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 October 17, 1994, through October 28, 1994.

The last biweekly notice was published on October 26, 1994 (59 FR

53834).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

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

20555. The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By December 9, 1994, 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 20555 and at the local

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

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

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

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

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

and the Secretary or the designated Atomic Safety and Licensing Board

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

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555, Attention: Docketing and Services

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

the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the

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

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

the Commission by a toll-free telephone call to Western Union at 1-800-

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

should be given Datagram Identification Number N1023 and the following

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

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

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, and to the

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

Street, NW., Washington, DC 20555, and at the local public document

room for the particular facility involved.

Carolina Power & Light Company, et al.

Docket Nos. 50-325 and 50-324

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County,

North Carolina.

Date of amendments request: September 30, 1994.

Description of amendments request: The amendments would revise the

Technical Specifications to eliminate the scram and isolation trip

functions from the main steam line radiation monitor (MSLRM). This

change would specifically remove the reactor scram, main steam line

isolation valve closure, main steam line drain valve closure, reactor

water sample line isolation, and mechanical vacuum pump line isolation

actuated on a MSLRM High-High Radiation signal. The actuation signal

for isolation of the reactor water sample line will be replaced with a

low condenser vacuum signal. The isolation of the mechanical vacuum

pump line will be changed to a signal from the main stack radiation

monitor.

The MSLRMs will have both High Radiation and High-High Radiation

alarms. The setpoint for the MSLRM High Radiation alarm will be set at

or below 1.5 times the nominal full power background radiation adjusted

for Hydrogen water chemistry operation. The setpoint for the condenser

off-gas radiation monitor will be set at a value of 1.5 times

background radiation, but not less than 1.5 Rem per hour.

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 amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The deletion of the MSLRM trip function from the reactor scram and the

Group 1 isolation initiation logic removes a potential transient

initiation and therefore decreases the probability of plant transients

occurring due to inadvertent scrams resulting from this system.

The deletion of the MSLRM trip function from the Main Steam Drain

Valve, the Reactor Water Sample Isolation Valve, and the Mechanical

Vacuum Pump line isolation logic, does not affect the initiators of any

accident previously evaluated in the Safety Analysis Report. Therefore,

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

occurrence of any accident previously evaluated.

The NRC staff acceptance criterion for the Control Rod Drop

Accident is that the doses from the accident fall significantly below

the limits given in 10 CFR Part 100. The releases calculated for

accident during plant operations when the Steam Jet Air Ejectors (SJAE)

are operating and when the Mechanical Vacuum Pumps are operating are

within these acceptance limits.

In NEDO-31400, GE shows that the occurrence of a CRDA, with the MSL

high radiation isolation removed, and SJAE in operation, results in

offsite radiological exposures that are small fractions of 10CFR100

guidelines. Since the Brunswick specific CRDA doses are lower than the

[sic] calculated by GE and the GE dose parameters envelope those used

for the Brunswick analysis, it is concluded that the NRC's findings

that the radiological release consequence is within the staff's

acceptance criteria, even without the automatic MSIV trip, is

applicable to Brunswick.

While not specifically addressed in the GE evaluation, Carolina

Power and Light also proposes to eliminate the Main Steam Line Drain

valves, the Reactor Water Sample Line isolation valves, and the

mechanical vacuum line isolation valves from the MSLRM isolation logic.

Main Steam Line Drain Valves B21-F016 and B21-F019 drain to the main

condenser, which is the same flow path as the MSIVs. The discharge of

both the MSIV and MSL drain flow paths is processed through the offgas

system. Any radiation released through the drain valves during a

control rod drop accident will be negligible and, for Brunswick, is

bounded by the NEDO analysis.

The reactor water sample line provides a small amount of reactor

water to the Reactor Building Sample Panel. The discharge of the

Reactor Building Sample Panel is routed through the floor drain sump to

the liquid radwaste system. Any releases through this path would be

negligible and, for Brunswick, is bounded by the NEDO analysis.

The mechanical vacuum pumps are used only when the reactor is at

low power (less than 5%) and there is insufficient steam flow to

operate the Steam Jet Air Ejectors. The increase in radiation will be

detected by the MSLRMs and annunciated in the Main Control Room.

Operators will be instructed, in the annunciator response procedures,

to take action to stop the Mechanical Vacuum Pump(s) and isolate the

Mechanical Vacuum Pump line. The amount of radiation released prior to

isolating the line would represent the most limiting case for this

accident. However, it will still be well within 10 CFR Part 100 limits.

Additionally, the dose received in the Main Control Room as a result of

this accident is within General Design Criteria 19 (SRP 6.4) limits.

Therefore, since elimination of the MSIV [sic, MSLRM] scram and

isolation functions would not result in an increase in exposure above

NRC acceptance limits, the proposed changes will not significantly

increase the consequences of a previously evaluated accident.

2. The proposed amendments would not create the possibility of a

new or different kind of accident from any accident previously

evaluated. The function of a MSLRM trip is to detect abnormal fission

product release and isolate the steam lines, thereby stopping the

transport of fission products from the reactor to the main condenser.

The monitors do not perform a prevention function for any kind of

accident. The existence of a MSLRM trip does not prevent the occurrence

of a fuel failure event or any other type of event. The elimination of

these signals, which served only in a mitigative function, does not

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

those previously evaluated. Also, radiation monitors with alarm

functions will remain installed in the plant to warn the operators of a

high radiation condition in the main steam lines, or in the off-gas

system. Thus no new or different accident can be postulated by the

proposed changes.

3. The proposed amendments do not involve a significant reduction

in a margin of safety. As shown in the topical report, the changes

represent an overall improvement in plant safety. Safe operation of the

plant is further enhanced by elimination of the unnecessary scram and

isolation of the reactor vessel. With implementation of these changes,

1) the primary heat sink (main condenser) remains available, 2) large

transients on the reactor vessel, as well as challenges to the ESF, are

avoided, and 3) the Offgas system remains available to control the

pathway of potential releases. As such, the margin of safety is

enhanced by the proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Acting Project Director: Michael L. Boyle.

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: October 7, 1994.

Description of amendment request: The proposed amendment would

revise the introduction to TS Section 6.9.3.3 to require the approved

revision number for the referenced analytical methods be listed in the

Core Operating Limits Report. The methodology referenced in 6.9.3.3.b.f

(XN-NF-82-49(A)) will be updated to clarify that all supplements are

included. New methodologies ANF-89-151(A) and EMF-92-081(A) will be

added to TS Section 6.9.3.3.b.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The proposed changes will have no influence on the probability of an

accident previously evaluated. No changes will be made to any safety

related equipment, systems, or setpoints used in determining the

probability of an evaluated accident. The plant design basis will not

be altered. Therefore, there will be no significant increase in the

probability of an accident previously evaluated.

Consequences are dependent on the type of accident and the

mitigating response of safety related equipment. Furthermore, the

magnitude of consequences are calculated, directly or through

supporting calculations, by use of NRC approved methodologies. The

proposed license amendment will not alter the function of safety

related equipment designed to mitigate the consequences of an accident

previously evaluated or allow operation of the facility outside any

current limitations or restrictions. Also, this amendment will not

alter the requirement that evaluation of the consequences of an

accident previously evaluated by determined/supported with NRC reviewed

and approved methodologies. The change to TS Section 6.9.3.3.b's

introductory wording satisfies an administrative commitment and the

requirements it adds are administrative in nature. Accordingly the

proposed license amendment will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The addition of and update to NRC previously reviewed and approved

methodologies in TS Section 6.9.3.3.b will not result in any design or

function changes to any safety related equipment designed to prevent

and/or mitigate accidents, to any setpoints or systems, or to any

portion of the plant design basis. Operation of the facility will

remain within all required limitations and/or restrictions. The change

to TS Section 6.9.3.3.b's introductory wording satisfies an

administrative commitment and the requirements it adds are

administrative in nature. Therefore, the proposed amendment will not

create the possibility of a new kind of accident from any accident

previously evaluated.

The addition of and update to NRC previously reviewed and approved

methodologies in TS Section 6.9.3.3.b will not result in any design or

function changes to any safety related equipment designed to prevent

and or mitigate accidents, to any setpoints or systems, or to any

portion of the plant design basis. Operation of the facility will

remain within all required limitations and/or restrictions. The changes

to TS Section 6.9.3.3.b's introductory wording satisfies an

administrative commitment and the requirements it adds are

administrative in nature. Therefore, the proposed amendment will not

create the possibility of a different kind of accident from any

accident previously evaluated.

3. The proposed amendment does not involve a significant reduction

in the margin of safety. The proposed license amendment is defined as

administrative in nature. No current operational limits, restrictions,

or operating modes of the facility and its equipment, safety related or

otherwise, designed to preserve the margin of safety will be changed or

affected by the proposed amendment. There will be no changes to

setpoints or to the plant design basis. The methodology proposed for

addition to TS Section 6.9.3.3.b and the methodology that will be

updated has been previously reviewed and approved by the NRC. The

change to TS Section 6.9.3.3.b's introductory wording satisfies an

administrative commitment and the requirements it adds are

administrative in nature. Accordingly the proposed license amendment

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

Attorney for licensee: R.E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Project Director: William H. Bateman.

Entergy Operations, Inc., et al.

Docket No. 50-416

Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi.

Date of amendment request: October 12, 1994.

Description of amendment request: The proposed amendment requests

the closure and deletion of License Condition 2.C.(26) related to

turbine disk integrity.

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. No significant increase in the probability or consequences of an

accident previously evaluated results from this change.

The proposed change would close and delete License Condition

2.C.(26). The approved methodology currently used to evaluate the

probability of rotor failure and the inspection interval will not be

changed. The closure and deletion of the license condition is an

administrative change and will affect any accident previously

evaluated.

The bounding accident for the turbine-generator as analyzed in the

Grand Gulf Nuclear Station (GGNS) Updated Final Safety Analysis Report

(UFSAR) is the occurrence of an external missile resulting from the

failure of a low pressure (LP) turbine disc. The probability of this

incident occurring is less than 1 x 10-5 per year, which is the

NRC acceptable failure criterion for probability.

Any extension to the service interval in the future will be

evaluated in accordance with the current methodology. The original

acceptable levels of failure will be maintained. Therefore, no

significant increase in the probability or consequences of a previously

evaluated accident results from this change.

2. The change would not create the possibility of a new or

different kind of accident from any previously evaluated.

The proposed change does not involve a change to the control logic

or operating procedures for the turbine but rather transfers the

control of the LP turbine disc inspection interval from the Operating

License to administrative control. The current approved methodology

will continue to be used when determining future inspection intervals.

Therefore, this change does not create the possibility of a new or

different kind of accident from any previously evaluated.

3. The change would not involve a significant reduction in a margin

of safety.

Closing and deleting the current license condition for LP turbine

disc inspections and controlling the inspection interval

administratively has no adverse effects to the margin of safety. The

current approved methodology for failures will continue to be used and

any changes to future inspection intervals will be evaluated by the

methodology. This change does not affect any previous safety analysis

presented in the UFSAR and does not affect the criteria used to

establish safety limits, the basis for limiting safety system settings,

the basis for limiting conditions of operation, a change to the

technical specifications or a change in plant operations.

Therefore, this change does not involve a significant reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

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

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

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

NRC Project Director: William D. Beckner.

Florida Power and Light Company

Docket Nos. 50-250 and 50-251

Turkey Point Plant, Units 3 and 4, Dade County, Florida.

Date of amendment request: October 20, 1994.

Description of amendment request: The licensee proposes to change

Turkey Point, Units 3 and 4 Technical Specifications (TS) by revising

TS 1.9, Definitions--CORE ALTERATIONS to only address activities which

may, in actuality, affect core reactivity. In addition, the licensee

proposes to revise TS 3.9.4, Containment Building Penetrations to allow

both containment personnel airlock (PAL) doors to be open during core

alterations and movement of irradiated fuel in containment provided (a)

that at least one PAL door is capable of being closed; (b) the plant is

in Mode 6 with at least 23 feet of water above the fuel; and (c) a

designated individual is available outside the PAL to close the door.

The licensee also proposes a revision to the footnote of TS 3.9.4, to

remove the description of the purpose for imposing administrative

controls.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The change in the definition of CORE ALTERATIONS would allow the

movement of a temporary source range detector or other small

components, such as cameras, tools, etc., within the reactor vessel

without the activity being considered CORE ALTERATIONS. The potential

exists, however small, that an object can be dropped into the reactor

vessel. However, the justification for this change, is that the

insertion of small components into the reactor vessel will have no

effect on core reactivity since these items displace a small volume of

borated water, and sufficient borated water will surround the

components and provide the necessary neutron absorption to

neutronically isolate the components from the reactor. The consequences

of dropping one of these small components into the vessel are bounded

by the In-Containment Fuel Handling Accident Analysis discussed in

Chapter 14.2.1 of the Turkey Point Updated Final Safety Analysis Report

(UFSAR). Therefore, the proposed change is bounded by the current and

the proposed In-Containment Fuel Handling Accident Analyses and will

not involve a significant increase in the probability or consequences

of an accident previously evaluated.

The proposed change to TS 3.9.4 would allow the containment

personnel airlock (PAL) doors to be open during fuel movement and core

alterations. Currently, a single PAL door is closed during fuel

movement and core alterations to prevent the escape of radioactive

material in the event of a in-containment fuel handling accident. The

PAL is not an initiator of an accident. Whether the PAL doors are open

or closed during fuel movement and core alterations has no affect on

the probability of any accident previously evaluated.

Allowing the PAL doors to be open during fuel movement and core

alterations does not increase the consequences from a fuel handling

accident. The calculated offsite doses are well within the limits of 10

CFR Part 100. In addition, the calculated doses are larger than the

expected doses because the calculation does not incorporate the closing

of the PAL door after the containment is evacuated. The proposed change

should significantly reduce the dose to workers in containment in the

event of a fuel handling accident by reducing the time required to

evacuate the containment. The proposed change will also significantly

decrease the wear on the PAL doors and, consequently, increase the

availability of the PAL doors in the event of an accident.

The proposed change to the footnote of TS 3.9.4 is administrative

in nature, and does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The changes being proposed do not affect assumptions contained in

plant safety analyses or the physical design of the plant, nor do they

affect Technical Specifications that preserve safety analysis

assumptions. Therefore, operation of the facility in accordance with

the proposed amendments would not involve a significant increase in the

probability or consequences of an accident previously analyzed.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different kind

of accident from any accident previously evaluated.

The change in the definition of CORE ALTERATIONS would allow the

movement of a temporary source range detector or other small

components, such as cameras, tools, etc., within the reactor vessel

without the activity being considered CORE ALTERATIONS. The potential

exists however small, that an object can be dropped into the reactor

vessel. However, the justification for this change, is that the

insertion of small components into the reactor vessel will have no

effect on core reactivity since these items displace a small volume of

borated water, and sufficient borated water will surround the

components and provide the necessary neutron absorption to

neutronically isolate the components from the reactor. The consequences

of dropping one of these small components into the vessel are bounded

by the In-Containment Fuel Handling Accident Analysis discussed in

Chapter 14.2.1 of the Turkey Point UFSAR. Therefore the proposed change

is bounded by the current and the proposed In-Containment Fuel Handling

Accident Analyses and will not create the possibility of a new or

different kind of accident.

The proposed change to Specification 3.9.4 affects a previously

evaluated accident, i.e., in-containment fuel handling accident. Both

the current and the proposed In-Containment Fuel Handling Accident

Analysis assume that all of the iodines and noble gases that become

airborne within the containment escape and reach the site boundary and

low population zone with no credit taken for the containment building

barrier or for decay or deposition taken. Since the proposed change

does not involve the addition or modification of equipment nor does it

alter the design of plant systems and the revised analysis is

consistent with the current In-Containment Fuel Handling Accident

Analysis, the proposed change does not create the possibility of a new

or different kind of accident from any accident previously evaluated.

The proposed change to the footnote of TS 3.9.4 is administrative

in nature and does not create the possibility of a new or different

kind of accident from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The change in the definition of CORE ALTERATIONS would allow the

movement of a temporary source range detector or other small

components, such as cameras, tools, etc., within the reactor vessel

without the activity being considered CORE ALTERATIONS. The potential

exists however small, that an object can be dropped into the reactor

vessel. However, the justification for this change, is that the

insertion of small components into the reactor vessel will have no

effect on core reactivity since these items displace a small volume of

borated water, and sufficient borated water will surround the

components and provide the necessary neutron absorption to

neutronically isolate the components from the reactor. The consequences

of dropping one of these small components into the vessel are bounded

by the Fuel Handling Accident Analysis discussed in Chapter 14.2.1 of

the Turkey Point UFSAR. Therefore, the proposed change is bound by the

current In-Containment Fuel Handling Accident Analyses and as a result

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

The margin of safety as defined by 10 CFR Part 100 has not been

reduced. There is no increase in calculated offsite dose resulting from

a fuel handling accident in containment and the calculated dose is a

small fraction of the limits given in 10 CFR Part 100. The proposed

changes do not alter the bases for assurance that safety-related

activities are performed correctly or the basis for any Technical

Specification that is related to the establishment of or maintenance of

a safety margin. Therefore, operation of the facility in accordance

with the proposed amendments would not involve a significant reduction

in a margin of safety.

The proposed change to the footnote of TS 3.9.4 is administrative

in nature and does not relate to or modify the safety margins defined

in, and maintained by, 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: Florida International

University, University Park, Miami, Florida 33199.

Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,

P.C., 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: Mohan C. Thadani, (Acting)

Florida Power and Light Company

Docket Nos. 50-250 and 50-251

Turkey Point Plant, Units 3 and 4, Dade County, Florida.

Date of amendment request: October 20, 1994.

Description of amendment request: This supersedes the licensee's

original request dated July 19, 1994, and noticed in the Federal

Register on August 3, 1994 (59 FR 39588). The licensee proposes to

change Turkey Point, Units 3 and 4 Technical Specifications (TS)

4.8.1.1.2e. and 4.8.1.1.2f., which address Emergency Diesel Generator

(EDG) fuel oil testing, by replacing the specific EDG fuel oil

Surveillance Requirements with the requirement to verify new and stored

EDG fuel oil in accordance with the Diesel Fuel Oil Testing Program. In

addition, the licensee proposes the addition of ACTION statements g.

and h., to TS 3.8.1.1, to address the required action in the event the

diesel fuel oil does not meet the Diesel Fuel Oil Testing Program

limits. The Diesel Fuel Oil Testing Program will be described in both

TS 6.8.4 and the BASES Section to the Technical Specifications. In

addition, FPL proposes revising TS 6.8.1 to include the requirement

that written procedures shall be established, implemented and

maintained for implementation of the Diesel Fuel Oil Testing Program.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed changes to the Technical Specifications will permit

the Technical Specification required testing of Emergency Diesel

Generator (EDG) fuel oil in accordance with the Turkey Point, Units 3

and 4 Diesel Fuel Oil Testing Program. The proposed change will permit

FPL to use more recent editions of the American Society for Testing and

Materials (ASTM) standards currently listed in Technical Specification

Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f. Prior to changing

the Diesel Fuel Oil Testing Program, the proposed change will be

evaluated pursuant to Title 10 Code of Federal Regulations Sec. 50.59

(10 CFR Sec. 50.59), ``Changes, tests, and experiments.'' Title 10 CFR

Sec. 50.59 permits a licensee to make changes in the procedures as

described in the safety analysis report without prior Commission

approval, provided that the proposed changes does not involve an

unreviewed safety question.

Title 10 CFR Sec. 50.59(a)(2) states that a proposed change

involves an unreviewed safety question (i) if the probability of

occurrence or the consequences of an accident or malfunction of

equipment important to safety previously evaluated in the safety

analysis report may be increased. Consequently, since any change to the

Diesel Fuel Oil Testing Program, including the ASTM standard or ASTM

edition standard to be used to evaluate EDG fuel oil acceptability, the

change must be evaluated relative to the more restrictive evaluation

criterion of 10 CFR Sec. 50.59, then operation of the facility in

accordance with the proposed amendments would not involve a significant

increase in the probability or consequences of an accident previously

evaluated. The EDG fuel oil TS Surveillance Requirements will be

replaced with a requirement to test the EDG fuel oil in accordance with

the Turkey Point Units 3 and 4 Diesel Fuel Oil Testing Program.

ACTION statement g. of TS 3.8.1.1 is added to address the required

action in the event the new fuel oil properties do not meet the Diesel

Fuel Oil Testing Program limits. A failure to meet the American

Petroleum Institute (API) gravity, kinematic viscosity, flash point or

clarity limits is cause for rejecting the new fuel oil prior to the

addition to the Diesel Fuel Oil Storage Tanks, but does not represent a

failure to meet the Limiting Condition for Operation (LCO) of TS

3.8.1.1, since the new fuel oil has not been added to the storage

tanks. Provided these new fuel oil properties are met subsequent to the

addition of the new fuel oil to the storage tanks, 30 days is provided

to complete the analyses of the other fuel oil properties specified in

Table 1 of ASTM-D975-81, except sulfur which may be performed in

accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the other

new fuel oil properties specified in Table 1 of ASTM-D975-81 are not

met, ACTION statement g. of TS 3.8.1.1 provides an additional 30 days

to meet the Diesel Fuel Oil Testing Program limits. This additional 30

day period is acceptable because the fuel oil properties of interest,

even if they are not within limits, would not have an immediate effect

on EDG operation.

ACTION statement h. of TS 3.8.1.1 is added to address the required

action in the event the stored fuel oil total particulates do not meet

the Diesel Fuel Oil Testing Program limits. Fuel oil degradation during

long term storage shows up as an increase in particulate, due mostly to

oxidation. The presence of particulate does not mean the fuel oil will

not burn properly in a diesel engine. The frequency for performing

surveillance on stored fuel oil is based on stored fuel oil degradation

trends which indicate that particulate concentration is unlikely to

change significantly between surveillances.

Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil

Testing Program, FPL will need to determine if the proposed program

change is at least as, if not more, effective, in detecting

unsatisfactory fuel oil. The EDGs will thus continue to function as

designed and the probability or consequences of previously evaluated

accidents will be unaffected.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different kind

of accident from any accident previously evaluated.

The proposed changes to the Technical Specifications will permit

the Technical Specification required testing of Emergency Diesel

Generator fuel oil using more recent editions of the American Society

for Testing and Materials (ASTM) standards currently listed in

Technical Specification Surveillance Requirements 4.8.1.1.2e. and

4.8.1.1.2f. Prior to changing the edition of the previously approved

ASTM standard being used to evaluate the EDG fuel oil, the proposed

edition standard will be evaluated pursuant to 10 CFR Sec. 50.59,

``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits a

licensee to make changes in the procedures as described in the safety

analysis report without prior Commission approval, provided that the

proposed changes does not involve an unreviewed safety question. Title

10 CFR Sec. 50.59(a)(2) states that a proposed change involves an

unreviewed safety question (ii) if a possibility for an accident or

malfunction of a different type than any evaluated previously in the

safety analysis report may be created. Consequently, since any change

to the edition of the ASTM standard to be used to evaluate EDG fuel oil

acceptability must be evaluated relative to the more restrictive

evaluation criterion of 10 CFR Sec. 50.59, then operation of the

facility in accordance with the proposed amendments would not create

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

accident previously evaluated.

ACTION statement g. of TS 3.8.1.1 is added to address the required

action in the event the new fuel oil properties do not meet the Diesel

Fuel Oil Testing Program limits. A failure to meet the API gravity,

kinematic viscosity, flash point or clarity limits is cause for

rejecting the new fuel oil prior to the addition to the Diesel Fuel Oil

Storage Tanks, but does not represent a failure to meet the Limiting

Condition for Operation (LCO) of TS 3.8.1.1, since the new fuel oil has

not been added to the storage tanks. Provided these new fuel oil

properties are met subsequent to the addition of the new fuel oil to

the storage tanks, 30 days is provided to complete the analyses of the

other fuel oil properties specified in Table 1 of ASTM-D975-81, except

sulfur which may be performed in accordance with ASTM-D1552-79 or ASTM-

D2622-82. In the event the other new fuel oil properties specified in

Table 1 of ASTM-D975-81 are not met, ACTION statement g. of TS 3.8.1.1

provides an additional 30 days to meet the Diesel Fuel Oil Testing

Program limits. This additional 30 day period is acceptable because the

fuel oil properties of interest, even if they are not within limits,

would not have an immediate effect on EDG operation.

ACTION statement h. of TS 3.8.1.1 is added to address the required

action in the event the stored fuel oil total particulates does not

meet the Diesel Fuel Oil Testing Program limits. Fuel oil degradation

during long term storage shows up as an increase in particulate, due

mostly to oxidation. The presence of particulate does not mean the fuel

oil will not burn properly in a diesel engine. The frequency for

performing surveillance on stored fuel oil is based on stored fuel oil

degradation trends which indicate that particulate concentration is

unlikely to change significantly between surveillances.

Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil

Testing Program, FPL will need to determine if the proposed program

change is at least as, if not more, effective, in detecting

unsatisfactory fuel oil. Since the proposed changes do not involve a

change in the design of any plant system or component, and since the

proposed changes will need to evaluate the effect of any ASTM standard

edition change on the level of EDG reliability, the change proposed

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

from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes to the Technical Specifications will permit

the Technical Specification required testing of Emergency Diesel

Generator (EDG) fuel oil using more recent editions of the American

Society for Testing and Materials (ASTM) standards currently listed in

Technical Specification Surveillance Requirements 4.8.1.1.2e. and

4.8.1.1.2f. Prior to changing the edition of the previously approved

ASTM standard being used to evaluate the EDG fuel oil, the proposed

edition standard will be evaluated pursuant to 10 CFR Sec. 50.59,

``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits a

licensee to make changes in the procedures as described in the safety

analysis report without prior NRC approval, provided that the proposed

changes does not involve an unreviewed safety question. Title 10 CFR

Sec. 50.59(a)(2) states that a proposed change involves an unreviewed

safety question (iii) if the margin of safety as defined in the basis

for any technical specification is reduced. Consequently, since any

change to the edition of the ASTM standard to be used to evaluate EDG

fuel oil acceptability must be evaluated relative to the more

restrictive evaluation criterion of 10 CFR Sec. 50.59, then operation

of the facility in accordance with the proposed amendments would not

involve a significant reduction in a margin of safety.

ACTION statement g. of TS 3.8.1.1 is added to address the required

action in the event the new fuel oil properties do not meet the Diesel

Fuel Oil Testing Program limits. A failure to meet the API gravity,

kinematic viscosity, flash point or clarity limits is cause for

rejecting the new fuel oil prior to the addition to the Diesel Fuel Oil

Storage Tanks, but does not represent a failure to meet the Limiting

Condition for Operation (LCO) of TS 3.8.1.1, since the new fuel oil has

not been added to the storage tanks. Provided these new fuel oil

properties are met subsequent to the addition of the new fuel oil to

the storage tanks, 30 days is provided to complete the analyses of the

other fuel oil properties specified in Table 1 of ASTM-D975-81, except

sulfur which may be performed in accordance with ASTM-D1552-79 or ASTM-

D2622-82. In the event the other new fuel oil properties specified in

Table 1 of ASTM-D975-81 are not met, ACTION statement g. of TS 3.8.1.1

provides an additional 30 days to meet the Diesel Fuel Oil Testing

Program limits. This additional 30 day period is acceptable because the

fuel oil properties of interest, even if they are not within limits,

would not have an immediate effect on EDG operation.

ACTION statement h. of TS 3.8.1.1 is added to address the required

action in the event the stored fuel oil total particulates does not

meet the Diesel Fuel Oil Testing Program limits. Fuel oil degradation

during long term storage shows up as an increase in particulate, due

mostly to oxidation. The presence of particulate does not mean the fuel

oil will not burn properly in a diesel engine. The frequency for

performing surveillance on stored fuel oil is based on stored fuel oil

degradation trends which indicate that particulate concentration is

unlikely to change significantly between surveillances.

Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil

Testing Program, FPL will need to determine if the proposed program

change is at least as, if not more, effective, in detecting

unsatisfactory fuel oil. Since the proposed changes will require a

safety evaluation to assure that the reliability of the EDGs using fuel

oil tested in accordance with the different ASTM standard edition

maintains the current margin of safety, the proposed changes do not

involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,

P.C., 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: Mohan C. Thadani, Acting.

Florida Power Corporation, et al.

Docket No. 50-302

Crystal River Nuclear Generating Plant, Unit No. 3, Citrus County,

Florida.

Date of amendment request: September 30, 1994.

Description of amendment request: The proposed amendment would

revise the Crystal River 3 (CR3) Nuclear generating Plant Technical

Specifications (TS) to allow an increase in the rated thermal power

(RTP) for CR-3 from the current 2544 level to 2568 Megawatt thermal

(Wt). Accordingly, in TS 1.1, ``Definitions,'' would be revised to

indicate the new power level of 2568 MWt. The proposed change would not

require any hardware modifications.

Basis for proposed no significant hazards consideration

determination: Currently, CR-3 is operating at a maximum RTP of 2544

MWt. The licensee proposes to operate at a maximum RTP of 2568 MWt, an

increase of 24 MWt over the current licensed power of 2544 MWt.

The licensee states that the Babcock and Wilcox (B&W) 177 Fuel

Assembly (FA) Nuclear Steam Supply System (NSSS) in the CR3 design is

capable of operating at a thermal power level of 2772 MWt. Due to

limitations in the secondary area of the plant, the licensee requests

authorization to operate at 2568 MWt which is less than the design

level of 2772 MWt. The licensee performed a detailed engineering study

on this power increase.

As required by 10 CFR 50.91(a), the licensee has provided its

analysis of the issue of no significant hazards consideration, which is

presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the probability

of occurrence or consequences of an accident previously evaluated. The

thermal-hydraulic and nuclear characteristics of the reactor core were

originally designed for a rated thermal power of 2568 MWt or higher.

Therefore, the proposed thermal power increase to the reference power

level of 2568 MWt does not change the original design assumptions and

analyses for the reactor core. Most of the design basis accidents and

transients were originally evaluated at the proposed power level. As

described more fully in this submittal, those transients and accidents

that were not originally evaluated at 2568 MWt were re-evaluated using

CR-3 FSAR [Final Safety Analysis Report] Chapter 14 accident sequence

of events, reactor protection criteria, and approved calculational

methods. Based on this evaluation and initial plant design evaluations,

FPC [Florida Power Corporation, the licensee for CR3] has determined

that the probability and consequences of design basis transients and

accidents are not significantly increased and that the radiological

consequences from the design basis transients and accidents remain well

below 10 CFR 100 limits.

FPC has also reviewed CR-3 balance of plant and safety related

systems to determine which systems and components could be affected by

the proposed power increase. The changes to the reactor coolant system

and secondary conditions and parameters are discussed in this

submittal. These changes are minor in nature. The only Technical

Specification change is to revise the reference power to 2568 MWt. No

facility modifications will be required. FPC evaluated the systems and

components and concluded that these systems and components will

continue to perform within their design parameters with the unit

operating at 2568 MWt.

Based on the foregoing, the proposed amendment does not

significantly increase the probability or consequences of an accident

previously evaluated.

2. The proposed thermal power increase does not create the

possibility of a new or different kind of accident from previously

evaluated accidents. As noted above, the thermal-hydraulic and nuclear

characteristics of the reactor core were originally designed for

operation at the proposed thermal power. Therefore, operation at the

proposed power level does not introduce new or different performance

characteristics that create the possibility of a new or different kind

of accident.

FPC has also reviewed CR-3 safety-related systems and balance of

plant systems to determine which systems could be affected by the

proposed power increase and the resultant minor changes in plant

parameters and operating conditions. Systems that could be affected

were evaluated using the licensing basis criteria described in the CR-3

FSAR to assure their adequacy at the increased power level. Included in

these evaluations were plant features that are not power level related

or directly affected by an increase in power level, as well as,

associated issues such as environmental considerations. Equipment

performance and plant operation were evaluated with respect to actual

performance versus projected operating conditions to identify any

hardware modifications required to achieve the upgraded power. Based on

this evaluation, FPC has determined that all systems will continue to

perform within their design parameters at 2568 MWt and that no physical

modifications to these systems will be necessary to accommodate a 2568

MWt rating. Only minor re-calibration of plant instrumentation to

reflect the increased power will be needed. The proposed power level

does not introduce any new performance characteristics or modes of

operation for plant systems and components, and does not introduce any

new failure modes.

Based on the foregoing, the proposed amendment does not create the

possibility of a new or different kind of accident.

3. The proposed amendment does not involve a significant reduction

in a margin of safety. The thermal-hydraulic and nuclear

characteristics of the reactor core were originally designed for

operation at the proposed power level. Most of the design basis

transients and accidents were originally analyzed assuming a power

level of 2568 MWt or higher. As described more fully in this submittal,

those transients and accidents that were not originally analyzed at

2568 MWt were re-evaluated using CR-3 FSAR Chapter 14 accident sequence

of events, reactor protection criteria, and approved calculational

methods. FPC has determined that operation with the proposed thermal

power will be bounded by the original analyses. In addition, FPC's

evaluation of affected plant systems and components revealed that plant

systems and components will continue to operate within their design

parameters with no significant change in a margin of safety.

Based on the foregoing, the proposed amendment 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 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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

Attorney for licensee: A. H. Stephens, General Counsel, Florida

Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida

33733.

NRC Project Director: Mohan C. Thadani, (Acting).

Indiana Michigan Power Company

Docket Nos. 50-315 and 50-316

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan.

Date of amendment request: August 3, 1994.

Description of amendment requests: The proposed amendments would

allow the radiological effluent technical specifications (TS) to be

relocated to other controlled documents. Procedural details contained

in the current radiological effluents TS have been relocated to either

the Offsite Dose Calculation Manual (OCDM) or the Process Control

Program (PCP), as applicable. Proposed revisions to the OCDM and PCP

have been prepared in accordance with the proposed changes to the

administrative controls section of the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The changes described above in no way negatively impact the

requirements of the T/Ss. Separating the turbine room sump releases

from the others is purely a clarification of the method we handle

releases. The six ground monitoring wells added to the T/S table

updates our current monitoring practice. With the six extra wells to

monitor, we exceed the monitoring requirements of the T/Ss. Therefore,

it is concluded that the proposed changes do not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

Criterion 2

No changes to the LCOs for either T/S are proposed as part of this

amendment request. The proposed change does not involve any physical

changes to the plant or any changes to plant operations. The changes

merely propose to update our methods of implementing the T/S with our

current practices. Thus, the proposed change does not create the

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

previously evaluated.

Criterion 3

The changes described above in no way negatively impact the

requirements of the T/Ss. Separating the turbine room sump releases

from the others is purely a clarification of the method we handle

releases. The six ground monitoring wells added to the T/S table

updates our current monitoring practice. With the extra wells to

monitor, we exceed the monitoring requirements called for in the T/Ss.

Therefore, it is concluded that the proposed changes do 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 requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

and Trowbridge, 2300 N Street, NW, Washington, DC 20037.

NRC Project Director: John N. Hannon.

Niagara Mohawk Power Corporation

Docket No. 50-410

Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.

Date of amendment request: October 5, 1994.

Description of amendment request: The proposed license amendment

would revise the applicability requirements of Technical Specification

(TS) 3.7.3 to require operability of the Control Room Outdoor Air

Special Filter Train System in Operational Conditions 1, 2, 3 and **

(when irradiated fuel is being handled in the reactor building and

during CORE ALTERATIONS and operations with a potential for draining

the reactor vessel and uncovering irradiated fuel) rather than in all

Operational Conditions and * * *. The applicability requirements for

Action Statement b of TS 3.7.3 and for the Radiation Monitoring

Instrumentation required operable by TS Tables 3.3.7.1-1 and 4.3.7.1-1

would be changed in a similar manner. The proposed amendment would also

add a notation to Action Statement b.1 of TS 3.7.3 stating that the

provisions of Specification 3.0.4 are not applicable provided an

operable control room filter train is in the emergency pressurization

mode of operation. The licensee stated that these proposed changes are

consistent with the requirements of the NRC's Improved Standard

Technical Specifications (NUREG-1433) and with Generic Letter 87-09,

``Section 3.0 and 4.0 of the Standard Technical Specifications (STS) on

the Applicability of Limiting Conditions for Operation and Surveillance

Requirements.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The Control Room Outdoor Air Special Filter Train System is not an

initiator or precursor to an accident. The Control Room Outdoor Air

Special Filter Train System responds to a release of radioactivity to

the outside environment as detected in the air supply to the control

room by providing a radiologically controlled environment within the

control room. In operational conditions 4 and 5, the probability and

consequences of a design basis accident are reduced due to the pressure

and temperature limitations in these operational conditions. Therefore,

maintaining the chiller subsystem operable is not required in

operational conditions 4 and 5, except for the * * * operational

condition. Therefore, a change to applicability and action statements

of LCO [Limiting Condition For Operation] 3.7.3 cannot affect the

probability of a previously evaluated accident.

All accidents which take credit for operation of the Control Room

Outdoor Air Special Filter Train System in the emergency pressurization

mode of operation are analyzed and presented in Chapter 15 of the USAR

[Updated Safety Analysis Report]. These accidents can only occur in

operational conditions 1, 2, 3 and * * *.

Accordingly, the proposed change in the applicability of LCO 3.7.3

from all operational conditions (i.e., 1, 2, 3, 4, 5 and * * *) to

operational conditions 1, 2, 3 and * * * does not significantly

increase the consequences of an accident previously evaluated. The

proposed change to action statement b of LCO 3.7.3 and to Tables

3.3.7.1-1 and 4.3.7.1-1 of LCO 3.3.7.1 is consistent with the above

change.

Sections 15.7.4 and 15.7.5 of the USAR evaluate a fuel handling

accident and a spent fuel cask drop accident, respectively. The

radiological evaluation of these accidents considers the unfiltered

radioactivity that enters the control room prior to the automatic

operation of the Control Room Outdoor Special Filter Train System in

the emergency pressurization mode of operation. The radiological

consequences of these accidents are within the limits of GDC [General

Design Criterion]-19.

With one control room filter train inoperable and prior to entering

the operational condition, the proposed change to action statement b.1

of LCO 3.7.3 would require an operable control room filter train be

placed in the emergency pressurization mode of operation. During an

accident involving the release of radioactivity to the environment, an

operable control room filter train would already be running in the

emergency pressurization mode and performing its safety function,

thereby preventing the entry of unfiltered radioactivity into the

control room. Therefore, if a fuel handling accident or a spent fuel

cask drop accident were to occur and release radioactivity, the control

room personnel radiological doses would be less than the doses depicted

in the USAR. Accordingly, the Technical Specification change to action

statement b.1 does not significantly increase the consequences of a

previously evaluated accident.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

This amendment does not involve any accident precursors or

initiators. In addition, this amendment does not require any changes to

plant equipment.

During an accident involving the release of radioactivity to the

environment an operable control room filter train would already be

running in the emergency pressurization mode and performing its safety

function. Furthermore, the operating status of a running control room

filter train would be unaffected by the receipt of an automatic start

signal due to high radiation in either air intake to the Control Room

Outdoor Air Special Filter Train System. Therefore, the proposed

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

accident from any accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed change in the applicability of LCO 3.7.3 from all

operational conditions (i.e., 1, 2, 3, 4, 5 and * * *) to operational

conditions 1, 2, 3 and * * * is consistent with the safety analysis

contained in the USAR. The proposed changes to action statement b of

LCO 3.7.3 and to Tables 3.3.7.1-1 and 4.3.7.1-1 of LCO 3.3.7.1 is

consistent with the above change.

Entry into the ** operational condition for LCO 3.7.3 with one

control room filter train inoperable and the other control room filter

train operable and operating in the emergency pressurization mode

provides a comparable level of safety to two operable non-running

control room filter trains. The remedial measure prescribed by

Technical Specification action statement b.1 (placing an operable

control room filter train in the emergency pressurization mode of

operation) for which the exception to LCO 3.0.4 is proposed provides a

sufficient level of protection to permit operational mode changes and

safe long-term operation of NMP2 [Nine Mile Point Unit 2] consistent

with the licensing basis described in the USAR. Therefore, the proposed

change to action statement b.1 is consistent with Generic Letter 87-09,

``Sections 3.0 and 4.0 of the Standard Technical Specifications (STS)

on the Applicability of Limiting Conditions for Operation and

Surveillance Requirements.'' Accordingly, this change will not

significantly reduce the margin of safety.

This proposed amendment is consistent with the Improved Standard

Technical Specifications, NUREG-1433. Accordingly, as determined by the

analysis above, this proposed amendment involves no significant hazards

consideration.

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: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: Ledyard B. Marsh.

Niagara Mohawk Power Corporation

Docket No. 50-410

Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.

Date of amendment request: October 21, 1994.

Description of amendment request: The proposed amendment would add

a footnote to Technical Specification (TS) 4.8.1.1.2.e.8 which would

permit performance of the 24-hour functional test of the emergency

diesel generators (EDGs) during power operation. TS 4.8.1.1.2.e.8

currently requires the 24-hour functional test of the EDGs be performed

at least once per 18 months during shutdown; the proposed amendment

would permit this testing to be performed during power operation

provided the other two EDGs are operable. If either of the other two

EDGs become inoperable, the test would be aborted.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change to permit the 24 hour functional test of the

diesels to be performed during power operation does not increase the

chances for a previously analyzed accident to occur. The function of

the diesels is to supply emergency power in the event of a loss of

offsite power. Operation of the diesels is not a precursor to any

accident. Furthermore, the diesel generator being tested will remain

operable and will be available to supply emergency loads within the

required time. In addition, the two remaining diesel generators will be

operable during the test. Consequently, if an offsite disturbance were

to occur that affected the operability of the diesel being tested, the

two remaining diesels would be capable of feeding the loads necessary

for safe shutdown of the plant. This addresses the concerns raised in

Information Notice 84-69 regarding the operation of emergency diesel

generators connected in parallel with offsite power. In summary, the

proposed changes do not adversely affect the performance or the ability

of the diesel generators to perform their intended function.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any previously evaluated.

The proposed amendment to the 24 hour functional surveillance test

will not affect the operation of any safety system or alter its

response to any previously analyzed accident. The diesel will

automatically transfer from the test mode if necessary to supply

emergency loads in the requried time. The test mode is used for the

monthly surveillance of the diesel generators as well, therefore, no

new plant operating modes are introduced. In the event the diesel fails

the functional test it will be declared inoperable and the actions

required for an inoperable diesel will be performed. The remaining two

diesel generators will be operable and are capable of feeding the loads

necessary for safe shutdown of the plant.

Therefore, the proposed change will not create the possibility of a

new or different kind of accident from any previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed amendment will not reduce availability of the diesel

generator being tested to provide emergency power in the event of a

loss of offsite power. If a loss of offsite power or a loss of coolant

accident occurs during the surveillance test, the emergency bus would

de-energize and shed load. The diesel generator would then transfer

from the test mode to the emergency mode. It would then be available to

automatically supply emergency loads. In addition, the two remaining

generators will be operable during the test. Consequently, if an

offsite disturbance were to occur that affected the operability of the

diesel begin tested, the two remaining diesels would be capable of

feeding the loads necessary for safe shutdown of the plant. The time

required for the diesel being tested to pick up emergency loads will

not be affected by performing the 24 hour functional test during power

operation.

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 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: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: Ledyard B. Marsh.

Northeast Nuclear Energy Company et al.

Docket No. 50-336.

Millstone Nuclear Power Station, Unit No. 2, New London County,

Connecticut.

Date of amendment request: October 18, 1994.

Description of amendment request: The proposed amendment would

require three type A overall Integrated Containment Leakage Tests be

conducted at approximately equal intervals during shutdowns during each

10-year service period. For the third Type A test for the second 10-

year period, it would be conducted during the thirteenth refueling

outage extending the second 10-year service period to the end of the

thirteenth refueling outage. The amendment would also change the

Containment Leakage Bases by reflecting the conditions of a proposed

exemption to 10 CFR 50, Appendix J, that would remove the requirement

that the third Type A test for each 10-year period be conducted when

the plant is shutdown for the 10-year plant inservice inspection.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

* * * The proposed changes do not involve a SHC [significant

hazards consideration] because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

Type A tests are performed to ensure that the total leakage from

containment does not exceed the maximum allowable primary containment

leakage rate at the design pressure. This ensures compliance with the

dose limits of 10 CFR 100.

The proposal to revise Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 2 Technical Specifications will increase the

flexibility for scheduling the Type A tests. It does not modify the

maximum allowable leakage rate at the design containment pressure, does

not impact the design basis of the containment, and does not make any

physical or operational changes to existing plant structures, systems,

or components.

The first two Type A tests of the second 10-year service period for

Millstone Unit No. 2 have been conducted. The results of these tests

demonstrate that Millstone Unit No. 2 has maintained control of

containment integrity by maintaining margin between the acceptance

criterion and the ``As-Found'' and ``As-Left'' leakage rates.

Historically, Type A tests have a relatively low failure rate where

Type B and C testing (local leakage rate tests) could not detect the

leakage path. Most Type A test failures are attributed to failures to

Type B or C components (containment penetrations and isolation valves).

Type B and C components are tested per Surveillance Requirement

4.6.1.2.d for the Millstone Unit No. 2 Technical Specifications. These

tests are required to be conducted at intervals no greater than 24

months, and the acceptance criterion for the combined leakage rate for

all penetrations and valves subject to the Type B and C tests is 0.6

La. These local leakage rate tests provide assurance that

containment integrity is maintained. The relatively low ``As-Left''

Type B and C total leakage resulting from the past outage indicates

that the leakage has been maintained within the technical specification

acceptance criterion. The Type B and C tests will continue to be

performed in accordance with the requirements of Surveillance

Requirement 4.6.1.2.d. However, on September 26, 1994, NNECO submitted

a request for a one-time technical specification change, request for

enforcement discretion, and a request for a scheduler exemption from

Appendix J to 10 CFR 50 regarding the Schedule for Type B and C

testing. The NRC verbally granted enforcement discretion on September

24, 1994, and written enforcement discretion on September 30, 1994. The

schedular exemption request was granted on October 12, 1994.

The previous Type A, B, and C tests demonstrate that Millstone Unit

No. 2 has maintained control of containment integrity by maintaining a

conservative margin between the acceptance criterion and the ``As-

Found'' and ``As-Left'' leakage results. Based on this, the Millstone

Unit No. 2 containment is considered to be in sound condition. No

operations are known to have occurred which would suggest any

substantial degradation of these results.

Based on the above, the proposal to revise Surveillance Requirement

4.6.1.2.a of the Millstone Unit No. 2 Technical Specifications does not

involve a significant increase in the probability or consequences of an

accident previously analyzed.

2. Create the possibility of a new or different kind of accident

from any previously analyzed.

The proposal to revise Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 2 Technical Specifications will increase the

flexibility in scheduling the Type A tests. It does not make any

physical or operational changes to existing plant structures, systems,

or components. In addition, the proposal does not modify the acceptance

criterion for the Type A tests. Maintaining the leakage through the

containment boundary to the atmosphere within a specific value ensures

that the plant complies with the requirements of 10 CFR 100. The

containment boundary serves as an accident mitigator; it is not an

accident initiator. Therefore, the proposal to revise Surveillance

Requirement 4.6.1.2.a does not create the possibility of a new or

different kind of accident from any previously analyzed.

3. Involve a significant reduction in the margin of safety.

The proposal to revise Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 2 Technical Specifications will increase the

flexibility for scheduling the Type A tests. It does not modify the

maximum allowable leakage rate at the design containment pressure, does

not impact the design basis of the containment, and does not make any

physical or operational changes to existing plant structures, systems,

or components.

The first two Type A tests of the second 10-year service period for

Millstone Unit No. 2 have been conducted. The results of these tests

demonstrate that Millstone Unit No. 2 has maintained control of

containment integrity by maintaining margin between the acceptance

criterion and the ``As-Found'' and ``As-Left'' leakage rates.

Additionally, the results of the last Type B and C tests had

significant margin with respect to the acceptance criterion. Based on

the previous Type A, B, and C tests, the Millstone Unit No. 2

containment is considered to be in sound condition. No operations are

known to have occurred which would suggest any substantial degradation

of these results.

Based on the above, the proposal does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al.

Docket No. 50-423

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut.

Date of amendment request: September 30, 1994.

Description of amendment request: The licensee has proposed to

revise the Technical Specifications (1) to clarify the definition of

core alterations, (2) to change the verbiage in the Limiting Condition

For Operation (LCO) addressing the refueling operations, (3) to make

changes to three surveillance requirements involving source range

instrumentation, and (4) to change the LCO regarding the Residual heat

Removal and coolant circulation water levels to be consistent with the

guidance provided in NUREG-1431.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Boron Dilution in Mode 6--A boron dilution in Mode 6 is precluded

by technical specification requirements to close and lock all dilution

source valves. There is a provision for dilution valves to be opened

under administrative controls; in this case, cautionary measures will

be taken to control and monitor the reactivity addition. Deletion of

the source range analog operational test prior to core alterations will

not impact an accident previously evaluated since the sources range

monitors are verified operable prior to entry into Mode 6 and every 7

days thereafter. The change in definition for a core alteration means

that components which do not effect reactivity may be moved within the

reactor vessel without any additional condition such as direct

supervision of an SRO.

Since a boron dilution would not be initiated by movement of

nonfuel components within the reactor vessel, it is not impacted by the

change in definition of a core alteration.

Inadvertent Loading of a Fuel Assembly--Movement of a fuel assembly

would be performed as a core alteration under the supervision of an

SRO, therefore, it would not be impacted by the change to the

definition of a core alteration. The change to the source range

monitors also will not affect the probability of occurrence of a

misloaded fuel assembly since this accident is precluded by

administrative controls, as well as the source range monitors. Also,

there will be no degradation in the reliability or accuracy of the

source range monitors due to this change. The deletion of the

requirement to perform the analog channel operational test within eight

hours prior to core alterations will not impact performance of the

monitors, since they have to be checked prior to entry into Mode 6 and

every 7 days thereafter.

Fuel Handling Accident--Movement of fuel will not affect this

accident, because it will still be considered a core alteration.

Therefore, there is no effect on the probability of a fuel handling

accident. The source range monitors are not involved in the occurrence

of a fuel handling accident. The fuel handling accident is the only

accident considered here with radiological consequences. It will not be

impacted by the proposed changes.

Loss of RHR in Mode 6--The probability of this accident will not be

changed since the new requirement is the same as before. As before, RHR

may be secured for up to one hour per eight-hour period and boron

dilution operations may not be performed with RHR secured (although

this requirement is being added to the notes, the requirement is also

given elsewhere in the technical specifications). Additionally, the

existing reactor coolant system (RSC) temperature limits must still be

met.

Based on the above, the proposed changes do not involve a

significant increase in the probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or different kind of accident

from any previously analyzed.

All required systems will continue to operate as before. Therefore,

there is no possibility of a new or different kind of accident. The

deletion of the source range analog channel operational test within

eight hours prior to core alterations will not affect the performance

of the monitors since they will have had this test completed prior to

entry into Mode 6 and every 7 days thereafter. The change in definition

of a core alteration cannot create the possibility of a new type of

accident because those initiating events for accidents will remain

classified as core alterations.

3. Involve a significant reduction in the margin of safety.

The margin of safety for the above listed accidents will remain as

before.

a. Boron dilution in Mode 6--This accident calculates the time from

receipt of a shutdown margin monitor dilution alarm until the core

reaches criticality. Since this time is not changed, there is no

reduction in the margin of safety. In this case, the dilution is

precluded by administrative controls which will not be impacted by the

proposed changes.

b. Inadvertent Loading of a Fuel Assembly--Technical Specification

3.9.1.1 protects against this accident by requiring sufficient boron in

the RCS to prevent criticality for any core configuration including two

stuck RCCAs [rod cluster control assemblies] in the fully withdrawn

position. Since this requirement will not change, the margin of safety

will not change.

c. Fuel Handling Accident--The margin of safety for the

radiological limits is not changed.

d. Loss of RHR--Changes are editorial due to the revised definition

of a core alteration. There is no change to the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L.M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

Northern States Power Company

Docket Nos. 50-282 and 50-306

Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue

County, Minnesota.

Date of amendment requests: October 3, 1994.

Description of amendment requests: The proposed amendment would

revise Prairie island Nuclear Generating Plant Technical Specification

4.6, ``Periodic Testing of Emergency Power Systems.'' Specifically, the

proposed amendment would modify the emergency diesel generator (EDG)

24-hour load test requirements to provide a indicated load range of

103-110% of the continuous rating. The proposed amendment would also

rephrase various EDG test requirements to provide clarity and delete

the requirement to verify that the auto-connected loads do not exceed

3000 kw (Unit 2 5100kw).

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.

Changing the specification from ``unit'' to ``diesel generator''

does not change the intent of the specification, it merely clarifies

the original intent and therefore cannot involve a change in the

probability or consequences of an accident.

Changing the 22-hour lower range limit from a load of 90% to an

indicated load of 92% removes possible ambiguity from the specification

but does not change the actual requirement, therefore it cannot involve

a change in the probability or consequences of an accident.

Removing the 22-hour upper range limit from the specification does

not reduce the conservatism of the test since operating at a higher

load provides more evidence of the ability of the machine to carry the

accident loads. For this reason, this change will not involve any

increase in the consequences of an accident. Also, increasing the load

at which the diesel generator is tested cannot affect the probability

of an accident.

The NRC staff has pointed out, in Generic Letter 88-15, the hazards

of testing the Diesel Generator at a load greater than the design

rating. The proposed change is intended to ensure that the design

rating is not inadvertently exceeded. Since the recent installation of

two additional emergency diesel generators, the highest anticipated

event loads are: Unit 1-2414kW, Unit 2-3813 kW. For these diesel

generators, then, 103% of the continuous ratings:

Unit 1, 103% of 2750 kW (continuous rating) = 2832.5 kW

represents 117.3% of the highest anticipated event load and;

Unit 2, 103% of 5400 kW (continuous rating) = 5562 kW

represents 145.9% of the highest anticipated event load.

A test load of 103%, therefore would still be significantly greater

than the load required during accident conditions. Since an adequate

level of electrical load carrying capacity of the diesel generators

(and thus their accident mitigating functions) would still be

demonstrated by the surveillance test, the consequences of an accident

would be unaffected by the proposed change. The probability of

occurrence of a previously evaluated accident would be unaffected since

testing a diesel generator at load between 103 and 110 percent instead

of at load between 105 and 110 percent could not cause or contribute to

the initiation of an accident. For these reasons, this change could

have no effect on the probability or consequences of an accident

previously evaluated.

Allowing momentary transients outside of the test band does not

affect the conduct of the test, it merely allows momentary swing

outside the specified band to not invalidate the test. Not allowing

momentary transients would not prevent them, it would only require

conducting the test longer until the specified time period was achieved

without moving outside the band. Since the machine will not be operated

any differently, this specification change cannot affect the

probability or consequences of an accident previously evaluated.

Proposed changes A, B, C, D, and the first part of E [identified as

such in the submittal] are intended to clarify the meaning of the

existing specifications without changing the requirements. For this

reason, these proposed changes to the Technical Specifications will not

change the manner in which the plant is operated or maintained. These

administrative changes, therefore, will effect on the probability or

consequences of an accident previously evaluated.

The second part of E (verification of the bypass of diesel

generator trips during a simulated safety injection signal vs

concurrent safety injection and loss of offsite power signals) does not

change the intended function which is to be tested but, rather, reduces

the special conditions (temporary electrical jumpers to simulate the

loss of offsite power) in which the plant needs to be placed in order

to perform the test.

Proposed change F (removal of the verification that the auto-

connected load do not exceed 3000 or 5100 kW) does not reduce the

assurance of the ability of the diesel generators to perform the

accident mitigation functions since this verification is performed by

other, more pertinent, means.

Therefore, these changes cannot increase the probability or

consequences of an accident previously evaluated.

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

or different king of accident from any accident previously analyzed.

Changing the specification from ``unit'' to ``diesel generator''

does not change the intent of the specification, it merely clarifies

the original intent and therefore cannot create the possibility of a

new or different kind of accident.

Changing the 22-hour lower range limit from a load of 90% to an

indicated load of 92% removes possible ambiguity from the specification

but does not change the actual requirement.

Removing the 22-hour upper range limit from the specification does

not change the manner in which the surveillance is performed. It only

affects whether the time spent above 100% load can be counted toward 22

hours in the 22-hour portion of the test. This change would not allow

any new modes of operation nor does it allow any modification to the

plant.

As stated above, testing a diesel generator at a load between 103

and 110% instead of between 105 and 110% could not cause or contribute

to the initiation of an accident.

Allowing momentary transients outside of the test band does not

affect the conduct of the test, it merely allows momentary swings

outside the specified band to not invalidate the test. Not allowing

momentary transients would not prevent them, it would only require

conducting the test longer until the specific time period was achieved

without moving outside the band.

Therefore, for these reasons, operation of the facility in

accordance with the proposed amendment will not create the possibility

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

analyzed.

As stated above [for changes A-F], the proposed changes will not

cause a change in the way in which the plant is operated or maintained,

excepted for the reduction of the special conditions in which the plant

needs to be placed in order to test the bypass of the diesel generator

trips. Therefore, these administrative changes will not create the

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

previously analyzed.

3. The proposed amendment will not involve a significant reduction

in a margin of safety.

Changing the specification from ``unit'' to ``diesel generator''

does not change the intent of the specification, it merely clarifies

the original intent and therefore cannot affect the margin of safety.

Changing the 22-hour lower range limit from a load of 90% to an

indicated load of 92% removes possible ambiguity from the specification

but does not change the actual requirement and therefore cannot affect

the margin of safety.

The margin of safety is not affected by removal of the 22-hour

upper range limit on the operation of the diesel generators during

surveillance testing since the margin of safety is related to the

magnitude of the accident loads and the maximum capacity of the machine

to carry load and this margin would be unaffected by this change.

The capacity of each diesel generator to carry electrical load can

not be diminished by being tested at a lower load. Also, load testing

to less than 105% but more than 103% does not lessen the confidence in

the ability of the diesel generators to carry adequate load for this

facility since these diesel generators have significantly greater load

capacity than required by Standard Review Plan guidance in this regard

(the guidance allows peak accident load up to 100% of the continuous

rating versus Unit 1 diesel generators peak accident load of 87.8% and

Unit 2 diesel generators peak accident load of 70.6%). Therefore, this

change will not involve a significant reduction in the margin of

safety.

Allowing momentary transients outside of the test band does not

affect the conduct of the test, it merely allows momentary swings

outside the specified band to not invalidate the test. Not allowing

momentary transients would not prevent them, it would only require

conducting the test longer until the specified time period was achieved

without moving outside the band. Since the machine will not be operated

any differently per the new specification, the margin of safety is

unaffected.

As stated above [for changes A-F], the proposed changes will not

cause a change in the way in which the plant is operated or maintained,

except for the reduction of the special conditions in which the plant

needs to be placed in order to test the bypass of the diesel generator

trips. Therefore, these administrative 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 requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet mall, Minneapolis,

Minnesota 55401.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW, Washington, DC 20037.

NRC Project Director: John N. Hannon.

Omaha Public Power District

Docket No. 50-285

Fort Calhoun Station, Unit No. 1, Washington County, Nebraska.

Date of amendment request: October 7, 1994.

Description of amendment request: The proposed amendment to the

Technical Specifications (TSs) would (1) delete the surveillance

requirements contained in TS 3.6(3)a for the raw water backup valves to

the containment cooling coils, (2) delete the surveillance requirements

contained in TS 3.2, Table 3-5, item 6, for raw water valves, and (3)

revise the basis of TS 2.4 to reflect these 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) The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The deletion of surveillance requirements contained in Technical

Specifications (TS) 3.2, Table 3-5, Items 6 and 3.6(3)a does not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

TS 3.6(3)a requires the Raw Water (RW) backup valves to the

containment air coolers to be tested each refueling outage. In 1990,

during the process of reviewing several open items created by the

design basis reconstitution project, an engineering analysis determined

that RW direct cooling of the containment air cooling coils should not

be used after an accident that has created elevated temperature

conditions inside containment. The high containment air temperatures,

in conjunction with the low back pressure in the containment cooling

coils when in the RW direct cooling mode, introduces the possibility of

vaporization inside the coils. Therefore, the use of RW direct cooling

for the containment air coolers has been discontinued in post-Loss of

Coolant Accident (LOCA) or post-Main Steam Line Break (MSLB)

situations. The issue of not being able to utilize RW direct cooling to

the containment air cooling coils was reported to the NRC in LER-90-25,

dated October 29, 1990 and LER-90-25 Revision 1, dated December 17,

1990.

Raw water direct cooling of the containment air coolers is possible

if the containment atmospheric temperatures are less that 150 deg.F. If

RW direct cooling of the containment air coolers was utilized after a

LOCA or MSLB accident, it could only be used for long-term containment

atmospheric cooling. These conditions are essentially equivalent to

that associated with conditions in containment during normal plant

operation. RW direct cooling of the containment air coolers is not a

required post-accident function to maintain containment pressure below

60 psig. Since these valves are not required to perform a post-accident

function, deletion of the requirements to test these valves does not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

TS 3.2, Table 3-5, Item 6 requires that valves in the RW system be

tested every refueling outage. The valves tested by this surveillance

that could perform a safety function are already tested in accordance

with TS 3.3(1). Therefore testing of these valves under TS 3.2, Table

3-5, Item 6 is redundant to TS 3.3(1)a.

(2) The proposed changes do not create the possibility of a new or

different kind of accident from any previously analyzed.

There will be no physical alterations to the plant configuration,

changes to setpoint values, or changes to the implementation of

setpoints or limits as a result of this proposed change. Valves that

are required to be repositioned during an accident to mitigate the

consequences will still be tested on a refueling frequency. The

proposed change only deletes unnecessary or redundant testing

requirements from the TS. Therefore, the proposed change does not

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

previously analyzed.

(3) The proposed changes do not involve a significant reduction in

a margin of safety.

The proposed changes delete unnecessary or redundant surveillance

requirements within the TS. The deletion of TS 3.2, Table 3-5 Item 6,

only deletes testing requirements that are already required to be

conducted by TS 3.3(1)a. The deletion of the requirement to test the RW

backup valves to the containment air coolers in TS 3.6(3) only deletes

an unnecessary surveillance. RW direct cooling of the containment air

coolers is not required to maintain containment pressure below the

design limit of 60 psig. Therefore, the proposed 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 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: LeBoeuf, Lamb, Leiby, and MacRae, 1875

Connecticut Avenue, N.W., Washington, D.C. 20009-5728.

NRC Project Director: Theodore R. Quay.

Pennsylvania Power and Light Company

Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania.

Date of amendment request September 26, 1994.

Description of amendment request: The amendment would remove the

requirement for operability of the Average Power Range Monitors (APRMs)

while the plant is in Operational Condition 5. However, the requirement

for the APRMs to be operable during a shutdown margin demonstration,

when the mode switch is in Startup, will remain unchanged.

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

construction, which is presented below:

I. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Not requiring APRMs to be OPERABLE in OPCON 5 will not increase the

probability of inadvertent reactor critically during refueling

operations. Refueling Interlocks, NMS [Neutron Monitoring System] (SRMs

[Source Range Monitor], IRMs [Intermediate Range Monitor]), and

procedural restrictions provide assurance that inadvertent criticality

does not occur due to the simultaneous withdrawal or removal of two

control rods or due to the inadvertent insertion of a fuel bundle into

a core location with a control blade removed.

The FSAR [Final Safety Analysis Report] Section 15.4.1 discusses

the potential for a control rod withdrawal error during refueling and

start-up operations. The discussion concludes that the withdrawal of

one control rod does not require a safety action because the total

worth of one control rod is not sufficient to cause criticality. The

attempted withdrawal of two control rods, assuming an operator error

and a single active failure, would result in a control rod block

initiated by the Refueling Interlocks. The safety-related IRM

subsystem, which is required by Technical Specifications to be OPERABLE

while in OPCON 5, is designed to generate a rod block or reactor scram

on high neutron flux and is therefore a backup protective system for

the Refueling Interlocks during refueling.

The Safety-related IRM subsystem of the NMS is required by

Technical Specifications to be OPERABLE during OPCON 5 to support the

safety design bases of the NMS and RPS [Reactor Protection System]. The

SRM is not a safety-related subsystem but is important to plant safety

and is required by Technical Specifications to be OPERABLE in OPCON 5.

The SRM subsystem provides the plant operator with neutron flux levels

from startup conditions to the IRM operating range. The SRMs and IRMs

are designed to respond to local core conditions and would indicate and

respond (control rod block or scram) to an accident condition to

mitigate the transient. Thus, the APRMS are not necessary to be

OPERATOR in OPCON 5. The proposed Technical Specification change will

not alter the current requirements that the APRMs be OPERABLE during

shutdown margin demonstrations in OPCON 5 when the mode switch is in

Startup.

The proposed Technical Specification change would reduce the APRM

operability requirement in OPCON 5 and would not affect the FSAR

evaluation of the inadvertent criticality due to the withdrawal or

removal of the highest worth control rod or due to the insertion of

fuel bundles in uncontrolled cells. The FSAR concludes that the

Refueling Interlocks and plant procedures provide assurance that

inadvertent criticality does not occur during refueling.

The consequences of an accident will not be increased by the

proposed Technical Specification change because of the existing lines

of defense which prevent an inadvertent criticality event during

refueling, e.g., administrative restrictions, refueling procedures,

licensed plant operators, SRMs, Refueling Interlocks, and IRMs.

Furthermore, should the number of operator IRM or SRM channels be less

than that required by Technical Specifications, the Technical

Specifications require that core alteration activities be suspended and

all insertable control rods be inserted into the core.

Therefore, the proposed changes do not result in an increase in the

probability or consequences of an accident previously evaluated.

II. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes to the Technical Specifications will remove

the APRM operability requirement while in OPCON 5 (except for shutdown

margin demonstration testing); however, the SRMs and IRMs will still be

required to be OPERABLE in OPCON 5.

The IRMs are safety-related and are designed to detect and respond

to increases in neutron flux within the local core regions. Any

inadvertent increases in neutron flux during refueling would originate

at a local core location, i.e., rod withdrawal error or fuel bundle

insertion. Technical Specifications require IRM operability and will

generate an RPS scram or control rod block if neutron flux increased to

the setpoint. Therefore, removing the APRMs operability requirement in

OPCON 5 would not effect any safety related equipment or equipment

important to safety.

The APRMs provide core power information to the control room

operator and also provide trip signals to the RMCS [Reactor Manual

Control System] and RPS as required. The absence of an APRMs input

signal will not affect these systems during refueling operations.

Removing the APRMs operability in OPCON 5 will not affect the

response of safety-related equipment as previously evaluated in the

FSAR. The proposed changes to the Technical Specifications do not

affect any safety-related equipment or equipment important to safety.

The proposed changes to the Technical Specifications would remove

the APRMs operability requirement during refueling operations.

Technical Specifications require IRM operability and will generate an

RPS scram or control rod block if neutron flux increased to the

applicable setpoint.

No new types of accidents would be introduced since the SRMs and

IRMs are available and required to be OPERABLE in OPCON 5. Both SRMs

and IRMs would indicate and provide a control rod block or scram

signal, as appropriate, to an increase in neutron flux to mitigate a

transient event. Furthermore, should the number of OPERABLE IRM or SRM

channels be less than that required by Technical Specifications, the

Technical Specifications require that core alteration activities be

suspended and all insertable control rods be inserted into the core.

Therefore, the proposed Technical Specification changes do not

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

accident previously evaluated.

III. This change does not involve a significant reduction in a

margin of safety.

For the reasons discussed in items 1 and 2 above and because the

Technical Specification Bases do not discuss or require APRMs

operability during OPCON 5, Refueling, the proposed Technical

Specification changes do not involve a significant reduction in a

margin of safety.

The NRS 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: Osterhout Free Library, Reference

Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW., Washington, DC 20037.

NRC Project Director: John F. Stolz.

Philadephia Electric Company

Docket Nos. 50-352 and 50-353

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania.

Date of amendment request: August 22, 1994.

Description of amendment request: The amendment consists of five

(5) sections of Technical Specifications changes which reflect the

Improved Standard Technical Specifications (NUREG-1433):

Section 1: Control Rod Block Instrumentation,

Section 2: Standby Liquid Control System Operability in Mode 5,

Section 3: Scram Discharge Volume Valve Testing,

Section 4: Optional Method of Scram Timing, and

Section 5: Definition of Core Alteration.

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:

Section 1: Control Rod Block Instrumentation

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 can be divided into two general categories,

the deletion of the ``S/U'' requirements, and the change in frequency

of the SRM [Source Range Monitor] and IRM [Intermediate Range Monitor]

Calibration and Functional Tests. In each case in which the ``S/U''

requirement has been deleted, the normal surveillance frequency

specified for the required Operating Condition remains. The equipment's

associated probability of failure remains unchanged. In the case of the

surveillance frequency changes proposed for the SRMs and IRMs, the

probability of an accident evaluated in the SAR [Safety Analysis

Report] occurring does not increase since there is no credit taken in

the SAR for those Control Rod Block functions with respect to an

accident. As such, the proposed changes will not result in an increase

in the probability of occurrence of an accident previously evaluated in

the SAR. The proposed TS changes do not alter the method of operation

or performance of the equipment in carrying out associated Control Rock

Block functions. Thus, the consequences of an accident previously

evaluated in the SAR are not increased.

Therefore, the proposed TS changes do not involve an increase in

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 do not alter the configuration of the plant

or the way that the plant is operated. The equipment can perform no

other function than it is presently capable of, or cause or permit any

other accident than is now possible. Thus, the possibility of an

accident of a different type than previously evaluated in the SAR

cannot be created.

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.

Since the proposed TS changes affect only the surveillance

frequency intervals and do not change the plant configuration or

associated instrument setpoints, there is no quantitative or

qualitative reduction in the margin of safety. Thus, the margin of

safety as defined in the bases of any Technical Specification is not

reduced.

Therefore, the proposed TS changes do not involve a reduction in a

margin of safety.

Section 2: Standby Liquid Control System Operability in Mode 5

1. The proposed Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed TS change will remove the SLCS operability requirement

in OPCON 5. The purpose of the SLCS is to bring the reactor to and

maintain it in a cold shutdown condition from normal power operations

following failure to scram during power operations. Initiation of the

SLCS is not a precursor to any accident. Therefore, inoperability of

the SLCS in OPCON 5 cannot increase the probability of an accident

previously evaluated.

The proposed TS change does not involve a physical change in any

system's configuration and no new modes of operation are introduced.

The SLCS has not analyzed function OPCON 5. The probability of fuel

failure will not be increased by this change. Shutdown margin, in

conjunction with TS requirements and procedural controls, will assure

that an inadvertent criticality event will not occur during refueling.

In addition, the Reactor Protection System (RPS) and Control Rod System

will provide protection in the unlikely event that an inadvertent

criticality should occur.

Therefore, the proposed TS change does not involve an increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed TS changes does not involve a physical change in any

system's configuration and no new modes of operation are introduced.

The SLCS's only purpose is to mitigate the consequences of a failure to

scram during power operation. In OPCON 5, the SLCS has no analyzed

function, therefore, the proposed TS change will not create the

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

evaluated.

3. The proposed TS change does not involve a significant reduction

in a margin of safety.

The purpose of the SLCS is to bring the reactor to and maintain it

in a cold shutdown condition from normal power operations following a

failure to scram during power operations. The SLCS is not designed to

terminate an inadvertent criticality during OPCON 5. Shutdown margin,

either demonstrated or analytically determined, in conjunction with

Technical Specifications and procedural controls, will assure that an

inadvertent criticality event will not occur during refueling

operations. In addition, the RPS and Control Rod System, which are

extremely reliable, will provide protection in the unlikely event that

an inadvertent criticality does occur. Therefore, the proposed TS

change does not involve a reduction in a margin of safety.

Section 3: Scram Discharge Volume Valve Testing

1. The proposed Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The Scram Discharge Volume (SDV) is not an accident initiator.

Deletion of the requirement that the SDV be determined OPERABLE by

testing the SDV vent and drain valves when control rods are scram

tested from a normal control and configuration of less than or equal to

50% rod density at least once per 24 months, as proposed, will have no

effect on the probability or consequences of an accident previously

evaluated.

This proposed TS will have a negligible impact on the conditions

experienced by the vent and drain valves as they stroke closed, since

the SDV is initially vented to the atmosphere, and the valves close

before the SDV becomes pressurized, even during a scram at full reactor

power. Reactor pressure and Control Rod Drive (CRD) discharge flow

conditions do not influence the SDV vent and drain closure rates, since

the SDV is of sufficient volume and initially vented such that peak

pressure prior to the SDV complete isolation will not be substantial.

In addition, lower coolant temperatures expected during testing at

shutdown conditions will also have a negligible impact on the

performance of the test. Although, there could be some variation in the

performance [of] the SDV vent and drain valves to re-open when

performing the test during shutdown conditions, as opposed to

conducting the test during power operation, the ability of the valves

to re-open is demonstrated after each reactor scram during power

operation.

In the event and SDV vent or drain valve failed to open, increasing

SDV level during reactor operation would cause 1) an alarm in the Main

Control Room (MCR), 2) a control rod block, and finally a reactor scram

initiated by the Reactor Protection System (RPS) if action is not taken

to drain the SDV. Therefore, the ability to shut down the reactor is

not impaired. If a SDV vent or drain valve fails to close, the

redundant valve's closure would provide the required function. If both

valves failed to close, a loss of reactor coolant in the form of water

discharged from the CRD system would occur. The amount of water

discharged will be relatively small, and is more of a concern from the

standpoint of contamination to the Secondary Containment rather than a

loss of reactor water inventory. A structural failure of the SDV, which

bounds this case of an open SDV vent or drain line, has been previously

evaluated in NUREG-0808, ``Generic Safety Evaluation Report Regarding

Integrity of BWR Scram System Piping.'' In this evaluation, the NRC

concluded that, for a bounding leakage case corresponding to a rupture

of the SDV, the offsite doses would be well within the limits of

10CF100, and that adequate core cooling would be maintained.

Deletion of the requirement that the SDV be determined OPERABLE by

testing the SDV vent and drain valves, as proposed in this TS Change

Request, will have an insignificant effect on the probability of

occurrence of malfunction of any plant equipment. The conditions in the

SDV at the time of vent and drain valve closure are not appreciably

different whether a scram is initiated from power operation or during

shutdown conditions. In addition, this proposed TS change eliminates

the potential need for an additional startup and shutdown cycle, along

with the associated challenges to all systems and components, that

would be required to satisfy the current TS requirements in the event a

unit were to trip off-line shortly before a planned outage when the

surveillance was scheduled to be performed. Furthermore, this proposed

TS changes does not affect the testing frequency for the valves.

This proposed TS change will not result in appreciably different

conditions experienced by the valves as they close, and their ability

to re-open is confirmed following each reactor scram from power

conditions. The consequences resulting from a failed closed or failed

open SDV vent or drain line have been evaluated and determined not to

result in offsite doses that would exceed the limits specified in

10CFR100, or jeopardize adequate reactor core cooling capability.

Therefore, the consequences of a malfunction of equipment important to

safety previously evaluated is not increased.

Therefore, the proposed TS change does not involve an increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The SDV is not an accident initiator. Deletion of the requirement

that the SDV be determined OPERABLE by testing the SDV vent and drain

valves from a configuration of less than or equal to 50% rod density,

as proposed, will not create the possibility of a different type [of]

accident than any previously evaluated.

No plant equipment is added or deleted as a result of this proposed

change. Since the initial conditions of pressure, temperature, and CRD

system discharge flowrate have no appreciable effect on the SDV vent

and drain valve performance, no different type of malfunction of any

equipment important to safety is created.

Therefore, the proposed TS change does not create the possibility

of a new different kind of accident from any previously evaluated.

3. The proposed TS change does not involve a significant reduction

in a margin of safety.

Since the initial test conditions of pressure, temperature, and CRD

discharge flowrate will have no appreciable effect on the SDV vent and

drain valve performance, conducting the surveillance test during

shutdown conditions, as specified in this proposed TS change, will not

affect the validity of the surveillance results with respect to the

operability of the SDV to perform its intended safety function.

Furthermore, every reactor scram is a serious plant transient and a

potential challenge to safety-related systems and equipment. The

potential decrease in future scrams which could result from this

proposed TS change will represent an improvement in overall safety.

Therefore, the proposed TS change does not involve a reduction in a

margin of safety.

Section 4: Optional Method of Scram Timing

1. The proposed Technical Specification (TS) changes involves a

significant increase in the probability or consequences of an accident

previously evaluated.

Scram testing control rods at zero reactor coolant pressure will

not increase the probability of any control rod related transient or

accident discussed in the UFSAR [Updated Final Safety Analysis Report].

UFSAR Sections 15.4.1.1 and 15.4.1.2 discuss the consequences of

inadvertent reactivity insertion errors due to the withdrawal of one or

more control rods. The probability of one of these events occurring is

a function of operator error and equipment malfunction and is not

related to scram insertion times.

An inadvertent reactivity insertion error is prevented by existing

system hardware interlocks and procedural controls that are not

affected by scram time testing, e.g., core design, control and design,

one-rod-out interlocks, refueling interlocks, control rod sequence

designations, and neutron monitoring systems.

USFAR Section 15.4.9 discusses the control rod drop accident

(CRDA). The CRDA assumes that a control rod suddenly drops out of the

core due to equipment malfunction. The probability of occurrence of

this accident is based on an equipment malfunction and is not affected

by scram testing.

Engineering analysis and control rod scram test data demonstrate

that a control rod drive that will meet the 2.0 second, scram insertion

time, test criteria at zero reactor coolant pressure will also meet all

scram insertion criteria during reactor startup and up to 40% rated

thermal power.

The 2.0 second criterion was chosen to conservatively envelop scram

time criteria and reactivity insertion criteria during reactor startup

and up to 40% rated power conditions. Therefore, scram testing affected

control rods at zero reactor pressure will not increase the

consequences of an accident previously evaluated.

UFSAR Sections 15.4.1.1 and 15.4.1.2 evaluate reactivity insertion

transients at low power conditions due to inadvertent control rod

withdrawal errors. The UFSAR concludes that rod withdrawal errors at

low power are adequately precluded by refueling interlocks, rod worth

minimizer, operating procedures, core design, and control rod hardware

design. However, should operator errors followed by equipment

malfunctions result in an inadvertent criticality event, the IRMs would

provide the necessary rod blocks or reactor scram to preclude the

operational transient. Scram insertion time limits for the continuous

rod withdrawal error during startup is 5.0 seconds. This scram time

criterion will be met by a control rod that scrams within 2.0 seconds

at zero reactor pressure. The 2.0 second scram criterion was

established to ensure that affected control rods will meet scram

requirements from zero reactor pressure up to 40% core thermal power.

Also, during low power operation (UFSAR Subsection 15.4.1.2) the

rod worth minimizer (RWM) prevents the operator from selecting and

withdrawing an out-of-sequence control rod. During reactor operation in

the power range (UFSAR subsection 15.4.2) the rod block monitor (RBM)

prevents a rod withdrawal error by inhibiting inadvertent control rod

withdrawal. The RWM and RBM do not rely on a scram function to

mitigation the consequences of a rod withdrawal error, and therefore

the consequences of an accident evaluated in the UFSAR will not be

affected by the proposed changes to the Technical Specifications.

The consequences of a control rod drop accident (UFSAR Section

15.4.9) would not be affected by scram testing a control rod at zero

reactor pressure. The design basis accident of the rod drop accident

assumes that control rods scram within 5.0 seconds. This 5.0 second

scram test requirement will be met by control rods that meet the 2.0

second criterion at zero reactor pressure.

Therefore, the proposed TS changes do not involve an increase in he

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The changes to the Technical Specifications will allow control rods

to be scram tested at zero reactor pressure and then again at rated

reactor pressure prior to achieving 40% rated reactor power. No new

types of accidents will be introduced since control rods that meet the

2.0 second scram criterion at zero reactor pressure will also meet all

scram test criteria during reactor startup and at rated reactor

pressure.

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 basis for shutdown margin (TS Bases 3/4.1.1) states that the

reactor shall be made subcritical by all certain margin in all

operating and shutdown conditions. The proposed changes to the

Technical Specifications will not affect the shutdown margin

requirements. Adequate shutdown margin is assured by core design, the

one-rod-out interlock, and administrative controls.

The basis for the control rod insertion times (TS Bases 3/4.1.3)

states that the scram times are to be consistent with those used in the

transient and accident analysis. The proposed Technical Specifications

changes will add an additional scram test verification for affected

control rods at zero reactor pressure. The zero reactor pressure scram

limit (2.0 seconds) was designed to ensure that the scram times assumed

in the transient analysis will remain bounding from zero reactor

pressure up to 40% rated core thermal power.

The basis for the control rod drop accident (TS Bases 3/4.1.3)

states that the potential effects of a CRDA are limited. The proposed

Technical Specifications changes will not effect the control rod drop

results as the changes do not affect the reactivity of the rod or the

rod drop velocity. The CRDA analysis is based on a 5.0 second scram

insertion time criterion. The 2.0 second time criterion was established

to ensure that the 5.0 second scram time criterion was valid from zero

reactor pressure to 950 psig reactor pressure.

The basis for MCPR limits (TS Bases 3/4.1.3 and 2.3) states the CRD

system must bring the reactor subsubcritical at a rate fast enough to

prevent MCPR from becoming less than the fuel cladding safety limit

during the limiting power transient analyzed in the UFSAR. The proposed

changes to the Technical Specifications will not affect the scram

insertion rates that are used as input to the transient analysis. The

zero reactor pressure scram limit of 2.0 seconds was developed to

ensure that the control rods would meet their design scram insertion

times from zero reactor pressure up to 40% rated power.

The proposed changes to the Technical Specifications will not

increase the probability of inadvertent criticality because the changes

do not affect the reactivity worth of control rods.

Therefore, the proposed TS changes do not involve a reduction in a

margin of safety.

Section 5: Definition of Core Alteration

1. The proposed TS change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The proposed definition change removes the requirement to have a

SRO or LSRO supervise control rod withdrawal in an off-loaded cell

(i.e. no fuel assemblies). The evaluated accident potentially affected

by this change is a control rod movement error during refueling

resulting in inadvertent criticality. The supervision by a SRO or LSRO

does not solely preclude inadvertent criticality and was not relied

upon in the accident analysis contained in Section 15.4 of the LGS

Updated Final Safety Analysis Report (UFSAR). The LGS reactor core is

designed to have adequate shutdown margin with the highest-reactivity-

worth control rod withdrawn. The withdrawal of a second rod with fuel

assemblies loaded in the associated cell is prevented by a combination

of the refueling, one-rod-out interlock, and the Limiting Conditions

for Operation (LCO) requirement of TS 3.9.10.2. The LCO requirements

ensure adequate shutdown margin is present prior to control rod

withdrawal. This is accomplished by testing during startup following a

refueling outage or by analytical calculations during refueling. The

refueling interlock will provide a rod block upon an attempt to

withdraw a second control rod and is required to be operable in

accordance with TS 3.9.10.2 except for rods which have no fuel

assemblies in the associated cell. The removal of the fuel assemblies

from a cell eliminates the need for the reactivity control function of

the associated rod. The physical removal of a control blade from the

core by means of the refueling floor, first requires the removal of the

four associated fuel assemblies in the cell. This design inherently

prevents inadvertent criticality. Finally, this change is consistent

with NUREG-1433 ``Standard Technical Specifications.'' Since current

analysis permits the withdrawal of a control rod blade, provided the

associated cell is unloaded, and refueling mode interlocks,

administrative TS requirements and the physical design of the control

blade and fuel cell, which preclude inadvertent criticality, will

remain unchanged, this proposed change to the TS definition of CORE

ALTERATION will not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The LGS UFSAR currently permits control rod withdrawal and or

removal, provided there are no fuel assemblies in the associated fuel

cell. The definition change removes the requirement to have a SRO or

LSRO supervise rod withdrawal in an off-loaded cell. The change

potentially [a]ffects a control rod movement error during refueling

resulting in inadvertent criticality which has been previously

evaluated. In addition, the proposed change will make no physical

changes to equipment or remove administrative controls which solely

preclude inadvertent criticality. Therefore, this change will not

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

accident previously evaluated.

3. The proposed TS change does not involve a significant reduction

in a margin of safety.

The LGS TS bases address reactivity concerns, radiological

releases, control rods, and monitoring of the facility related to this

change. With the four fuel assemblies removed from a cell, the control

rod/blade in the associated cell has no reactivity function. The

reactivity issues addressed by TS are therefore unaffected. The rod/

blade coupling integrity is maintained by the requirement to perform a

coupling check following maintenance. Section 15.4 of the UFSAR states

that there are no radiological releases in association with a rod

withdrawal error during refueling. This conclusion is maintained by the

administrative requirements of TS 3.9.10.2, the refueling interlocks

for one-rod-out, and the physical design of the blade and cell. Lastly,

the TS requirements for Emergency Core Cooling, Plant System,

Containment, and Electrical Power Distribution System, which provide

the systems necessary to mitigate the effects of a radiological release

during control rod movement in an unloaded cell were reviewed and were

found not to be adversely [a]ffected by the proposed change. Therefore,

this 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Philadelphia Electric Company

Docket Nos. 50-352 and 50-353

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania.

Date of amendment request: August 31, 1994.

Description of amendment request: The proposed amendments, which

are consistent with the Improved Standard Technical Specifications

(NUREG-1433), involve the following six (6) sections of TS changes:

Section 1: Relocation of Turbine Overspeed Protection System

Requirements;

Section 2: Relocation of Primary Containment Conductor Protection

Devices Requirements;

Section 3: Feedwater/Main Turbine Trip System Actuation Instrumentation

Requirements;

Section 4: Permit Operability of Low Pressure Coolant Injection While

Aligned to Shutdown Cooling;

Section 5: Remove Temperature Requirement for Operational Condition

[OPCON] 5; and

Section 6: Reduce Frequency of Alternate Decay Heat Demonstration

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:

Section 1: Relocation of Turbine Overspeed Protection System

Requirements

1. The proposed TS change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The proposed change relocates requirements from the TS, to licensee

controlled documents. The licensee controlled documents containing the

relocated requirements will be maintained using the provisions of 10

CFR 50.59 and are subject to the change control process in the

Administrative Controls Section 6.0 of the TS. Since changes to

licensee controlled documents will be evaluated per 10 CFR 50.59, no

increase (significant or insignificant) in the probability or

consequences of an accident previously evaluated will be allowed.

Therefore, this change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident previously evaluated.

This change relocates requirements to licensee controlled

documents. This change will not alter the plant configuration (no new

or different type of equipment will be installed) or make changes in

methods governing normal plant operation. This change will not impose

different requirements and adequate control of information will be

maintained. This change will not alter assumptions made in the safety

analysis and licensing basis. Therefore, this change will not create

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

accident previously evaluated.

3. The proposed TS change does not involve a significant reduction

in a margin of safety.

This change relocates requirements from the TS to licensee

controlled documents. This change will not reduce a margin of safety

since it has no impact on any safety analysis assumptions. In addition,

the requirements to be transferred from the TS to licensee controlled

documents are the same as the existing Technical Specifications. Since

any future changes to these licensee controlled documents will be

evaluated per the requirements of 10 CFR 50.59, no reduction

(significant or insignificant) in [a] margin of safety will be allowed.

Therefore, this change will not involve a significant reduction in a

margin of safety.

The existing requirements for NRC review and approval of revisions,

in accordance with 10 CFR 50.59, to these details and requirements

proposed for relocation, does not have a specific margin of safety upon

which to evaluate. However, since the proposed change is inconsistent

with the BWR [boiling-water reactor] Improved Standard Technical

Specifications (NUREG-1433 approved by the NRC Staff) and the change

controls for proposed relocated details and requirements provide an

equivalent level of regulatory authority, revising the TS to reflect

the approved level of detail and requirements ensures no reduction to

the margin of safety.

Section 2: Relocation of Primary Containment Conductor Protection

Devices Requirements

1. The proposed TS change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

This proposed change relocates requirements from the TS to licensee

controlled documents. The licensee controlled documents containing the

relocated requirements will be maintained using the provisions of 10

CFR 50.59 and are subject to the change control process in the

Administrative Controls Section 6.0 of the TS. Since changes to these

licensee controlled documents will be evaluated per 10 CFR 50.59, no

increase (significant or insignificant) in the probability or

consequences of an accident previously evaluated will be allowed.

Therefore, this change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

This change relocates requirements to licensee controlled

documents. This change will not alter the plant configuration (no new

or different type of equipment will be installed) or make changes in

methods governing plant operation. This change will not impose

different requirements and adequate control of information will be

maintained. This change will not alter assumptions made in the safety

analysis and licensing basis. Therefore, this change will not create

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

accident previously evaluated.

3. The proposed TS change does not involve a significant reduction

in a margin of safety.

This change relocates requirements from the TS to licensee

controlled documents. This change will not reduce a margin of safety

since it has no impact on any safety analysis assumptions. In addition,

the requirements to be transferred from the TS to the licensee

controlled documents are the same as the existing TS. Since any future

changes to these licensee controlled documents will be evaluated per

the requirements of 10 CFR 50.59, no reduction (significant or

insignificant) in [a] margin of safety will be allowed. Therefore, this

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

The existing requirements for NRC review and approval of revisions,

in accordance with 10 CFR 50.59, to these details and requirements

proposed for relocation, does not have a specific margin of safety upon

which to evaluate. However, since the proposed change is consistent

with the BWR Improved Standard TS (NUREG-1433 approved by the NRC

Staff) and the change controls for proposed relocated details and

requirements provide an equivalent level of regulatory authority,

revising the TS to reflect the approved level of detail and

requirements ensures no reduction to the margin of safety.

Section 3: Feedwater/Main Turbine Trip System Actuation Instrumentation

Requirements

1. The proposed TS change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

For the proposed TS change, in the event of a Reactor Vessel Water

Level--High Level 8 transient, operator action per existing plant

procedures would terminate the event and prevent damage to the Main/RFP

[reactor feed pump] Turbine due to water carry over. The Main/RFP

Turbine do not serve a safety function, also at 25% RTP. Therefore, the proposed TS change

to the operability requirements for the feedwater/main turbine trip

system actuation instrumentation does not involve a reduction in a

margin of safety.

Section 4: Permit Operability of Low Pressure Coolant Injection While

Aligned to Shutdown Cooling

1. The proposed Technical Specifications change does not involve a

significant increase in the probability or consequences of an accident

previously evaluated.

The LPCI [low pressure coolant injection] mode of RHR is an

accident mitigator, not an initiator. Currently, the LPCI mode of RHR

is an automatic Emergency Core Cooling System (ECCS) function during

OPCONs 4 and 5. However, shutdown cooling has been an accident

initiator in many industry events. Reliance on this loop of RHR for

LPCI does not increase the probability of an accident in shutdown

cooling, but the alignment for LPCI will, in itself, terminate the

draindown event by exiting the shutdown cooling mode. This proposed

change will permit the operability of one LPCI subsystem while the

components of that subsystem are aligned and operating in the Shutdown

Cooling mode of RHR, provided all other components of that subsystem

are operable and can be manually realigned from the Main Control Room,

if required. The required number of operable Emergency Core Cooling

Systems (ECCS) remains unchanged, thus maintaining the TS required

subsystem redundancy (TS Section 3.5.2 requires two operable ECCS

subsystems with exception for Reactor level). With this change, the

required number of LPCI subsystems are capable of performing their

function of limiting and/or mitigating the consequences of an accident,

by allowing the manual alignment of one LPCI subsystem, during OPCONs 4

and 5. This allowance is justified since the change only applies to

OPCONs 4 and 5, when reactor temperature, and associated heat loads are

sufficiently low to provide the operator sufficient time to perform the

manual realignment, from the Main Control Room, of the RHR pump suction

valves and restart of the pump following LPCI injection conditions.

Similar allowances for LPCI are currently permitted during OPCON 3,

since the decay heat loads are significantly reduced compared to OPCON

1, which is the mode of operation under which ECCS capability is

analyzed (Section 6.3 of the LGS [Limerick Generating Station] Updated

Final Safety Analysis Report (UFSAR)). The change will not increase the

probability of occurrence or consequences of a malfunction of equipment

since there will be no physical changes made to plant equipment nor the

method of their operation that would result in an unanalyzed condition.

PECO Energy [Philadelphia Electric Company] evaluated the need for

manual realignment of the pump minimum flow path since operating in

Shutdown Cooling typically results in the isolation of the pump minimum

flow path to prevent inadvertent draining of the reactor vessel. The

associated pump is still operable since this change is limited to

OPCONs 4 and 5, when reactor pressure is sufficiently low to allow

immediate injection to the reactor vessel without a minimum flow path.

In situations, while in OPCON 4, where reactor pressure may not be

sufficiently low to allow injection, the RHR system will not be aligned

for Shutdown Cooling, since the reactor vessel pressure will be greater

than the RHR ``cut-in'' permissive pressure. In addition,

Administrative Controls are currently in place to realign RHR to the

LPCI mode for planned pressure increases. Finally, this change is

consistent with NUREG-1433 ``Standard Technical Specifications.''

Therefore, these changes will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The LPCI mode of RHR is an accident mitigator, not an initiator.

This change will not reduce the number of required ECCS during OPCONs 4

and 5. This change will permit the operability of one LPCI subsystem

while the components of that subsystem are aligned and operating in the

Shutdown Cooling mode of RHR. The change does not alter current methods

of plant operation nor does the change make a physical change to plant

equipment resulting in an unanalyzed malfunction of equipment.

Therefore, this change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed TS change does not involve a significant reduction

in a margin of safety.

The basis of TS Section 3.5.2 is to ensure sufficient ECCS capacity

to maintain core cooling in OPCONs 4 and 5. This proposed change does

not affect the required number of ECCS during OPCONs 4 and 5;

therefore, adequate capability through subsystem redundancy is

maintained. The amount of time required to obtain rated LPCI conditions

is increased due to the manual realignment, from the Main Control Room,

of the suction valves and restart of the RHR pump following LPCI

injection conditions. This change is in conformance with the current TS

bases, since the operator has sufficient time to perform the manual

realignment, during OPCONs 4 and 5, ensuring sufficient ECCS capability

to maintain core coverage. In addition, NUREG-1433 BASES states, in

part, ``One LPCI subsystem may be aligned for decay heat removal and

considered OPERABLE for the ECCS function, if it can be manually

realigned (remote or local) to the LPCI mode and is not otherwise

inoperable. Because of low pressure and low temperature conditions in

MODES 4 and 5, sufficient time will be available to manually align and

initiate LPCI subsystem operation to provide core cooling prior to

postulated fuel uncover.'' Therefore, this change will not involve a

significant reduction in a margin of safety.

Section 5: Remove Temperature Requirement for Operational Condition 5

1. The proposed Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed TS change does not involve a physical change in the

configuration of any systems important to safety. The elimination of a

temperature requirement from the definition of OPCON 5 was reviewed for

potential effect on reactor coolant system materials and for potential

effect on reactivity. This TS change does not result in system

temperature and pressure change or reactivity changes not previously

analyzed. The reactor pressure vessel will still be restricted to the

temperature and pressure limits of TS Section 3/4.4.6 which includes

heatup/cooldown rates and minimum boltup limits. The reactor pressure

vessel temperature and pressure limits will still ensure proper

protection of the reactor coolant system materials. Therefore, this TS

change does not increase the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed TS change does not involve any physical change in

plant configuration, and reactor coolant system temperature and

pressure are still restricted per TS Selection 3/4.4.6. The decrease in

moderator density corresponding to the potential change in temperature

(i.e., above 140 deg.F and below 200 deg.F) would have a negligible,

however conservative effect on shutdown margin. Therefore, this TS

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

accident from any accident previously evaluated.

3. The proposed TS change does not involve a significant reduction

in a margin of safety.

This proposed TS change does not change the reactor coolant system

material restrictions as defined in TS Section 3/4.4.6. Therefore, the

reactor pressure vessel will still be maintained under the current

temperature and pressure restrictions as well as the current boltup

limits.

The decrease in moderator density corresponding to the potential

temperature change from 140 deg.F to 200 deg.F is insignificant and

would afford approximately the same moderator effect. Therefore,

shutdown margin could only be improved (although marginally) at these

evaluated temperatures. The actual coolant temperature will be

administratively controlled to provide for personnel safety. Therefore,

this change will not involve a reduction in a margin of safety.

Section 6: Reduce Frequency of Alternate Decay Heat Demonstration

1. The proposed TS change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The proposed TS change does not involve any physical changes to

plant systems or equipment. This proposed TS change will allow the use

of either an ``analytical approach'' (i.e., calculation) or

``demonstration'' to ensure the operability of an alternate decay heat

removal method. This proposed TS change does not involve any physical

changes to plant systems or components, nor does it affect the

capability, availability, or operability of any decay heat removal

systems/methods (e.g., Shutdown Cooling). The Shutdown Cooling mode of

operation of the Residual Heat Removal (RHR) system, and Residual Heat

Removal Service Water (RHRSW) system, are not impacted by this proposed

TS change, and will continue to function as designed to remove decay

heat loads from the reactor primary coolant system. The RHRSW system

and various modes of operation of the RHR system, e.g., Low Pressure

Coolant Injection (LPCI) are not accident initiators, since these

systems function to mitigate the consequences of an accident. This

proposed TS change is consistent with the criteria delineated in the

Improved Standard TS (i.e., NUREG-1433, ``Standard Technical

Specifications, General Electric Plants, BWR/4,'' dated September 28,

1992).

Therefore, the proposed TS change does not involve an increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

This proposed TS change does not involve any physical changes to

plant systems or equipment. The proposed TS change will allow the use

of a ``calculation'' or ``demonstration'' as the means for determining

the operability of an alternate decay heat removal method. The proposed

TS change does not involve any physical changes to plant systems or

equipment. This proposed TS change will not affect the operation of the

Shutdown Cooling mode of the RHR system. This mode of operation will

continue to function as designed to remove decay heat loads from the

reactor primary coolant system. This proposed TS change will not impact

the operation of the other modes of operation of the RHR system (e.g.,

LPCI), nor will it affect the operation of the RHRSW system. These

systems will continue to function as designed, which is to mitigate the

consequences of an accident. This proposed TS change will not introduce

the potential for equipment malfunctions or failures. This proposed TS

change is consistent with the criteria delineated in the Improved

Standard TS (i.e., NUREG-1433).

Therefore, the proposed TS change does not create the possibility

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

3. The proposed TS change does not involve a significant reduction

in a margin of safety.

The proposed change to the TS does not involve any physical changes

to plant systems or equipment. This proposed TS change does not make

any physical modifications to plant systems or equipment, and is

consistent with the criteria delineated in the Improved Standard TS

(i.e., NUREG-1433). The proposed TS change will not impact any mode of

operation of the RHR system or the RHRSW system.

This proposed TS change involves revising TS ACTION statements, and

associated supporting Bases sections, to allow for the use of a

``calculation'' or ``demonstration'' to ensure the operability of an

alternate decay heat removal method. The bases for the TS sections

affected by this proposed change indicate that sufficient heat removal

capability, system redundancy, and coolant circulation will be

available to facilitate decay heat removal and mixing to assure

accurate temperature indication.

This proposed TS change does not affect the function or

availability of any decay heat removal system or method.

Therefore, the proposed TS change does not involve a 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Power Authority of the State of New York

Docket No. 50-333

James A. FitzPatrick Nuclear Power Plant, Oswego County, New York.

Date of amendment request: October 3, 1994.

Description of amendment request: The proposed amendment would

extend the functional test intervals and allowable out-of-service times

for some of the instruments subject to requirements of the Technical

Specifications (TSs). These proposed changes are based upon NRC-

approved Licensing Topical Reports prepared under the direction of the

Boiling Water Reactors Owners Group and intended to enhance plant

safety by reducing the potential for test related scrams, excessive

test cycles on equipment, and operator errors. The proposed amendment

would also: (1) Remove the Average Power Range Monitor (APRM) downscale

scram function from the TSs, remove instrument response time values

from the TSs in accordance with Generic Letter 93-08, and incorporate

various editorial changes and clarifications into the TSs. The proposed

amendment involves reactor protection system, primary containment

isolation, emergency core cooling, control rod block, and anticipated

transient without scram recirculation pump trip instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the proposed

Amendment would not involve a significant hazards consideration as

defined in 10 CFR 50.92, since it would not:

1. involve a significant increase in the probability or

consequences of an accident previously evaluated because:

a. Incorporate STI [Surveillance Test Interval] and AOT [Allowable Out-

Of-Service Time] Improvement--Category 1

The proposed changes are limited to an extension of the

surveillance testing intervals and allowable out-of-service times of

plant instrumentation. The changes do not introduce any new modes of

plant operation, make any physical changes, or alter any operational

setpoints. Therefore, the changes do not degrade the performance of any

safety system assumed to function in the accident analysis.

Consequently, there is no effect on the probability of occurrence of an

accident.

Regarding the consequences of an accident, the GE [General Electric

Company] Licensing Topical Reports (References 1 through 7) concluded

that the proposed extensions in the STI and AOT for the safety sy

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