rated speed was increased. However, increasing the HPCI

turbine/pump maximum rated speed is prevented due to HPCI pump

discharge piping overpressurization concerns. Further analysis has

shown that the HPCI system is capable of meeting its required ECCS

function design flowrate, and its required non-ECCS flowrate,

without any change to the current system operating parameters.

Therefore, the probability of a malfunction of the HPCI System is

not increased.

The Reactor Core Isolation Cooling (RCIC) system has been

evaluated for the impact of the proposed change. The analysis

determined that in order for the RCIC system to be capable of

injecting its design flowrate of 600 gpm at a reactor pressure of

1205 psig (lowest SRV setpoint of 1170 psig +3% tolerance) the

maximum rated speed of the RCIC turbine/pump is required to be

increased from 4575 rpm to 4625 rpm. This increase in the RCIC

turbine/pump maximum rated speed will reduce the margin to the

overspeed trip from 123% to 122.1%. This reduction in the margin to

the overspeed trip is acceptable due to the implementation of plant

Modification P00210, ``RCIC System Startup Transient Improvement,''

which reduced the amount of turbine/pump speed overshoot during

system startup. The RCIC overspeed trip setpoint will not be

changed; therefore, a failure of the RCIC turbine/pump (missile

hazard or system overpressurization) due to overspeed is not

increased. All other RCIC System components will continue to operate

within the currently specified design and operating limits.

Therefore, the probability of a malfunction of the RCIC System is

not increased.

The Standby Liquid Control (SLC) system has been evaluated for

the impact of the proposed change. The SLC system capability of

shutting down the reactor during a postulated event in which all or

some of the control rods cannot be inserted or during a postulated

Anticipated Transient Without Scram (ATWS) event is not impacted by

this proposed change. Therefore, the probability of a malfunction of

the SLCS is not increased.

The Control Rod Drive (CRD) system has been evaluated for the

impact of the proposed change. The CRD system capability of

controlling reactor power during normal plant operation and rapidly

inserting control rod blades (Scram) during abnormal plant

conditions is not impacted by the proposed change. Therefore, the

probability of a malfunction of the CRD system is not increased.

The Reactor Vessel Instrumentation System has been evaluated for

the impact of the proposed change. The Reactor Vessel

Instrumentation System will continue to be operated within the

current design pressure/temperature requirements; therefore, the

probability of a malfunction of the Reactor Vessel Instrumentation

System is not increased.

The LGS, Units 1 and 2, Generic Letter 89-10 Motor-Operated

Valve (MOV) Program has been evaluated for the proposed change. The

LGS MOV Program currently uses SRV nominal setpoints for

differential pressure determinations for valves in which reactor

pressure at the SRV setpoint is limiting. Use of nominal SRV

setpoints is consistent with current industry practice. Therefore,

the probability of a malfunction of a MOV is not increased.

Reducing the number of SRVs allowed to be out-of-service does

not make the consequences of a malfunction of a SRV more severe,

since the number of SRVs required to maintain the reactor vessel

within ASME Code and TS Safety Limits will be maintained OPERABLE.

The proposed change does not result in any changes to the

interactions of any system, structure, or component. All systems,

structures, and components will continue to function as designed.

Therefore, the proposed TS changes do not significantly increase

the probability or consequences of an accident previously evaluated.

2. The proposed TS changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed TS changes allow for an increase in the as-found

SRV setpoint tolerance from plus or minus 1% to plus or minus 3%.

The proposed TS changes also reduce the allowable number of SRVs to

be out-of-service from three (3) to two (2). Generic and plant

specific analyses and evaluations indicate that the plant response

to any previously evaluated event will remain unchanged. All plant

systems, structures, and components will continue to be capable of

performing their required safety function as required by event

analysis guidance.

The proposed TS changes do not alter the SRV nominal lift

setpoints or SRV lift setpoint test frequency. The operation and

response of the affected Equipment Important to Safety is unchanged.

All systems, structures, and components will continue to be operated

within acceptable operating and/or design parameters. No system,

structure, or component will be subjected to a condition that has

not been evaluated and determined to be acceptable using the

guidance required for specific event analysis.

Therefore, the proposed TS changes do not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed TS changes allow for an increase in the as-found

SRV setpoint tolerance from plus or minus 1% to plus or minus 3%.

The proposed TS changes also reduce the allowable number of SRVs to

be out-of-service from three (3) to two (2). The proposed TS changes

do not alter the SRV nominal lift setpoints or SRV lift setpoint

test frequency. The operation and response of the affected Equipment

Important to Safety is unchanged. All systems, structures, and

components will continue to be operated within acceptable operating

and/or design parameters. While the calculated peak reactor vessel

pressure for the ASME overpressure event and the ATWS Pressure

Regulator Failure-Open (PREGO) event are higher than those

calculated without the increase in setpoint tolerance, both are

still within the respective licensing acceptance limits associated

with these events. These licensing acceptance limits have been

determined by the NRC to provide a sufficient margin of safety.

The increase in the RCIC system turbine/pump maximum rated speed

is within the capability of the system design. The reduction in the

margin to the overspeed trip is not a reduction in the margin of

safety, since the operation of the RCIC System has demonstrated

minimal speed overshoot on system initiation due to the installation

of plant Modification P00210, ``RCIC System Startup Transient

Improvement.''

The inability of the HPCI system to be capable of injecting 5600

gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint

of 1170 psig +3% tolerance) is not a reduction in the margin of

safety, since analysis for events that would result in high reactor

vessel pressure indicate that the HPCI System is capable of

providing adequate coolant injection.

The increase in SRV steam flow and reactor vessel pressure does

not reduce the margin of safety associated with the SRVs and

associated components and structures since the increased SRV steam

flow rate and reactor vessel pressure are bounded by the current

design analysis.

The margin of safety for fuel thermal limits and 10CFR50.46

limits is unaffected by the proposed change.

The margin of safety for the Containment is unaffected by the

proposed change.

The capability of the SLC system to perform its safety function

during all required events, using the required guidance for event

analysis, is maintained. Therefore, the proposed changes do not

reduce the margin of safety provided by the SLC system.

Therefore, these proposed TS changes do not involve a

significant reduction in a margin of safety.

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

this

[[Page 9196]]

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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101.

NRC Project Director: William M. Dean.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating

Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: January 25, 1999.

Description of amendment request: The proposed Technical

Specification (TS) Change Request revises the TS Surveillance

Requirement frequencies for Sections 4.8.1.1.2.e.1, 4.8.1.1.2.e.8.a,

and 4.8.1.1.2.e.8.b for the Emergency Diesel Generator maintenance

inspection outages, the 24-hour endurance run, and for the hot restart

test from 18 to 24 months.

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 Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The maintenance inspection interval change and the corresponding

interval change for the associated 24 hour endurance test and hot

restart test which are normally performed in conjunction with the

diesel preventive maintenance overhaul inspections, as well as the

programmatic improvements addressed here do not involve physical

changes that would affect the ability of the EDGs [emergency diesel

generators] to perform their safety function. The Emergency Diesel

Generator System is not an accident initiator.

The Surveillance Testing requirements of Technical Specification

Section 3/4.8 will continue to verify the operability and

reliability of the Emergency Diesel Generator system.

The proposed changes do not affect the ability of the EDGs to

mitigate the consequences of an accident, including the Loss of

Coolant Accident (LOCA) coupled with Loss Of Offsite Power accident

analyses as presented in Chapter 15 of the LGS [Limerick Generating

Station] UFSAR [Updated Final Safety Analysis Report]. EDG

unavailability due mostly to outage inspections is more than 2 times

higher than EDG unplanned unavailability. An extension of the outage

inspection frequency to 24 months will result in increased EDG

availability to mitigate the consequences of a potential accident.

When this program is taken in its entirety the extended maintenance

intervals coupled with the defined enhancements is judged to result

in an overall increase in EDG availability and reliability.

Therefore, the probability or consequences of an accident previously

evaluated is not increased.

2. The proposed TS changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The Emergency Diesel Generator system is not an accident

initiator. The operation and design of the onsite emergency power

system (including the EDGs) is not being changed; only the overhaul

inspection interval coupled with the program improvements and the

corresponding interval change for the associated 24 hour endurance

test and hot restart test, (which are normally performed in

conjunction with the diesel preventive maintenance overhaul

inspections), are changed. The EDG system meets the single failure

criteria at the EDG unit level, i.e., the SAR [safety analysis

report] states that with one EDG failed or out-of-service, the

standby AC system is capable of furnishing sufficient power for the

minimum Class 1E load demand, assuming a limiting design basis

accident has occurred. The proposed changes involve a routine

preventive maintenance and inspection time interval change along

with the corresponding surveillance test interval changes, and also

include programmatic improvements to reduce the likelihood of a

failure of an individual EDG unit; the proposed changes do not

involve any physical design or operational changes that could create

a malfunction extending beyond an individual EDG nor do they

increase the potential for a common-mode EDG failure. Therefore, it

is not possible to create a new or different type of accident

through implementation of these changes.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The changes to bring the frequencies of the EDG overhaul, the 24

hour endurance test and the associated hot restart test into

alignment with the current 2 year operating cycle, and the detailed

programmatic changes to achieve conformance with the FMOG [Fairbanks

Morse Owners Group] recommended maintenance program, will increase

the reliability and availability of the EDG system. This will

enhance the margin of safety as the amount of time the EDGs are out-

of-service will decrease and the system will be single-failure proof

for more clock hours when the nuclear reactor(s) are operating. The

changes discussed here do not result in operation of the emergency

diesel generator system nor any other plant system in a manner

beyond their original design basis, and thus does not reduce any

explicit or implicit Technical Specification 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101.

NRC Project Director: William M. Dean.

Portland General Electric Company, et al., Docket No. 50-344, Trojan

Nuclear Plant, Columbia County, Oregon

Date of application for amendment: February 12, 1997.

Brief description of amendment: The proposed amendment would delete

a portion of the Trojan site from the 10 CFR 50 license when that

portion of the site, designated for use as an independently licensed

spent fuel storage installation (ISFSI), receives a part 72 license.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensees' analysis

against the standards of 10 CFR 50.92(c). The licensee's analysis is

summarized below:

The proposed changes would not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed change is administrative in nature and has no impact on

the probability or consequences of accidents previously evaluated. The

physical structures, systems, and components of the Trojan Nuclear

Plant and the operating procedures for their use are unaffected by this

proposed change. The proposed action would eliminate the ISFSI area

from the Part 50 license when the Part 72 license is issued. The 10 CFR

72 licensing controls for the area will assure an adequate level of

safety for the area during normal operation of the ISFSI and during

abnormal events or accidents. Therefore the proposed Part 50 amendment

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes would not create the possibility of a new or

different kind of accident from any accident previously evaluated. The

proposed action would eliminate the ISFSI area from the Part 50 license

when the Part 72 license is issued. The proposed change is

administrative in

[[Page 9197]]

nature and has no impact on plant systems, structures, or components or

on any procedures for operating the plant equipment. The ISFSI will be

separately licensed under Part 72 and physically separated from the

Part 50 licensed structures and equipment. Therefore, the proposed

change does not create the possibility of a new or different kind of

accident from those previously evaluated.

The proposed changes do not involve reduction in the margin of

safety. The Trojan Permanently Defueled Technical Specifications (PDTS)

contain four limiting conditions of operation that address: 1) Spent

Fuel Water Level, 2) Spent Fuel Pool Boron Concentration, 3) Spent Fuel

Pool Temperature, and 4) Spent Fuel Pool load restrictions. These PDTS

will remain in effect as long as spent fuel is stored in the Spent Fuel

Pool, which is in accordance with their applicability statements. The

ISFSI area is physically separated from the Spent Fuel Pool area and

the Fuel Building and will have no effect on spent fuel water level,

spent fuel pool boron concentration, spent fuel pool temperature, or

loads over the Spent Fuel Pool. The proposed change is administrative

and does not affect plant equipment, operating parameters, or

procedures. Based on the above, the proposed change will not reduce the

margin of safety.

Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library,

Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,

Portland, Oregon 97207.

Attorney for licensees: Leonard A. Girard, Esq., Portland General

Electric Company, 121 S. W. Salmon Street, Portland, Oregon 97204.

NRR Project Director: Seymour H. Weiss.

Portland General Electric Company, et al., Docket No. 50-344, Trojan

Nuclear Plant, Columbia County, Oregon

Date of application for amendment: January 7, 1999.

Brief description of amendment: The proposed amendment would allow

loading and handling of spent fuel transfer and storage casks in the

Trojan Fuel Building.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensees' analysis

against the standards of 10 CFR 50.92(c). The licensee's analysis is

summarized below:

The proposed changes would not involve a significant increase in

the probability or consequences of an accident previously evaluated.

With the permanent cessation of operations, the number of potential

accidents was reduced to those types of accidents associated with the

storage of irradiated fuel and radioactive waste storage and handling.

Additional events were postulated for decommissioning activities due to

the difference in the types of activities that were to be performed.

The postulated accidents in the Defueled Safety Analysis Report (DSAR)

are generally classified as: (1) radioactive release from a subsystem

or component, (2) fuel handling accident and, (3) loss of spent fuel

decay heat removal capability. The postulated events described in the

Decommissioning Plan are grouped as: (1) decontamination,

dismantlement, and materials handling events, (2) loss of support

systems (offsite power, cooling water, and compressed air), (3) fire

and explosions, and (4) external events (earthquake, external flooding,

tornadoes, extreme winds, volcanoes, lightning, toxic chemical

release). These types of accidents are discussed below.

Radioactive release from a subsystem or component involves failure

of a radioactive waste gas decay tank (WGDT) or failure of a chemical

and volume control system holdup tank (HUT). For a failure of a WGDT,

the radioactive contents are assumed to be principally the noble gases

krypton and xenon, the particulate daughters of some of the krypton and

xenon isotopes and trace quantities of halogens. For the failure of a

HUT, the assumptions were full power operations with 1-percent failed

fuel, 40 weeks elapsed since power operation, and 60,000 gallons of

120 deg. F liquid released over a 2-hour period. However, the WGDT's

and HUT's are no longer active and have been emptied. Therefore, cask

loading and transfer activities cannot increase the probability of

occurrence of a failure or the consequence of a failure of the WGDT's

or HUT's.

The fuel handling accident involves a stuck or dropped fuel

assembly that results in damage of the cladding of the fuel rods in one

assembly and the release of gaseous fission products. Spent fuel

handling and loading will involve moving the spent fuel assemblies one

by one, from the Spent Fuel Pool to the baskets which will be located

in the Cask Loading Pit. The fuel handling equipment will be the same

as had been previously analyzed with the exception of special tools

which will be used to manipulate failed fuel. These special tools will

be similar in size and weight to the existing tools used for underwater

manipulation and therefore will not present a new hazard. In addition,

the same administrative controls and physical limitations imposed on

any fuel handling operation will be used for spent fuel loading and

handling. The potential release, 100 percent of gap noble gas, from a

fuel assembly is not affected (although the fission product inventory

in a fuel assembly continues to decrease with time). Thus there is no

increase in the probability of occurrence or consequences of a fuel

handling accident over what would be expected for any routine fuel

handling operation.

The loss of spent fuel decay heat removal capability involves the

loss of forced spent fuel cooling with and without concurrent Spent

Fuel Pool inventory loss. The only requirement to assure adequate decay

heat removal capability for the spent fuel is to maintain the water

level in the Spent Fuel Pool so that the fuel assemblies remain covered

(i.e. the capability to make up water to the Spent Fuel Pool must be

available when required). The potential events which could result in a

loss of spent fuel decay heat removal include external events

(explosions, toxic chemical, fires, ship collision with intake

structure, oil or corrosive liquid spills in the river, cooling tower

collapse, seismic events, severe meteorological events), and internal

events including Spent Fuel Pool makeup water system malfunctions

(Service Water System, electrical power, instrument air). Spent fuel

loading and handling will not require the use of explosive materials

(the gases used for electric arc welding are inert), toxic chemicals or

flammable materials (routine use of contamination control materials is

not considered to present a significant hazard). The probability of

other external events (e.g. cooling tower collapse) is not effected by

the spent fuel handling and loading activities inside the Fuel

Building. Spent fuel loading and handling activities will not directly

interface with the Spent Fuel Pool makeup water systems, therefore does

not affect their probability of failure. (The Cask Loading Pit will be

filled with borated water from the Spent Fuel Pool that will be cooled

by the Spent Fuel Cooling System, but use of this water in the Cask

Loading Pit does not increase the failure probability of

[[Page 9198]]

the Spent Fuel Pool or makeup water systems.) As described in the

licensees' safety evaluation, the safe load path and handling height

limitations will ensure that a load drop does not adversely affect the

Spent Fuel Pool or the makeup water systems. Therefore there is no

significant increase in the probability or consequences of a loss of

spent fuel decay heat removal capability.

The events postulated in the Decommissioning Plan are similar to

the DSAR with the exception of the decontamination, dismantlement, and

materials handling events. Decontamination events involve gross liquid

leakage from in-situ decontamination equipment (e.g. tanks) or

accidental spraying of liquids containing concentrated contamination.

Dismantlement events involve segmentation of components and structures,

or removal of concrete by rock splitting, explosives, or electric and/

or pneumatic hammers. Dismantlement events potentially result in

airborne contamination. Material handling events involve the dropping

of contaminated components, concrete rubble, filters, or packages of

particulate materials. Licensee administrative controls will be

implemented to ensure that spent fuel loading and handling activities

and decommissioning activities will not be performed concurrently if

they interact with each other and could increase the probability or

consequences of a postulated event of accident. Therefore, neither the

probability nor the consequences of decontamination, dismantlement, and

materials handling events will not be significantly increased.

The proposed changes would not create the possibility of a new or

different kind of accident from any accident previously evaluated. As

described in the licensees' safety evaluation the potential accidents

associated with fuel handling and loading were similar to fuel handling

accidents, material handling events and pressurized line break

previously analyzed. Additionally the potential consequences were a

small fraction of Environmental Protection Agency (EPA) Protective

Action Guides (PAG's). Therefore, fuel loading and handling does not

present new or different types of accidents.

The proposed changes do not involve a significant reduction in the

margin of safety. The Trojan Permanently Defueled Technical

Specifications (PDTS) contain four limiting conditions of operation

that address: (1) Spent fuel water level, (2) spent fuel pool boron

concentration, (3) spent fuel pool temperature, and (4) spent fuel pool

load restrictions. These PDTS will remain in effect as long as spent

fuel is stored in the Spent Fuel Pool, which is in accordance with

their applicability statements. The spent fuel loading and handling

activities will not affect these PDTS or their bases.

The Cask Loading Pit, where the spent fuel will be loaded into the

basket, is immediately adjacent to the Spent Fuel Pool. The gate

between the Cask Loading Pit and Spent Fuel Pool will be open to allow

transfer of spent fuel assemblies from storage racks in the Spent Fuel

Pool to the basket in the Cask Loading Pit. Opening the gate between

them will allow free exchange of water between the Cask Loading Pit and

the Spent Fuel Pool. The Cask Loading Pit will be filled with borated

water at approximately the same concentration and temperature as the

Spent Fuel Pool prior to opening the gate. This will maintain the

limiting conditions for operation for Spent Fuel Pool boron

concentration, temperature, and water level and the margin of safety

will not be affected.

Spent fuel loading and handling activities will involve lifting and

moving heavy loads (e.g. transfer cask, basket). Loads that will be

carried over fuel in the Spent Fuel Pool racks and the heights at which

they will be carried will be limited to preclude impact energies over

240,000 in-lbs if the loads were dropped. This is in accordance with

limiting condition for operation 3.1.4 ``Spent Fuel Pool Load

Restrictions.'' With this precaution, the limiting condition for

operation pertaining to load restrictions over the Spent Fuel Pool will

be satisfied and the margin of safety will be unaffected. The safe load

paths for heavy loads being lifted outside the Spent Fuel Pool will be

sufficiently far from the Spent Fuel Pool so as to not have an

interaction in the unlikely event of a load drop. In addition

mechanical stops and electrical interlocks on the Fuel Building

overhead crane will provide additional assurance that heavy loads are

not carried over the Spent Fuel Pool racks.

Based on the above, the spent fuel loading and handling activities

will not reduce the margin of safety.

Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library,

Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,

Portland, Oregon 97207.

Attorney for licensees: Leonard A. Girard, Esq., Portland General

Electric Company, 121 S. W. Salmon Street, Portland, Oregon 97204.

NRR Project Director: Seymour H. Weiss.

Portland General Electric Company, et l., Docket No. 50-344, Trojan

Nuclear Plant, Columbia County, Oregon

Date of application for amendment: January 27, 1999.

Brief description of amendment: The proposed amendment would allow

unloading of spent fuel transfer casks in the Trojan Fuel Building.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The licensee's analysis is

summarized below:

The proposed changes would not involve a significant increase in

the probability or consequences of an accident previously evaluated.

With the permanent cessation of operations, the number of potential

accidents was reduced to those types of accidents associated with the

storage of irradiated fuel and radioactive waste storage and handling.

Additional events were postulated for decommissioning activities due to

the difference in the types of activities that were to be performed.

The postulated accidents in the Defueled Safety Analysis Report (DSAR)

are generally classified as: (1) Radioactive release from a subsystem

or component, (2) fuel handling accident and, (3) loss of spent fuel

decay heat removal capability. The postulated events described in the

Decommissioning Plan are grouped as: (1) Decontamination,

dismantlement, and materials handling events, (2) loss of support

systems (offsite power, cooling water, and compressed air), (3) fire

and explosions, and (4) external events (earthquake, external flooding,

tornadoes, extreme winds, volcanoes, lightning, and toxic chemical

release). These types of accidents are discussed below.

Radioactive release from a subsystem or component involves failure

of a radioactive waste gas decay tank (WGDT) or failure of a chemical

and volume control system holdup tank (HUT). For a failure of a WGDT,

the radioactive contents are assumed to be principally the noble gases

krypton and xenon, the particulate daughters of some

[[Page 9199]]

of the krypton and xenon isotopes and trace quantities of halogens. For

the failure of a HUT, the assumptions were full power operations with

1-percent failed fuel, 40 weeks elapsed since power operation, and

60,000 gallons of 120 deg. F liquid released over a two hour period.

However, the WGDT's and HUT's are no longer active and have been

emptied. Therefore, cask loading and transfer activities cannot

increase the probability of occurrence of a failure or the consequence

of a failure of the WGDT's or HUT's.

The fuel handling accident involves a stuck or dropped fuel

assembly that results in damage of the cladding of the fuel rods in one

assembly and the release of gaseous fission products. Spent fuel cask

unloading will involve moving the spent fuel assemblies one by one,

from the baskets which will be located in the cask loading pit to the

spent fuel pool. The fuel handling equipment will be the same as had

been previously analyzed. In addition, the same administrative controls

on physical limitations imposed on fuel handling and fuel loading

operations will be used for fuel unloading. The potential release, 100

percent of noble gases within the gap, from a fuel assembly is not

affected (although the inventory in a radioactive stored fuel assembly

continues to decrease with time). Thus, there is no increase in the

probability of occurrence or consequences of a fuel handling accident

over what would be expected for any routine fuel handling operation or

loading of fuel into a cask.

The loss of spent fuel decay heat removal capability involves the

loss of forced spent fuel cooling with and without concurrent spent

fuel pool inventory loss. The only requirement to assure adequate decay

heat removal capability for the spent fuel is to maintain the water

level in the spent fuel pool so that the fuel assemblies remain covered

(i.e., the capability to make up water to the spent fuel pool must be

available when required). The potential events that could result in a

loss of spent fuel decay heat removal include external events

(explosions, toxic chemical, fires, ship collision with intake

structure, oil or corrosive liquid spills in the river, cooling tower

collapse, seismic events, and severe meteorological events), and

internal events including spent fuel pool makeup water system

malfunctions (service water system, electrical power, and instrument

air). Spent fuel cask unloading will not require the use of explosive

materials, toxic chemicals or flammable materials (routine use of

contamination control materials is not considered to present a

significant hazard). The probability of other external events (e.g.

cooling tower collapse) is not effected by the spent fuel unloading

activities inside the fuel building. Spent fuel cask unloading

activities will not directly interface with the spent fuel pool makeup

water systems, and therefore does not affect their probability of

failure. (The cask loading pit will be filled with borated water from

the spent fuel pool that will be cooled by the spent fuel cooling

system, but use of this water in the cask loading pit does not increase

the failure probability of the spent fuel pool or makeup water

systems). As described in the licensees' safety evaluation, the safe

load path and handling height limitations will ensure that a load drop

does not adversely affect the spent fuel pool or the makeup water

systems. Therefore, there is no significant increase in the probability

or consequences of a loss of spent fuel decay heat removal capability.

The events postulated in the Decommissioning Plan are similar to

the DSAR with the exception of the decontamination, dismantlement, and

materials handling events. Decontamination events involve gross liquid

leakage from in-situ decontamination equipment (e.g. tanks) or

accidental spraying of liquids containing concentrated contamination.

Dismantlement events involve segmentation of components and structures,

or removal of concrete by rock splitting, explosives, or electric and/

or pneumatic hammers. Dismantlement events potentially result in

airborne contamination. Material handling events involve the dropping

of contaminated components, concrete rubble, filters, or packages of

particulate materials. Licensee administrative controls will be

implemented to ensure that spent fuel cask unloading activities and

decommissioning activities will not be performed concurrently if they

interact with each other and could increase the probability or

consequences of a postulated event of accident. Therefore, neither the

probability nor the consequences of decontamination, dismantlement, and

materials handling events will be significantly increased.

The proposed changes would not create the possibility of a new or

different kind of accident from any accident previously evaluated. As

described in the licensee's safety evaluation the potential accidents

associated with fuel cask unloading were similar to fuel handling

accidents, material handling events and pressurized line break

previously analyzed. Additionally the potential consequences were a

small fraction of Environmental Protection Agency (EPA) Protective

Action Guides (PAGs). Therefore, fuel loading and handling does not

present new or different types of accidents.

The proposed changes do not involve a significant reduction in the

margin of safety. The Trojan Permanently Defueled Technical

Specifications (PDTS) contain four limiting conditions of operation

that address: (1) spent fuel pool water level, (2) spent fuel pool

boron concentration, (3) spent fuel pool temperature, and (4) spent

fuel pool load restrictions. These PDTS will remain in effect as long

as spent fuel is stored in the spent fuel pool, which is in accordance

with their applicability statements. The spent fuel cask unloading

activities will not affect these PDTS or their bases.

The cask loading pit, where the spent fuel will be unloaded from

basket, is immediately adjacent to the spent fuel pool. The gate

between the cask loading pit and spent fuel pool will be open to allow

transfer of spent fuel assemblies from the basket in the cask loading

pit to the storage racks in the spent fuel pool. Opening the gate

between them will allow free exchange of water between the cask loading

pit and the spent fuel pool. The cask loading pit will be filled with

borated water at approximately the same concentration and temperature

as the spent fuel pool prior to initial cask loading. This will

maintain the limiting conditions for operation for spent fuel pool

boron concentration, temperature, and water level and the margin of

safety will not be affected.

Spent fuel cask unloading activities may involve lifting and moving

heavy loads (e.g. transfer cask, basket). Loads that will be carried

over fuel in the spent fuel pool racks and the heights at which they

will be carried will be limited to preclude impact energies over

240,000 in-lbs if the loads were dropped. This is in accordance with

limiting condition for operation 3.1.4 ``Spent Fuel Pool Load

Restrictions.'' With this precaution, the limiting condition for

operation pertaining to load restrictions over the spent fuel pool will

be satisfied and the margin of safety will be unaffected. The safe load

paths for heavy loads being lifted outside the spent fuel pool will be

sufficiently far from the spent fuel pool so as to not have an

interaction in the unlikely event of a load drop. In addition,

mechanical stops and electrical interlocks on the fuel building

overhead crane will provide additional assurance that heavy loads are

not carried over the spent fuel pool racks.

[[Page 9200]]

Based on the above, the spent fuel cask unloading activities will

not reduce the margin of safety.

Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library,

Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,

Portland, Oregon 97207.

Attorney for licensees: Leonard A. Girard, Esq., Portland General

Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.

NRR Project Director: Seymour H. Weiss.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: October 16, 1998, as supplemented

January 28, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to

relocate the Chemical Volume and Control System (CVCS) TS 3.2 from the

TSs to the IP3 Operational Specifications.

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 proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response: Relocation (i.e., removal from TS) of TS 3.2, the

bases and the associated surveillances in Table 4.1-1 (items 12, 26,

and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not

involve a significant increase [in] the probability or consequences

of an accident since the relocation of the Technical Specifications

to administrative controls governed by 10 CFR 50.59 does not affect

the availability or function of charging and boric acid flow paths.

CVCS is not an initiator of an accident (the dilution event is

equipment malfunction that is manually terminated) and the proposed

change does not alter overall system operation, physical design,

system configuration, or operational setpoints. There will be no

significant increase in the consequences of an accident because the

required boration flow paths will continue to be available for

boration to the reactor coolant system.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response: Relocation (i.e., removal from TS) of TS 3.2, the

bases and the associated surveillances in Table 4.1-1 (items 12, 26,

and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not

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

any previously evaluated since it does not alter the overall system

operation, physical design, system configuration, or operational

setpoints. The plant systems for boration are operated in the same

manner as before and, consequently, the relocation does not

introduce any new accident initiators or failure mechanisms and does

not invalidate the existing dilution event response. The boration

function is not an accident initiator.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response: Relocation (i.e., removal from TS) of TS 3.2, the

bases and the associated surveillances in Table 4.1-1 (items 12, 26,

and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not

involve a significant reduction in margin of safety. The relocation

is a change to the administrative controls that are used to assure

system availability and those administrative controls are governed

by 10 CFR 50.59. The manner in which the system is operated does not

change and there is no change to physical design, system

configuration, or operational setpoints. Previous analyses of system

malfunction remain unchanged. The current Technical Specification

does not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion

in the technical specifications.

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: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Director.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek

Generating Station, Salem County, New Jersey

Date of amendment request: December 30, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Limiting Condition for Operation

(LCO) 3.7.3 and Table 3.7.3-1. The proposed changes would modify the

flood protection actions required during periods of elevated river

water level.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed TS revisions related to flood protection TS Action

Statements involve no hardware changes and no changes to existing

structures, systems or components. The proposed changes to the flood

protection TS Action Statements ensure that the supported systems

can perform their required safety functions under worst case design

basis conditions, consistent with limitations imposed by other TS.

The proposed flood protection TS ACTION Statements ensure that the

plant is directed to enter a safe shutdown condition whenever the

capability to withstand worst case design basis conditions is

affected. Since the flood protection changes will still ensure that

the plant remains capable of meeting applicable design basis

requirements and retains the capability to mitigate the consequences

of accidents described in the [Hope Creek] HC [Updated Final Safety

Analysis Report] UFSAR, the proposed changes were determined to be

acceptable. As a result, these changes will neither increase the

probability of an accident previously evaluated nor increase the

radiological dose consequences of an accident previously evaluated.

(2) The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes to the flood protection TS contained in

this submittal will not adversely impact the operation of any safety

related component or equipment. Since the proposed changes involve

no hardware changes and no changes to existing structures, systems

or components, there can be no impact on the potential occurrence of

any accident due to new equipment failure modes. The resulting

operational limits imposed by the flood protection LCO ensure that

the plant can either perform its design basis safety functions or an

appropriately conservative shutdown action statement is entered.

Furthermore, there is no change in plant testing proposed in this

change request that could initiate an event. Therefore, these

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

of accident from any accident previously evaluated.

(3) The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes for the flood protection TS retain the

plant's continued capability to withstand worst case design basis

conditions. The proposed flood protection TS ACTION Statements

ensure that the plant is directed to: (1) enter a safe shutdown

condition whenever the capability to withstand worst case design

basis conditions is lost; or (2) enter a conservatively short period

of continued operation when supported system redundancy is reduced.

Since the plant will still remain capable of meeting all applicable

design basis requirements and retaining the

[[Page 9201]]

capability to withstand worst case design basis events described in

the HC UFSAR, the proposed changes were determined to not result in

a significant reduction in a margin of safety.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: William M. Dean.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Description of amendment request: The proposed changes revise the

descriptive details of Technical Specification 4.7.1.2.1.a, regarding

performance testing of the Auxiliary Feedwater (AFW) pumps, to more

closely adhere to NUREG-1431, Improved Standard Technical

Specifications for Westinghouse Plants. This involves relocating the

surveillance-required numerical values for the AFW pump performance

test discharge pressure and flow rate to the South Texas Project

Updated Final Safety Analysis Report (UFSAR).

Date of amendment request: January 20, 1999.

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, which relocates descriptive details (i.e.,

numerical values for AFW pump discharge pressure and flow rate) of

the surveillance testing applicable to the AFW pumps, does not

involve a significant increase in the probability or consequences of

an accident previously evaluated. The affected AFW pump testing

pressure and flow descriptive details that are being removed from

SRs 4.7.1.2.1.a.1 and 4.7.1.2.1.a.2 are not related to any assumed

initiators of analyzed events and are not assumed to mitigate

accident or transient events. The requirement to perform testing on

a monthly, staggered basis is not altered by the proposed change,

and will remain in the Technical Specifications. The descriptive

details of the surveillance testing will be relocated from the

Technical Specifications to the USFAR and will be maintained

pursuant to 10CFR50.59. The proposed revised wording of SRs

4.7.1.2.1.a.1 and 4.7.1.2.1.a.2 (i.e., to verify the developed head

of each pump is greater than or equal to the required developed

head) and the relocation of pump testing details to the UFSAR is

consistent with the AFW pump test requirements in NUREG-1431. In

addition, the surveillance testing details are addressed in existing

surveillance procedures that are also controlled by 10CFR50.59 and

subject to the change control provisions imposed by plant

administrative procedures, which endorse applicable regulations and

standards. Therefore, this proposed 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 relocates descriptive details (i.e.,

numerical values for AFW pump discharge pressure and flow rate) of

surveillance testing applicable to the AFW pumps, which do not meet

the criteria for inclusion in Technical Specifications as identified

in 10CFR50.36(c)(3). The requirement to perform testing on a

monthly, staggered basis is not altered by the proposed change, and

will remain in the Technical Specifications. Additionally,

relocation of the descriptive testing details is consistent with the

wording of the AFW pump test requirements in NUREG-1431, which does

not specify minimum numerical pressure and flow limits. The proposed

change does not involve a physical alteration of the plant (no new

or different type of equipment will be installed) or make changes in

the methods governing normal plant operation. The change will not

impose different requirements, and any future changes to these

relocated surveillance testing details or to the applicable

surveillance procedures will be evaluated per the requirements of

10CFR50.59. This change will not alter assumptions made in the

safety analysis and licensing basis. Therefore, this change does not

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

any accident previously evaluated.

3. Does this change involve a significant reduction in a margin

of safety?

The proposed change, which relocates descriptive details (i.e.,

numerical values for AFW pump discharge pressure and flow rate) of

the surveillance testing applicable to the AFW pumps, will not

reduce a margin of safety since it has no impact on any safety

analysis assumptions. The requirement to perform AFW pump testing on

a monthly, staggered basis will not be altered by the proposed

change, and will remain in the Technical Specifications.

Furthermore, the proposed change will not affect the operability

requirements of the AFW system as delineated in Specification

3.7.1.2. Since any future changes to these relocated surveillance

testing details or to the applicable surveillance procedures will be

evaluated per the requirements of 10CFR50.59, there is no reduction

in a margin of safety. Finally, this proposed change is also

consistent with NUREG-1431, previously approved by the NRC Staff.

Revising the Technical Specifications to reflect the approved NUREG-

1431 content ensures no significant reduction in the margin of

safety. Therefore, this 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 standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.

NRC Project Director: John N. Hannon.

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: January 15, 1999 (TS 98-07).

Brief description of amendments: The proposed amendments would

change the Sequoyah (SQN) Technical Specification (TS) requirements by

adding a new action statement to TS 3.1.3.2, ``Position Indicating

Systems--Operating,'' that eliminates the need to enter TS 3.0.3

whenever two or more individual rod position indicators (RPIs) may be

inoperable per bank, while maintaining the appropriate overall level of

protection and adding flexibility to the initial determination of the

position of the non-indicating rod(s).

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), Tennessee Valley

Authority (TVA), the licensee, has provided its analysis of the issue

of no significant hazards consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change to TS 3.1.3.2 does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The potential for the new action statement to

impact the probability or consequences of the safety analyses for

the plant lies only in the area of operator-exacerbated reactivity

events due to

[[Page 9202]]

a loss of RCCA [rod control cluster assembly] position indication.

RCCA events such as: One or more dropped RCCAs, a dropped RCCA

bank or a RCCA ejection (FSAR [Final Safety Analysis Report]

Sections 15.2.3 and 15.4.6, respectively) are not impacted since the

new action statement does not involve a design change. Events such

as: Uncontrolled RCCA bank withdrawal at power, statically

misaligned RCCA or withdrawal of a single RCCA (FSAR Sections

15.2.2, 15.2.3, and 15.3.6, respectively) involve, or potentially

involve, operator action and are of interest. The uncontrolled RCCA

bank withdrawal at power is an ANS [American Nuclear Society]

Condition II transient that has been analyzed using a positive

reactivity insertion rate greater than that for the simultaneous

withdrawal of the two control banks having the maximum combined

worth at maximum speed. Whether the event is caused by a failure in

the rod control system or by operator error has no effect on the

positive reactivity insertion rate assumed in the analysis. The

protection systems assumed in the analysis are unaffected since

there is no change to the design. Loss of the RPIS would not result

in more frequent control rod movement by plant operators. Therefore,

the new action statement would not affect the analysis of this event

and departure from nucleate boiling ratio (DNBR) design basis would

still be met.

The most severe misalignment situation, with respect to DNBR,

arises from cases in which one RCCA is fully inserted or where Bank

D is fully inserted to its insertion limits with one RCCA fully

withdrawn. For these cases, as discussed in FSAR Section 15.2.3.2,

the DNBR remains above the safety analysis limit values. Also, the

control bank insertion limit alarms remain available to warn

operators that bank insertion limits have been reached.

A compensatory action associated with this new action statement,

placing the control rods under manual control, addresses concerns

associated with automatic rod motion due to the rod control system

and inadvertent operator contribution to these events.

The worst-case event of those described above, the withdrawal of

a single RCCA, is an ANS Condition III event. It has been analyzed

in FSAR Section 15.3.6, assuming that operators ignore RCCA position

indication or that multiple rod control system failures occur. No

single electrical or mechanical failure in the rod control system

could cause the accidental withdrawal of a single RCCA from an

inserted bank at full power operation. The operator could

deliberately withdraw a single RCCA in the control bank. This

feature is necessary in order to retrieve an accidentally dropped

rod. This new action statement does not change the plant design;

therefore, there would be no change in the probability of the event

being induced by the unlikely, simultaneous electrical failures

(FSAR Section 7.7.2.2).

The change in the time to determine the position of the non-

indicating rods, indirectly with the movable incore detectors, does

not involve a design change nor does it affect the immediate

response of the operator to the event, therefore, it does not affect

the results of the analyses described above.

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

new or different kind of accident from any accident previously

evaluated.

Since there is no change to the design associated with the

proposed change, it does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change involves a loss of the RPIS [Rod Position

Indication System] and establishes compensatory measures to maintain

control rod position consistent with the assumptions used in the

existing accident and transient analyses. The new action statement

provides sufficient time for troubleshooting while avoiding

unnecessary plant shutdowns per TS 3.0.3.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change to TS 3.1.3.2 does not involve a significant

reduction in a margin of safety. As discussed in Section IV.A above,

the results of the FSAR Chapter 15 safety analyses for the

applicable events, are not affected by the proposed changes.

Therefore, the safety margins demonstrated by these analyses remain

unchanged. The additional time to obtain the flux maps is consistent

with the 12-hour time frame allowed to verify shutdown margin when a

rod is misaligned from its group step counter height by more than

plus or minus 12 steps in TS 3.1.3.1 and remains within a shiftly

basis. Therefore, it does not reduce the margin of safety.

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

review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Va

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