Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating

LicensesInvolving 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 August 30, 1995, through September 15, 1995.

The last biweekly notice was published on Wednesday, September 13, 1995

(60 FR 47613).

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

[[Page 49930]]

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

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

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

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

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

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

intervene is discussed below.

By October 27, 1995, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

petition

[[Page 49931]]

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

Regulatory Commission, Washington, DC 20555, and to the attorney for

the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Carolina Power & Light Company, et al., Docket No. 50-324,

Brunswick Steam Electric Plant, Unit 2, Brunswick County, North

Carolina

Date of amendment request: August 4, 1995

Description of amendment request: The proposed amendment will allow

the loading and use of GE13 fuel assemblies in the Brunswick Steam

Electric Plant (BSEP), Unit 2, during Cycle 12 operation. The use of

GE13 fuel assemblies requires that the safety limit value for minimum

critical power ratio be revised. This safety limit is established to

maintain fuel cladding integrity. Use of GE13 fuel also requires an

increase in the concentration of sodium pentaborate solution required

by the Technical Specifications (TS) for the standby liquid control

system. This change provides the additional shutdown reactivity

necessary to permit use of this fuel.

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.

Proposed Change 1:

The proposed amendment will allow the loading and use of GE13

fuel assemblies in the Brunswick Unit 2 reactor core. The use of

GE13 fuel assemblies requires that the safety limit minimum critical

power ratio value also be revised. The safety limit minimum critical

power ratio is established to maintain fuel cladding integrity. The

GE13 fuel assembly design has been analyzed using methods that have

been previously approved by the Nuclear Regulatory Commission and

documented in General Electric Nuclear Energy's reload licensing

methodology Topical Report (NEDE-24011-P-A-10, ``General Electric

Standard Application for Reactor Fuel (GESTAR II)'' dated February

1991).

The proposed revision of the safety limit minimum critical power

ratio does not alter any plant safety-related equipment, safety

function, or plant operations that could change the probability of

an accident. The change does not affect the design, materials, or

construction standards applicable to the fuel bundles in a manner

that could change the probability of an accident.

A methodology that has been previously reviewed and accepted by

the Nuclear Regulatory Commission was used to derive the both

existing and updated safety limit minimum critical power ratio

value. The same methodology criteria have been applied to derive the

existing safety limit minimum critical power ratio of 1.07 as that

used to derive the updated safety limit minimum critical power ratio

value of 1.09. The updated safety limit minimum critical power ratio

assures that fuel cladding protection equivalent to that provided

with the existing safety limit minimum critical power ratio value is

maintained. This ensures that the consequences of previously

evaluated accidents are not significantly increased.

Proposed Change 2:

The standby liquid control system provides a means of reactivity

control that is independent of the normal reactivity control system.

The standby liquid control system must be capable of assuring that

the reactor core can be placed in a subcritical condition at any

time during reactor core life. Technical Specification Figure 3.1.5-

1 specifies the acceptable range of concentrations and volumes for

sodium pentaborate solution used as a neutron absorber (i.e., for

reactivity control). The portion of the sodium pentaborate

concentration range shown in Technical Specification Figure 3.1.5-1

applicable to the lower range of tank volumes is being revised to

increase the required concentration of sodium pentaborate solution.

This change is needed to account for the additional shutdown

reactivity needed based on the planned use of GE13 fuel assemblies

as reload fuel for the Unit 2 reactor core. Since the standby liquid

control system is independent from the normal means of controlling

reactor core reactivity and not used to control core reactivity

during normal plant operations, the proposed revision to the sodium

pentaborate concentration curve for the standby liquid control

system does not alter any plant safety-related equipment, safety

function, or plant operations that could change the probability of

an accident.

The current volume-concentration range of sodium pentaborate

used in the standby liquid control system will achieve a sufficient

concentration of boron in the reactor vessel to ensure reactor

shutdown. Based on the increased reactivity of the new GE13 reload

fuel assemblies, the required sodium pentaborate volume-

concentration range is being revised to ensure sufficient neutron

absorbing solution is available to achieve reactor shutdown;

therefore, the consequences of an accident previously evaluated are

not significantly increased.

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

new or different kind of accident from any accident previously

evaluated.

Proposed Change 1:

The GE13 fuel assembly has been designed and complies with the

acceptance criteria contained in General Electric Nuclear Energy's

standard application for reactor fuel (GESTAR-II), which provides

the latest acceptance criteria for new General Electric fuel

designs. The GE13 fuel assembly complies with GESTAR-II acceptance

criteria that have been previously reviewed and accepted by the

Nuclear Regulatory Commission. The similarity of the GE13 fuel

design to the previously accepted GE11 fuel design, in conjunction

with the increased critical power capability of the GE13 fuel

design, ensure that no new mode or condition of plant operation is

being authorized by the loading and use of the GE13 fuel type. The

proposed revision of the safety limit minimum critical power ratio

from 1.07 to 1.09 does not modify any plant controls or equipment

that will change the plant's responses to any accident or transient

as given in any current analysis. Therefore, the proposed change to

allow the loading and use of the GE13 fuel type and the revision of

the safety limit minimum critical power ratio value from 1.07 to

1.09 will not create the possibility for a new or different kind of

accident from any accident previously evaluated.

Proposed Change 2:

As discussed above, the standby liquid control system provides a

means of reactivity control that is independent of the normal

reactivity control system and is capable of assuring that the

reactor core can be placed in a subcritical condition at any time

during reactor core life. The proposed revision to the sodium

pentaborate concentration range does not modify the standby liquid

control system or its controls, does not modify other plant systems

and equipment, and does not permit a new or different mode of plant

operation. As such, the proposed revision to the minimum pentaborate

concentration value does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety.

Proposed Change 1:

As previously discussed, the GE13 fuel assembly design has been

analyzed using methods that have been previously approved by the

Nuclear Regulatory Commission and documented in General Electric

Nuclear Energy's reload licensing methodology Topical Report (NEDE-

24011-P-A-10, ``General Electric Standard Application for Reactor

Fuel (GESTAR II)'' dated February 1991). The safety limit minimum

critical power ratio value is selected to maintain the fuel cladding

integrity safety limit (i.e., that 99.9 percent of all fuel rods in

the core be expected to avoid boiling transition).

[[Page 49932]]

Appropriate operating limit minimum critical power ratio values are

established, based on the safety limit minimum critical power ratio

value, to ensure that the fuel cladding fuel integrity safety limit

is maintained. The operating limit minimum critical power ratio

values are incorporated in the Core Operating limits Report as

required by Technical Specification 6.9.3.1. The new GE13 safety

limit minimum critical power ratio value of 1.09 is based on the

same fuel cladding integrity safety limit criteria at that for the

GE11 safety limit minimum critical power ratio value of 1.07 (i.e.,

that 99.9 percent of all fuel rods in the core be expected to avoid

boiling transition); therefore, the proposed change does not result

in a significant reduction in the margin of safety.

Proposed Change 2:

As previously stated, the purpose of the standby liquid control

is to inject a neutron absorbing solution into the reactor in the

event that a sufficient number of control rods cannot be manually

inserted to maintain subcriticality. Sufficient solution is to be

injected such that the reactor will be brought from maximum rated

power conditions to subcritical over the entire reactor temperature

range from maximum operating to cold shutdown conditions. General

Electric reactor fuel methodology establishes a fuel type dependent

standby liquid control system shutdown margin to account for

calculational uncertainties. General Electric calculations show that

an in-vessel concentration of 660 ppm will provide an estimated

standby liquid control system minimum shutdown margin of 4.1% delta

k. To achieve an in-vessel concentration of 660 ppm, the acceptable

range of standby liquid control system tank concentrations is being

revised for the lower range of tank volumes. Thus, proposed revision

of the standby liquid control system sodium pentaborate volume-

concentration range ensures that there will not be a significant

reduction in the amount of available shutdown margin and, therefore,

not a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of 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 Project Director: David B. Matthews

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

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

Date of amendment request: April 10, 1995

Description of amendment request: The requested amendment would

modify Technical Specification 4.6.4.3 to allow a reduction in the

number of hydrogen mitigation system igniters that must be maintained

Operable. This would allow removal of the hydrogen igniters in the

incore instrument tunnel.

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 requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. No impact upon accident probabilities will be created,

since the EHM System is not an accident initiating system. In

addition, it has been demonstrated that based on the results of

computer analysis, and the review of results of an external study

performed for a similar type containment, that hydrogen

concentrations in the cavity during degraded core accidents will

remain within acceptable limits. No impact on the plant response to

any accident will be created (either design basis or beyond-design

basis).

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. As stated previously, the EHM System is not an accident

initiating system. No new accident causal mechanisms will be created

as a result of deleting the affected igniters. Plant operation will

not be affected by the proposed amendments and no new failure modes

will be created.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. No adverse impact upon any plant

safety margins will be created. As shown previously, applicable

computer analysis has successfully demonstrated that the affected

igniters could be removed with no adverse consequences. No fission

product barriers are being degraded. No change to the manner in

which the units are operated is being made.

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: September 1, 1995

Description of amendment request: Generic Letter 88-16 provided

guidance on removing cycle-specific parameters which are calculated

using NRC approved methodologies from Technical Specifications (TS).

The parameters are replaced in TS with a reference to a named report

which contains the parameters, and a requirement that the parameters

remain within the limits specified in the report. The proposed changes

incorporate NRC approved methodologies, approved revisions to

previously approved methodologies, or republished versions of

previously approved methodologies into Section 6.9 of the Catawba TS.

For Catawba, the limits to which these methodologies are applied are

explicitly listed in the TS. Since the proposed changes only

incorporate NRC approved methodologies into the TS the licensee

proposed that the changes are administrative in nature and can be

assumed to have no impact, or potential impact, on the health and

safety of the public.

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 will not create a significant hazards

consideration, as defined by 10 CRF 50.92, because:

1) The proposed changes will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes are administrative in nature, and do not

affect any system, procedure, or manipulation of any equipment which

could affect the probability or consequences of any accident.

2) The proposed changes will not create the possibility of any

new or different kind of accident from any accident previously

evaluated.

The proposed changes are administrative in nature, and cannot

introduce any new failure mode or transient which could create any

accident.

[[Page 49933]]

3) The proposed changes will not involve a significant reduction

in a margin of safety.

The proposed changes are administrative in nature, and will not

affect any operating parameters or limits which could result in 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: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: September 13, 1995

Description of amendment request: The proposed amendments modify

the notation for the overpower delta-temperature (OPDT) reactor trip

heatup setpoint penalty coefficient to be consistent with NUREG-0452,

Revision 4, ``Standard Technical Specifications for Westinghouse

Pressurized Water Reactors'' (STS). This change is necessary in order

to allow implementation of the modification to reduce the reactor

coolant system hot leg temperature as planned during the Unit 2 end-of-

cycle 7 refueling outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

As required by 10CFR50.91, this analysis is provided concerning

whether the requested amendments involve significant hazards

considerations, as defined by 10CFR50.92. Standards for

determination that an amendment request involves no significant

hazards considerations are if operation of the facility in

accordance with the requested 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.

Criterion 1

The proposed amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The amendments will have no impact whatsoever upon the

probability of any accident being initiated, since the reactor trip

system is an accident mitigating system. The amendments will have no

adverse impact upon any accident consequences or upon the function

of the OPDT setpoint. The reactor trip heatup setpoint penalty will

continue to be applied anytime T-avg is greater than T [double

prime] and will not be applied when T-avg is less than or equal to T

[double prime]. This is consistent with the intent of this function.

Criterion 2

The proposed amendments will not create the possibility of a new

or different kind of accident from any accident previously

evaluated. The function of the OPDT setpoint will not be altered by

the proposed changes. As stated previously, the reactor trip system

is an accident mitigating system, so no new failure modes can be

created. No change to any aspect of plant operation will result from

NRC approval of the proposed amendments.

Criterion 3

The proposed amendments will not involve a significant reduction

in a margin of safety. The changes are necessary to allow full

implementation of the T-hot reduction modification on Catawba Unit

2. The proposed changes are consistent with the terminology of both

NUREG-0452, Revision 4 and NUREG-1431, Revision 1. OPDT setpoint

behavior will not be adversely impacted by the proposed changes;

therefore, no impact upon any plant safety margins will result.

Based upon the preceding analyses, Duke Power Company concludes

that the requested amendments do not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: March 29, 1995

Description of amendment request: The amendments would revise the

Technical Specification 3.4.9.3 requirements for the Low Temperature

Overpressure Protection (LTOP) system and update the heatup and

cooldown curves. The intent of the proposed amendments is to enhance

overpressure protection during low temperature operations. These

enhancements can be fully implemented, improving startup and shutdown

operation of McGuire Units 1 and 2.

Specifically, these changes are categorized into five groups

identified as follows:

1) Revisions to the LCO requirements, the Action Statements and the

SR for the Reactor Coolant System Overpressure Protection System during

low temperature conditions,

2) A reduction in the Reactor Coolant System (RCS) vent requirement

from 4.5 square inches to 2.75 square inches,

3) The use of the Residual Heat Removal suction relief valve (1ND3

and 2ND3) for overpressure protection under restricted conditions. (RCS

greater than 107 deg.F and cooldown rate less than 20 deg.F/hr; or RCS

greater than 167 deg.F),

4) Revisions of the Pressure/Temperature curves to 16 EFPY,

including the incorporation of the latest radiation surveillance

capsule results and removal of instrumentation margins from the

Technical Specification figures, and

5) Changes to format and consistency.

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 for each of the five groups listed above.

FIRST STANDARD

(Amendment would not) involve a significant increase in the

probablility or consequences of an accident previously evaluated.

1) Revised LCO [limiting conditions for operation] and SR

[surveillance requirements] for LTOP:

The reduced maximum setpoint will prevent the violation of the

10 CFR 50 Appendix G pressure/temperature curves (as modified by the

provisions of ASME Code Case N-514) during overpressure transients

at low temperatures. Since the maximum setpoint is reduced, the peak

pressure for LTOP [low-temperature overpressure protection] events

will be reduced as well. Accordingly, the consequences of an LTOP

event would not change as result of the proposed changes.

The analysis performed to determine the setpoint is, in

accordance with the methods used in previous evaluations, found

acceptable by the NRC. The three possible transients evaluated are;

1) a mass input from an operable safety injection pump; 2) a mass

input from an operable centrifugal charging pump; and 3) a heat

input from a 50 deg.F temperature difference between the steam

generators and the NC system. The LTOP setpoint of the PORV [power-

operated relief valve] proposed by this technical specification

change is not considered to be an initiator of any of these three

transients. As such, the probability of an accident

[[Page 49934]]

previously evaluated would not be increased as a result of the proposed

changes.

Two additional conditions for operability of the LTOP system are

defined (accumulator isolation and only one NV or NI pump operable)

and new surveillance requirements are specified as well. They

provide additional limitations, requirements and restrictions that

currently do not exist within the technical specifications for

McGuire. The incorporation of these proposed changes are consistent

with what is specified within NUREG-1341. Therefore, these changes

do not increase the probability of consequences of an accident

previously evaluated.

2) Reduction in NC vent opening:

The bases for the size of the vent to be established per the

technical specifications is to ensure that the 10 CFR 50, Appendix G

pressure/temperature limits are not exceeded during an LTOP event.

The determination of the size of the opening continues to preserve

the above design basis. The evaluation performed demonstrated that a

2.75 square inch opening would provide adequate overpressure

protection for the combined capacity of a centrifugal charging pump

and a safety injection pump.

The only time that the vent path is to be established is when

the PORVs may not be available. Defining the size of the vent is not

considered to be an initiator of any LTOP events that have been

previously evaluated. As such, this change in the size of the vent

opening does not increase the probability of an overpressure event

during low temperature conditions. The analysis performed verifies

that the size opening specified is sufficient to mitigate the

consequences of an LTOP event. Accordingly, the change in the size

of the opening for the vent will not impact the consequences of LTOP

events.

3) Use of RHR [residual heat removal] suction relief valves:

By letter dated September 11, 1990, the NRC authorized the

deletion of the RHR autoclosure interlock circuitry. A modification

which removed the RHR system suction isolation valve autoclosure

interlocks has been completed. As such, the RHR suction relief valve

can be exposed to NC system pressure and would be available to

mitigate LTOP events.

The proposed amendments specify the necessary requirements and

controls to ensure proper ND system alignments and conditions will

exist to protect the pressure/temperature limits. This added

relieving capacity will enhance the current LTOP system at McGuire

in mitigating overpressure events at low temperatures. As such, the

mitigation of previously evaluated LTOP events would be improved by

the proposed technical specification changes. Further, the proposed

changes would not esult in the initiation of an LTOP event or cause

an overpressure transient. Accordingly, the proposed amendment would

not involve an increase in the consequences or the probability of an

accident previously evaluated.

4) Revised pressure/temperature curves to 16 EFPY [effective

full-power year]:

The proposed pressure/temperature curves, provided by this

amendment request, satisfy all regulatory required material

embrittlement considerations including: ASME Section XI Appendix G,

10 CFR 50 Appendix G, and Regulatory Guide 1.99, Revision 2. In

addition, the margins for instrument error have been removed from

the curves. Instrument error will be administratively handled by

incorporating them into the LTOP system setpoint selection

calculations and into appropriate controlling procedures for unit

operations.

The proposed changes to the pressure/temperature curves are not

considered to be an initiator of LTOP events. The changes to the

curves proposed by this amendment request will not cause an LTOP

event. The curves define the new limits that have been defined in

accordance with regulatory requirements by which both units are to

be operated within. Accordingly, the proposed amendment will not

increase the probability or the consequences of previously evaluated

accidents.

5) Format and consistency:

The changes associated within this group are considered to be

administrative in nature. They do not affect station operability or

require any modifications to the facility. Accordingly, the proposed

amendment request does not increase the probability or consequences

of any previously evaluated accident.

SECOND STANDARD

(Amendment would not) create the possibility of a new or

different kind of accident from any kind of accident previously

evaluated.

1) Revised LCO and SR for LTOP:

The only potential impact to plant systems, structures and

components, as a result of the proposed changes associated with this

group, would be the setting of the PORV low pressure setpoint. No

other changes to plant systems, structures or components would

occur. The proposed amendments, also, would not impact the plant

operation. Although the value for the PORV pressure setting

specified within the technical specification would be reduced per

the proposed amendment, the actual settings of the PORV are now

currently set low enough to comply with the proposed lower setpoint

value. As such, the proposed lower setpoint would not require any

changes to the plant nor how the plant is operated.

The additional requirements for LTOP operability will not

require any modifications to the plant nor how the plant is

operated. Currently, when entering LTOP conditions, the accumulators

are isolated and only one NV or NI pump is capable of injecting into

the reactor vessel. these actions are currently controlled and are

specified within the operating procedures for heatup and cooldown of

the respective units. The proposed changes will now specify these

current operating requirements within the technical specifications

as well.

Accordingly, the proposed revisions will not create a new or

different kind of accident than what has already been previously

evaluated.

2) Reduction in NC vent opening:

The proposed changes to the technical specifications associated

with this group involves the size of the vent opening. The proposed

amendment reduces the size of the vent opening from 4.5 square

inches to 2.75 square inches. The analysis that was performed has

determined that the proposed size for the vent opening is adequate

for overpressure events. Therefore, this proposed revision to the

technical specifications will not result in a new or different kind

of accident from any kind of accident previously evaluated.

3) Use of RHR suction relief valves;

The proposed amendment associated with this group will specify

the necessary requirements and controls to ensure the appropriate

use of the RHR suction relief valve for overpressure protection.

This added relieving capacity will enhance the current LTOP system

in mitigating overpressure events during low temperature conditions.

The analysis that has been performed demonstrates the adequacy of

the RHR suction relief valve, in conjunction with a PORV, in

mitigating overpressure events at low temperatures, assuming a worst

case single failure as well. As such, the use of the RHR suction

relief valve in the manner prescribed by the proposed technical

specification amendment will not create a new or different kind of

accident from those accidents that have been previously evaluated.

4) Revised pressure/temperature curves to 16 EFPY:

The changes associated with this group, provide new heatup and

cooldown curves for both Units 1 and 2, which will extend the

service period from 10 EFPY to 16 EFPY and will remove the

instrument error as well. The proposed [heatup] and cooldown curves

were developed in accordance with all regulatory required material

embrittlement criteria. Thus, operation of the units in accordance

with the proposed new pressure/temperature curves will not create

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

accident[s] that have been previously evaluated.

5) Format and consistency:

The changes associated within this group are considered to be

administrative in nature. They do not affect station operability or

require any modifications to the facility. Accordingly, the proposed

amendment will create the possibility of a new or different kind of

accident from that previously evaluated.

THIRD STANDARD

(Amendment would not) involve a significant reduction in a

margin of safety.

1) Revised LCO and SR for LTOP:

This proposed change will reduce the maximum PORV setpoint such

that, for LTOP events, the maximum pressure in the vessel would not

exceed 110% of the pressure/temperature limits that have been

established in accordance with ASME Appendix G. This is congruous

with the provisions of ASME Code Case N-514. Currently, the maximum

PORV setpoint for LTOP events ensure that the maximum pressure would

not exceed 100% of the pressure/temperature curves. As such, the

proposed change appears to involve a slight reduction in a margin of

safety.

Although the proposed change may involve a slight reduction in a

margin of safety, the proposed change will provide an

[[Page 49935]]

equivalent margins of safety to the reactor vessel during LTOP

transients and will satisfy the underlying purpose of 10 CFR 50.60

for fracture toughness requirements. By letter dated June 28, 1994,

an exemption request and authorization to use ASME Code Case N-514

at McGuire was submitted to the NRC for review and approval.

Approval for the use of the code case was granted on September 30,

1994. The proposed change to reduce the maximum PORV setpoint,

coupled with the September 30, 1994 NRC approval for the use of Code

Case N-514 satisfies current regulatory acceptance criteria.

Therefore, the proposed change would not involve a significant

reduction in a margin of safety.

This change group, also, defines two additional conditions for

the operability of the LTOP system (accumulator isolation and only

one NV or NI pump operable) and proposes new surveillance

requirements and restrictions that currently do not exist within the

technical specifications for McGuire. The incorporation of these

proposed changes are consistent with what is specified within NUREG-

1341. Therefore, these changes do not involve a significant

reduction in a margin of safety.

2) Reduction in NC vent opening:

The proposed changes to the technical specifications associated

with this group involves the size of the vent opening. The proposed

amendment reduces the size of the vent opening from 4.5 square

inches to 2.75 square inches. The basis for the size of the vent to

be established per the technical specifications is to ensure that

the 10 CFR 50, Appendix G pressure/temperature limits are not

exceeded during an LTOP event. The determination of the size of the

opening continues to preserve the above design basis. The evaluation

performed demonstrated that a 2.75 square inch opening would provide

adequate overpressure protection for the combined capacity of a

centrifugal charging pump and a safety injection pump. Accordingly,

the proposed changes would not involve a significant reduction in a

margin of safety.

3) Use of RHR suction relief valves:

The proposed amendment associated with this group will specify

the necessary requirements and controls to ensure the appropriate

use of the RHR suction relief valves for overpressure protection.

This added relieving capacity will enhance the current LTOP system

in mitigating overpressure events during low temperature conditions.

The analysis that has been performed demonstrates the adequacy of

the RHR suction relief valve, in conjunction with a PORV, in

mitigating overpressure events at low temperatures.

Further, by letter dated September 11, 1990, the NRC approved

amendments to delete a portion of the surveillance requirements

regarding periodic verification that the RHR suction isolation

valves automatically close on a RCS [reactor coolant system] signal

less than or equal to 560 psig. This action, in effect, authorizes

the removal of the RHR autoclosure interlock circuitry. As discussed

within the NRC Safety evaluation for the amendment, the Commission

and industry have recognized the safety benefits of removing the ACI

[automatic closure and interlock] circuitry from the RHR system to

minimize, and thus reduce the risk associated with loss of decay

heat removal events.

Therefore, the proposed amendments associated with this change

group will not involve a significant reduction in a margin of

safety.

4) Revised pressure/temperature curves to 16 EFPY:

The changes associated with this group provide new heatup and

cooldown curves for both Units 1 and 2, which will extend the

service period from 10 EFPY to 16 EFPY and will relocate the

instrument error as well. The proposed pressure/temperature curves

provided by this amendment request satisfy all regulatory required

material embrittlement considerations including; ASME Section XI

Appendix G, 10 CFR 50 Appendix G, and Regulatory Guide 1.99,

Revision 2. The instrument error will be administratively handled by

incorporating them into the LTOP system setpoint selection

calculations and into the controlling procedures for unit

operations.

The relocation of the instrument error to licensee controlled

documents is consistent with the NRC actions proposed within NUREG-

1431, new standard technical specifications for Westinghouse plants.

As prescribed within NUREG-1431, the pressure/temperature limit

curves are to be relocated to a licensee controlled document

entitled ``Pressure Temperature Limit Report (PTLR)''. Changes to

the heatup and cooldown curves would then be performed in accordance

with 10 CFR 50.59 criteria. For the situation proposed by this

amendment, updates and revisions of the instrument error associated

with the pressure/temperature limit curves will be processed in a

similar fashion. Thus, the proposed change to relocate the

instrument error to licensee controlled documents is analogous with

NRC acceptable practices.

Accordingly, the proposed changes will not reduce a margin of

safety.

5) Format and consistency:

The changes associated within this group are considered to be

administrative in nature. They do not affect station operability or

require any modifications to the facility. Accordingly, there is no

reduction in the margin of safety of the LTOP system due to the

incorporation of these editorial/administrative 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: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: June 21, 1995

Description of amendment request: The proposed amendments will

revise the action statements for a single inoperable Emergency Diesel

Generator (EDG), TS 3.8.1.1.b, to extend the allowed outage time (AOT)

from 72 hours to 7 days, and permit a 10 day AOT to be used once per

refueling cycle. This proposal is a result of a cooperative study by

participating Combustion Engineering Owners Group members which

concluded that the proposed AOT extension improves plant operational

flexibility while adequately controlling overall plant risk.

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments for St. Lucie Unit 1 and Unit 2 will

extend the action completion/allowed outage time (AOT) for a single

inoperable Emergency Diesel Generator (EDG) from 72 hours to 7 days,

with provisions for a 10 day AOT once per refueling cycle. The EDGs

are designed as backup AC power sources for essential safety systems

in the event of a loss of offsite power. As such, the EDGs are not

accident initiators, and an extended AOT to restore operability of

an inoperable diesel generator would not increase the probability of

occurrence of accidents previously analyzed.

The proposed technical specification revisions involve the AOT

for a single inoperable EDG, and do not change the conditions,

operating configuration, or minimum amount of operating equipment

assumed in the plant safety analyses for accident mitigation. In

addition, a Probability Safety Assessment (PSA) was performed to

quantitatively assess the risk impact of the proposed amendment. The

impact on the early radiological release probability for design

basis events was also evaluated. It was concluded that the risk

contribution from this proposed AOT is very small, and that the

impact will be negligible.

Therefore, operation of either facility in accordance with its

proposed amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

amendment would not

[[Page 49936]]

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

accident previously evaluated.

The proposed amendments will not change the physical plant or

the modes of plant operation defined in either Facility License. The

changes do not involve the addition or modification of equipment,

nor do they alter the design of plant systems. Therefore, operation

of either facility in accordance with its proposed amendment would

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

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendments are designed to improve EDG reliability

by providing flexibility in the scheduling and performance of

preventive and corrective maintenance activities. The surveillance

intervals or the operability requirements are not changed by the

proposal; only the AOT for a single inoperable EDG will be extended.

The proposed changes do not alter the basis for any technical

specification that is related to the establishment of, or the

maintenance of, a nuclear safety margin. Moreover, an integrated

assessment of the risk impact of extending the AOT for a single

inoperable EDG has determined that the risk contribution is very

small and can be offset by improvements in EDG reliability.

Therefore, operation of either facility in accordance with its

proposed amendment would 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B. Matthews

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: June 21, 1995

Description of amendment request: The proposed amendments will

revise TS 3.5.2 to allow up to 7 days to restore an inoperable Low

Pressure Safety Injection train to operable status. This proposal is a

result of a cooperative study by participating Combustion Engineering

Owners Group members which concluded that an extension of the allowed

outage time (AOT) from 72 hours to 7 days can improve plant operational

flexibility and is risk beneficial.

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments for St. Lucie Unit 1 and Unit 2 will

extend the action completion/allowed outage time (AOT) for a single

inoperable Low Pressure Safety Injection (LPSI) train from 72 hours

to 7 days. A LPSI train is designed as a part of each Emergency Core

Cooling System (ECCS) subsystem to supplement Safety Injection Tank

(SIT) inventory during the early stages of mitigating a Design Basis

Accident. As such, components of the LPSI system are not accident

initiators, and an extended AOT to restore operability of an

inoperable LPSI train would not increase the probability of

occurrence of accidents previously analyzed.

The safety analyses for both St. Lucie Units demonstrate that

ECCS performance acceptance criteria are satisfied with only one of

the two redundant ECCS subsystems operating during the postulated

Design Basis Accident. The proposed technical specification

revisions involve the AOT for a single inoperable LPSI train, and do

not change the conditions assumed for the minimum amount of

operating equipment needed for accident mitigation. Therefore, the

consequences of an accident previously evaluated will not be

significantly increased.

In addition to the preceding evaluation, a Probabilistic Safety

Analysis (PSA) was performed to quantitatively assess the risk

impact of the proposed amendments. It was concluded from the results

of that assessment that the risk contribution of the AOT extension

is very small, and that the net impact of the proposed amendment can

be risk beneficial.

Therefore, operation of either facility in accordance with its

proposed amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendments will not change the physical plant or

the modes of plant operation defined in either Facility License. The

changes do not involve the addition or modification of equipment nor

do they alter the design of plant systems. Therefore, operation of

either facility in accordance with its proposed amendment would 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

amendment would not involve a significant reduction in a margin of

safety.

The margin of safety associated with the ECCS system is

established by acceptance criteria for system performance defined in

10 CFR 50.46. The proposed amendments will not change this

acceptance criteria nor the operability requirements for equipment

that is used to achieve such performance as demonstrated in the

plant safety analyses. Moreover, an integrated assessment of the

risk impact of extending the AOT for a single inoperable LPSI train

has concluded that the risk contribution is very small, LPSI system

reliability can potentially be improved, and the net impact of the

proposed change can be risk beneficial. Therefore, operation of

either facility in accordance with its proposed amendment would 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B. Matthews

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: June 21, 1995

Description of amendment request: The proposed amendments will

revise the action statements and certain surveillances of TS 3/4.5.1,

Safety Injection Tanks (SIT). This proposal is based on the results of

a cooperative study performed by participating Combustion Engineering

Owners Group members which investigated the impact of a risk-based

allowed outage time (AOT) extension, and also included recommendations

for line-item TS improvements from NUREG-1366 and Generic Letter 93-05.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The license amendments proposed for St. Lucie Units 1 and 2

incorporate certain line-

[[Page 49937]]

item Technical Specifications (TS) improvements for the Safety

Injection Tanks (SIT), and include an extension of the required

action completion/allowed outage time (AOT) from one hour to 72

hours to restore an inoperable SIT (that is still able to perform

its safety function) to operable status. In addition, an AOT of 24

hours, based on risk assessment techniques, is proposed for an SIT

that may be unable to perform its design function.

The SITs are passive components of the Emergency Core Cooling

System (ECCS). As such, they are not accident initiators for any

transient evaluated in the plant safety analyses, and an extension

of the AOTs for restoring an inoperable SIT to operable status would

not increase the probability of occurrence of accidents previously

analyzed.

The SITs, in combination with other ECCS components, are used to

mitigate the consequences of a loss of coolant accident. The TS

revisions will provide a longer AOT for a single inoperable SIT, but

do not involve a change to the ECCS configuration or method of

operation. The proposed amendments will not change the conditions

assumed for the minimum amount of operating equipment needed for

accident mitigation. Therefore, the consequences of an accident

previously evaluated will not be significantly increased.

In addition to the preceding evaluation, a Probability Safety

Assessment (PSA) was performed to quantitatively assess the risk

impact of the 24 hour AOT proposal. The impact on the early

radiological release probability for design basis events was also

evaluated. It was concluded that the risk contribution from this AOT

is very small, and that the impact is negligible.

Therefore, operation of either facility in accordance with its

proposed amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendments will not change the physical plant or

the modes of plant operation defined in either Facility License. The

changes do not involve the addition or modification of equipment,

nor do they alter the design of plant systems. Therefore, operation

of either facility in accordance with its proposed amendment would

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

amendment would not involve a significant reduction in a margin of

safety.

The margin of safety associated with the ECCS system is

established by acceptance criteria for system performance defined in

10 CFR 50.46. The proposed amendments will not change this criteria

nor the operability requirements for equipment that is used to

achieve such performance as demonstrated by the plant safety

analyses. Moreover, an integrated assessment of the risk impact of

allowing 24 hours to restore an inoperable SIT to operable status

has concluded that this impact is very small, and can be offset by

averting an unnecessary transition to the shutdown modes. Therefore,

operation of either facility in accordance with its proposed

amendment would 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B. Matthews

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: August 16, 1995

Description of amendment request: The revisions will modify

Technical Specification 3.6.6.1, Shield Building Ventilation System

(SBVS), to more effectively address the design functions performed by

the SBVS for both the Shield Building (secondary containment) and the

Fuel Handling Building.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed license amendment for St. Lucie Unit 2 will clarify

the Applicability and the Actions required by Technical

Specification (TS) 3.6.6.1, and explicitly account for the dual

purpose of the Shield Building Ventilation System (SBVS) to perform

design functions for both the Shield Building (secondary

containment) and the Fuel Handling Building. The proposed amendment

is administrative in nature.

The SBVS only operates when actuated by automatic control

signals generated by systems detecting postulated accident

conditions. The SBVS is not an accident initiator, the proposed TS

changes do not involve any assumptions relative to accident

initiators used in the plant safety analyses, and the amendment,

therefore, will not impact the probability of occurrence for

accidents previously analyzed. Relative to accident consequences, at

least one train of the SBVS must operate to fulfill the design

function of evacuating filtered air from the Shield Building during

the postulated Loss of Coolant Accident; and likewise assumed in the

analysis for the Fuel Handling Building during a fuel handling

accident. The proposed changes simply remove elements of ambiguity

from TS 3.6.6.1; do not reduce the existing operability requirements

for the system; and provide further assurance that proper

compensatory measures will be taken in the event one or both SBVS

trains become inoperable.

Therefore, operation of the facility in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment is administrative in nature and will not

change the physical plant or the modes of plant operation defined in

the facility license. The changes do not involve the addition or

modification of equipment, nor do they alter the design or methods

of operation of plant systems. Plant configurations that are

prohibited by TS will not be created by this amendment. Therefore,

operation of the facility in accordance with the proposed amendment

would 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

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendment will not change the SBVS operability

requirements nor otherwise alter the basis for any technical

specification that is related to the establishment of, or the

maintenance of, a nuclear safety margin. The proposed changes are

administrative in nature, and are designed to provide assurance that

the SBVS capability to perform design functions assumed available in

the safety analyses will remain available during the various plant

operating modes. Therefore, operation of the facility in accordance

with the proposed amendment would 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: David B. Matthews

[[Page 49938]]

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: August 16, 1995

Description of amendment request: The proposed amendments revise

St. Lucie Units 1 and 2 Technical Specifications to relocate selected

Technical Specification Monitoring Instrumentation utilizing the Final

Policy Statement on Technical Specification Improvement for Nuclear

Power Reactors, 58 FR 39132, July 22, 1993. The proposed amendments

also include relocation of Technical Specifications related to the

Emergency and Security Plan review process utilizing the guidance

contained in NRC Generic Letter 93-07, ``Modification of the Technical

Specification Administrative Requirements for Emergency and Security

Plans.''

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 Selected Technical Specification

Requirements Related to Instrumentation are administrative in nature

in that the specifications for operation and surveillance of the

selected Technical Specification instrumentation will be relocated

from Appendix A of the facility operating license to the Updated

Final Safety Analysis Report (UFSAR) for each unit. Once relocated,

future changes will be controlled by 10 CFR 50.59 and the UFSARs

updated pursuant to 10 CFR 50.71(e). Relocation of these

requirements to the UFSAR is consistent with the NRC ``Final Policy

Statement on Technical Specifications Improvements for Nuclear Power

Reactors'' published in the Federal Register (58 FR 39132) dated

July 22, 1993.

The selected Technical Specification instruments are not

accident initiators nor a part of the success path(s) which function

to mitigate accidents evaluated in the plant safety analyses. The

proposed Technical Specification change does not involve any change

to the configuration or method of operation of any plant equipment

that is used to mitigate the consequences of an accident, nor do the

changes alter any assumptions or conditions in any of the plant

accident analyses. Therefore, operation of the facility in

accordance with the proposed amendment would not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The Technical Specifications changes associated with Emergency

Plan and Security Plan requirements are proposed in accordance with

Generic Letter 93 07. The changes being proposed are administrative

in nature and do not affect assumptions contained in plant safety

analyses, the physical design and/or operation 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 affect the probability or

consequences of an accident previously analyzed.

(2) Use of the modified specification would not create the

possibility of a new or different kind of accident from any

previously evaluated.

The proposed amendment to relocate the existing Technical

Specification requirements for selected Technical Specification

instrumentation to the UFSAR will not change the physical plant or

the modes of plant operation defined in the Facility License. The

change does not involve the addition or modification of equipment

nor does it alter the design or operation of plant systems.

Therefore, operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendments, in accordance with Generic Letter 93-

07, change the Technical Specifications to remove the audit of the

emergency and security plans and implementing procedures from the

list of responsibilities of the Facility Review Group. The changes

being proposed are administrative in nature and will not change the

physical plant or the modes of operation defined in the Facility

License. The change does not involve the addition or modification of

equipment nor does it alter the design or operation of plant

systems. Therefore, 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.

(3) Use of the modified specification would not involve a

significant reduction in a margin of safety.

The proposed changes are administrative in nature in that

operating and surveillance requirements for the selected Technical

Specification instrumentation will be relocated from Appendix A of

the facility license to the appropriate Updated Final Safety

Analysis Report for each unit. These selected instruments are not

used to actuate safety-related equipment, provide interlocks, or

otherwise perform plant control functions. Conditions evaluated in

plant accident and transient analyses do not involve these selected

instruments. The proposed changes do not alter the basis for any

technical specification that is related to the establishment of, or

the maintenance of, a nuclear safety margin. Therefore, operation of

the facility in accordance with the proposed amendment would not

involve a significant reduction in a margin of safety.

The proposed amendments, in accordance with Generic Letter 93-

07, change the Technical Specifications to remove the audit of the

emergency and security plans and implementing procedures from the

list of responsibilities of the Facility Review Group. The changes

being proposed are administrative in nature and 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 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800

M Street, N.W., Washington, DC 20036

NRC Project Director: David B. Matthews

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

Linn County, Iowa

Date of amendment request: July 21, 1995

Description of amendment request: The proposed amendment would make

administrative changes to various sections of the Duane Arnold Energy

Center (DAEC) Technical Specifications (TS). These changes replace a

conditional surveillance if one emergency service water (ESW) pump or

loop is determined to be inoperable (TS 4.8.E.2); credit successful

emergency diesel generator (EDG) tests performed in the previous 24

hours (TS 4.8.E.2); clarify the requirements governing spent and new

fuel storage in Section 5.5 of the DAEC TS; and eliminate the

Operations Committee reviews of procedures in support of the DAEC

Emergency Plan and Security Plan, as specified in Sections 6.5 and 6.8

of the TS. DAEC TS Section 4.8.E.2 states the surveillance requirement

applicable when one ESW pump or loop is determined to be inoperable.

This amendment request deletes the surveillance requirement to

physically test the opposite train's EDG and replaces it with a

requirement to verify OPERABILITY of the opposite train low pressure

core and containment cooling systems and EDG.

Basis for proposed no significant hazards consideration

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

[[Page 49939]]

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. The proposed revision does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The changes are administrative in nature and are

consistent with previously-published NRC guidance. The proposed

revision does not change any accident analysis, plant safety

analysis or calculations; degrade existing plant programs; or modify

any functions of safety related systems or accident mitigation

functions for which the DAEC has previously been credited. The

proposed revision to the Surveillance Requirements will continue to

assure OPERABILITY as required, but eliminate unnecessary operation

of an EDG.

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

new or different kind of accident from any accident previously

evaluated. The proposed revision does not alter any plant

parameters, revise any safety limit setpoint, or provide any new

release pathways. In addition, the proposed revision does not modify

the operation or function of any safety-related equipment, nor

introduce any new modes of operation, failure modes, or physical

changes to the plant.

3. The proposed revision does not involve a significant

reduction in a margin of safety. The proposed revision does not

alter any plant parameters, revise any safety limit setpoint, or

provide any new release pathways. In addition, the proposed revision

does not modify the operation or function of any safety-related

equipment, nor introduce any new modes of operation, failure modes,

or physical changes to the plant.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis

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

NRC Project Director: Gail H. Marcus

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: May 5, 1995, as revised by letter dated

July 14, 1995

Description of amendment request: The proposed changes would amend

the Cooper Nuclear Station (CNS) Technical Specifications (TS) sections

3/4.5.F.1, 3.5.F.2, 3.9.B.1, 3.9.B.2, 4.9.A.2, and the associated

bases. These changes would revise the TS to: 1) verify that the

redundant diesel generator is operable upon the loss of one diesel

generator, and implement provisions to verify that the operable diesel

generator does not have a common cause failure; 2) incorporate

provisions to allow a modified start for the diesel generators; and 3)

remove the requirement that the reactor power level be reduced to 25%

of rated power upon loss of both diesel generator units or both

incoming power sources (start-up and emergency transformers). In

addition, the period of time allowed for continued reactor operation

with both diesels inoperable would be reduced from 24 to two hours.

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:

10 CFR 50.91(a)(1), requires that licensee requests for

operating license amendments be accompanied by an evaluation of

significant hazards posed by the issuance of the amendment. NPPD has

reviewed the proposed changes in accordance with 10CFR50.92 and

concludes that the changes do not involve a significant hazards

consideration (SHC). The basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not compromised. The proposed changes

do not involve a SHC because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

Proposed Revision 1:

This proposed revision serves to ensure that an emergency diesel

generator is always available to perform on demand and that lowering

the number of demands to demonstrate operability reduces the

probability of equipment failure. The required action no longer

requires the redundant emergency diesel generator to be demonstrated

operable immediately. Therefore, this requirement has been deleted

from TS 4.5.F.1.

The proposed change includes provisions to determine if the

redundant diesel generator has been made inoperable by a common

cause failure or perform a demonstration test. The redundant

emergency diesel generator will remain in service during the entire

period of inoperability of the out of service emergency diesel

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

redundant diesel generator will be tested in accordance with the

surveillance requirements of TS section 4.9.A.2.a.1 to assure

operability.

Since this proposed revision does not affect the design or

negatively affect the performance of the diesel generators, the

change will not result in an increase in the consequences or

probability of an accident previously analyzed. This proposed

revision will increase diesel generator reliability and

availability, thereby increasing overall plant safety.

Proposed Revision 2:

This proposed revision only affects emergency diesel generator

periodic testing. The diesel generators are not accident initiators

and the method of testing the diesel generators cannot initiate an

accident and therefore will not increase the probability of an

accident. This change to the diesel generator testing method does

not impact any Updated Safety Analysis Report (USAR) safety

analysis. The proposed surveillances will still provide assurance

that the diesel generators are available to mitigate the

consequences of accidents previously evaluated. Thus the

consequences of an accident previously evaluated are not increased.

The revised periodic testing will still demonstrate that the

emergency diesel generators are ready to perform their safety

function. An overall improvement in diesel engine reliability and

availability can be gained by performing diesel generator starts for

surveillance testing using engine prelubes, warmups and other

manufacturer recommended practices to reduce engine stress and wear.

Since this proposed revision does not affect the design or

negatively affect the performance of the diesel generators, the

change will not result in an increase in the consequences or

probability of an accident previously analyzed. This proposed

revision will increase diesel generator reliability, thereby

increasing overall plant safety.

Proposed Revision 3:

This proposed revision does not affect the operation of the

emergency diesel generators or the incoming power sources (start-up

and emergency transformers). Both the diesel generators and the

incoming power sources function to mitigate the consequences of

postulated accidents. As such, removing the requirement to reduce

power level upon the loss of both redundant components in either of

these systems does not create an increase in the probability of an

accident. By eliminating this requirement, the potential for plant

transients during power reduction to 25% are also eliminated.

Eliminating this requirement will not increase the consequences of a

postulated accident because the redundant components will remain

available. Additionally, the loss of both offsite power sources

condition becomes more restrictive by requiring a plant shutdown

instead of notification within 24 hours.

The proposed changes do not alter the conditions or assumptions

in any of the Updated Safety Analysis Report (USAR) accident

analyses. Since the USAR accident analyses remains bounding, the

radiological consequences previously evaluated are not adversely

affected by the proposed changes. Therefore, no significant increase

in the probability or consequences of an accident previously

analyzed would occur.

The proposed rearrangement of information, and rewording of some

the TS requirements are included to enhance usability and alleviate

any possible confusion. These changes are strictly editorial have no

impact, and do not alter technical content or meaning of the

specifications. These editorial changes do not involve a significant

increase in the probability or consequences of an accident

previously analyzed.

[[Page 49940]]

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

from any previously analyzed.

Proposed Revision 1:

Accidents involving loss of off-site power and single failure

have been previously evaluated, and this proposed change does not

impact any of those assumptions. This proposed revision does not

introduce any new mode of plant operation or new accident

precursors, involve any physical alterations to plant

configurations, or make changes to system setpoints which could

initiate a new or different kind of accident. Operation of the

facility in accordance with the proposed revised changes does not

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

any previously evaluated.

Proposed Revision 2:

This proposed revision only affects emergency diesel generator

periodic testing. The diesel generators are not accident initiators

and the method of testing the diesel generators cannot initiate an

accident. This revision does not relieve the operation of the diesel

generator from existing requirements and the diesel generators

remain bounded by the assumptions in the USAR accident analysis. The

method of testing provides assurance that the diesel generators are

available when needed. The proposed revision does not involve any

changes in setpoints, plant equipment, plant operation, protective

functions, or the design basis of the plant. Therefore, a change in

the method of starting the diesel generators during periodic testing

would not create a different kind of accident than previously

evaluated.

Proposed Revision 3:

This proposed revision does not add or change any equipment or

logic, nor do the changes associated with this revision alter any

system operability requirements. The proposed changes for this

revision do not introduce any new failure modes for any plant system

or component important to safety nor has any new limiting failure

been identified as a result of the proposed revision. Since there

are no changes to the function, or operation of any system,

equipment, or component, the possibility of a new or different kind

of accident is not created.

The proposed rearrangement of information, and rewording of some

[of] the TS requirements are included to enhance usability and

alleviate any possible confusion. These changes are strictly

editorial have no impact, and do not alter technical content or

meaning of the specifications. These editorial changes do not create

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

previously analyzed.

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

Proposed Revision 1:

This proposed revision does not result in an overall reduction

in the margin of safety. The reduction in margin going from

``immediately'' testing an operable diesel generator to 24 hours to

determine no common cause, is offset by the increase in margin

resulting from increased diesel generator reliability and

availability associated with implementing the vendor recommendations

for testing and not exposing the diesel generator to potential grid

disturbances when a diesel generator is found to be inoperable. No

physical modification to the plant or change in the procedurally

prescribed operator actions result from the proposed changes

associated with this revision. Operation of the facility in

accordance with the proposed revision does not involve a significant

reduction in a margin of safety.

Proposed Revision 2:

This proposed revision is made to increase the reliability and

availability of the emergency diesel generators thus enhancing the

safety of the plant. Changing the way periodic testing of the diesel

generators is conducted does not involve a reduction in safety. The

test still demonstrates the ability of the diesel generator to start

within the time required, and reach rated voltage and frequency as

required in the accident analysis. The test also demonstrates the

ability of the diesel generator to start reliably, carry the

required load, and ensures the capabilities of the cooling system

and other support systems are operable. Therefore, assurance that

the diesel generators operate within the limits determined to be

acceptable continues to be provided. Implementing manufacturer's

recommendations to minimize stress and wear of the diesel engine

does not involve a significant reduction in the margin of safety,

but rather enhances safety.

Proposed Revision 3:

This proposed revision deletes the requirement to reduce reactor

power level to 25% of rated power upon the loss of either both

diesel generators or both incoming power sources. The elimination of

this requirement will allow the plant to maintain the existing power

level rather than subject the plant to an unnecessary transient.

Maintaining the plant at the existing power level provides a more

stable operating environment. The equipment and components of the

diesel generators or the incoming power sources are not impacted in

any way as a result of the proposed revisions. The margin of safety

for the diesel generators and the incoming power sources are not

significantly reduced since these systems are not altered in any

way, and will continue to be surveillance tested as required.

Assurance of operability is provided by the normal, scheduled

surveillances which have been established at a sufficient interval

to provide reasonable assurance of operability. Therefore, the

proposed changes do not involve a significant reduction in the

margin of safety.

The proposed rearrangement of information, and rewording of some

[of] the TS requirements are included to enhance usability and

alleviate any possible confusion. These changes are strictly

editorial have no impact, and do not alter technical content or

meaning of the specifications. These editorial changes do not

involve a significant reduction in the margin of safety.

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

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

satisfied. The licensee's July 14, 1995, letter revised the proposed

changes in their letter of May 5, 1995, to further limit the period of

time that continued reactor operation would be allowed with both

emergency diesel generators inoperable from 24 to two hours. This

revision to the proposed changes is more restrictive and does not

impact the licensee's analysis of the criteria of 10 CFR 50.92(c).

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, Nebraska 68602-0499

NRC Project Director: William D. Beckner

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,

Millstone Nuclear Power Station, Unit 1, New London County,

Connecticut

Date of amendment request: August 31, 1995

Description of amendment request: The proposed amendment modifies

the definition of HOT SHUTDOWN and COLD SHUTDOWN to specify that the

definitions are not applicable during the performance of an inservice

hydrostatic and leak test (IHLT). Technical Specification Section 3.6.B

and 4.6.B would be modified by adding Section 3.6.B.1.b and 4.6.B.1.b

to identify the requirements that must be satisfied to consider the

reactor in COLD SHUTDOWN during the performance of an IHLT. In

addition, the proposed amendment will change temperature specific

requirements on several pages to mode or condition specific

requirements; make several editorial changes; and change the associated

Bases.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

NNECO has reviewed the proposed changes in accordance with

10CFR50.92 and concluded that the changes do not involve a

significant hazards consideration (SHC). The bases for this

conclusion is that the three criteria of 10CFR50.92(c) are not

compromised. The proposed changes do not involve an SHC because the

changes would not:

[[Page 49941]]

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed changes will allow the reactor to be considered in

COLD SHUTDOWN during an IHLT with the average reactor coolant

temperature greater than 212 deg.F but less than 280 deg.F. The

change to allow the reactor to be in COLD SHUTDOWN during the

performance of IHLT will not increase the probability or

consequences of an accident. The probability of a leak in the

reactor pressure boundary during this testing is not increased by

considering the reactor to be in COLD SHUTDOWN. The IHLT is

performed near water solid, all control rods inserted, and with an

appropriate availability of engineering safety features. The stored

energy in the reactor core will be very low and the potential for

failed fuel and a subsequent increase in coolant activity are

minimal. In addition, secondary containment will be operable and

capable of handling airborne radioactivity from leaks that could

occur during the performance of an IHLT. Requiring secondary

containment to be operable will further ensure that potential

airborne radiation from leaks will be filtered by one or both trains

of SBGT [standby gas treatment], thereby limiting releases to the

environment. Therefore, the changes will not significantly increase

the consequences of an accident.

In the unlikely event of a large pressure boundary leak, the

reactor vessel would rapidly depressurize, allowing one or both of

the operable core spray systems to operate. Small system leaks would

be detected by leakage inspections before significant inventory loss

occurred, since leakage inspections are an integral part of the IHLT

program.

Based upon 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.

The IHLT conditions remain unchanged. The potential for a system

leak remains unchanged since the reactor coolant system is designed

for temperatures exceeding 500 deg.F with similar pressures. The

change in operable engineered safety features available to mitigate

a postulated accident does not reduce the ability to

safely mitigate a postulated accident. Adequate ECCS [emergency

core cooling system] equipment will be available to mitigate a LOCA

[loss of coolant accident] with an assumed single failure.

Therefore, this will not create the possibility of a new or

different kind of accident from any previously evaluated.

3. Will not involve a significant reduction in a margin of

safety.

The proposed changes will not have any significant impact on any

design basis accident or safety limit. The various engineered safety

features which are required by the proposed change will ensure

appropriate mitigation of postulated events. Since the test is

performed at a near water solid condition and at low decay heat

values, no fuel damage is expected in case of an accident such as a

LOCA. Nevertheless, secondary containment and the SBGT system will

be maintained operable to process air-borne radioactivity from a

steam leak that could occur during the performance of the IHLT.

Therefore, the proposed change does not constitute 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 Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket No. 50-336,

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

Connecticut

Date of amendment request: August 31, 1995

Description of amendment request: The proposed change to the

Millstone 2 Technical Specifications would remove the phrase ``other

than Millstone Unit No. 2'' from Section 6.3.1 on page 6-2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed change in accordance with

10CFR50.92 and concluded that the change does not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

compromised. The proposed change does not involve an SHC because the

change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed change does not affect any system or equipment of

Millstone Unit No. 2. The proposed change does not affect the

qualification of any of the licensed individuals involved in the

day-to-day operation of Millstone Unit No. 2. The proposed change

corrects a statement which could be interpreted such that an

individual who once held a Millstone Unit No. 2 SRO [Senior Reactor

Operator] license would not be eligible to be Operations Manager.

Since this change does not affect any equipment or operating

procedures, does not affect the level of expertise and

training required for on-shift personnel, and does not reduce

the level of expertise required of operations management, this

change 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.

This change does not affect any equipment or operating

procedures, does not affect the level of expertise and training

required for on-shift personnel, and does not reduce the level of

expertise required of operations management. Therefore, this change

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

accident.

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

This change eliminates a phrase which could be interpreted to

prevent an individual who had possessed a Millstone Unit No. 2 SRO

license from becoming the Operations Manager. The training and

experience necessary to possess a Millstone Unit No. 2 SRO license

is equivalent to that of other PWRs. Therefore, this proposed change

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

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

this review, it appears that the 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 Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania

Date of amendment request: July 28, 1995

Description of amendment request: The proposed amendment would

eliminate the Technical Specifications requirements to perform 10 CFR

50, Appendix J, Type C hydrostatic testing on certain valves that are

within closed systems and are assured a water seal following a Design

Basis Accident.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed TS changes do not involve a significant increase

in the probability or

[[Page 49942]]

consequences of an accident previously evaluated.

The primary containment (drywell and suppression pool) and the

affected closed systems are accident mitigators not accident

initiators. The proposed change to the scope of Appendix J, Type C

testing does not affect the probability of the DBA [Design Basis

Accident]. The valves will continue to be maintained in an operable

state, and in their current design configuration. There is no

correlation between the scope of Appendix J, Type C testing and

accident probability. There are no physical or operational changes

to the containment structure, system or components being made as a

result of the proposed changes. Therefore, the consequences of a

malfunction of equipment important to safety is not increased from

those previously evaluated.

The consequences of loss-of-coolant accidents (LOCAs) under the

proposed change were considered where a single active failure of a

containment isolation valve (CIV) or a passive failure of the closed

system were reviewed, within the limits of the existing licensing

basis. Under the existing licensing basis, a pipe rupture of the

seismically qualified ECCS piping does not have to be assumed

concurrent with the LOCA, except if it is a consequence of the LOCA.

Consideration of consequential failures can be eliminated, since a

LOCA inside containment is separated from the affected piping by the

containment structure. Consideration of consequential failures of

the ECCS piping from LOCAs outside containment are outside the

Appendix J design considerations. A single active failure of the

CIV, under the LOCA condition, can be accommodated since the closed

and water sealed system piping remains as the leakage barrier. The

ECCS passive failure criterion does require consideration of system

leaks, but not pipe breaks, beyond the initiating LOCA. The

capability to make-up water inventory to the suppression pool is

adequate to ensure that postulated seat leakage and pipe leakage

does not result in a condition that jeopardizes pool level. Make-up

capability exists for the suppression pool via the Condensate

Storage Tank and Ultimate Heat Sink Spray Pond. Operator actions to

make-up the suppression pool are delineated in existing Operating

Procedures.

The subject valves are single isolation valves associated with

lines that penetrate the primary containment, but are not connected

directly to the primary containment atmosphere or the reactor

coolant pressure boundary. This configuration is described in the

LGS UFSAR, Section 6.2.4.3.1.3.1, which states ``the systems which

the lines from the suppression pool connect to outside containment

are closed systems meeting the appropriate requirements of closed

systems.'' The integrity of these closed systems are also monitored

and controlled in accordance with TS Section 6.8.4.a. Any leakage

that may escape the confines of the closed system will be contained

within the Reactor Building, treated by standby gas and radwaste

systems, and, therefore, are within the existing LGS licensing

bases.

Finally, the affected penetrations will continue to be subjected

to the periodic 10 CFR 50, Appendix J, Type A test (Integrated

Containment Leakage Rate Test).

The suppression pool level is designed and operated so that

water level is maintained in accordance with current TS, and the

associated bases. The supply of water in the suppression pool is

assured for 30 days during all DBA, post-accident modes of

operation. The lowest water level which the suppression pool will

reach was analyzed, and it was determined that the affected lines

will remain below this minimum level, thereby assuring a water seal.

The valves will continue to be tested and maintained to ensure their

operability, and the closed systems' integrity will continue to be

monitored and controlled in accordance with TS 6.8.4.a and the

performance of the periodic 10 CFR 50, Appendix J, Type A test.

Therefore, the proposed changes will not increase the

probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes do not change the plant response to

accident scenarios, and do not introduce new or different scenarios.

The primary containment (drywell and suppression pool) and the

affected closed systems are accident mitigators not accident

initiators. The proposed change to the scope of Appendix J, Type C

hydrostatic testing maintains the existing barriers to primary

containment bypass leakage by the assurance that a water seal will

be maintained for 30 days during all DBA, post-accident modes of

operation. The valves will continue to be tested and maintained to

ensure their operability, and the closed systems' integrity will

continue to be monitored and controlled in accordance with TS

6.8.4.a. Therefore, the proposed changes cannot cause an accident,

and the plant response to the design basis events is unchanged,

whereby the change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The water seal provided by the assurance of a minimum

suppression pool level will prevent post-accident containment bypass

leakage. Appendix J does not require air leak testing of the valves

since the 30 day post-accident supply of water is maintained. In

addition, the closed systems' integrity is monitored and controlled

in accordance with TS 6.8.4.a. Any leakage that may escape the

confines of the closed system will be contained within the Reactor

Building, and is within the existing LGS licensing bases. Therefore,

the proposed TS changes do not involve a significant reduction in a

margin of safety.

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

this 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: July 28, 1995

Description of amendment request: The proposed amendments, which

are consistent with the Improved Standard Technical Specifications

(NUREG-1433), delete the operability and surveillance requirements

involving secondary containment differential pressure 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:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

Deleting the operability and surveillance requirements for the

secondary containment differential pressure instrumentation does not

involve any changes to the design, function, or operation of any

plant components or safety-related systems. There are no changes to

the separation, redundancy, qualification, quality assurance or fire

protection requirements for the associated components and systems,

nor are there any new failure modes created. This activity only

removes operability and surveillance requirements from the Technical

Specifications for selected plant components associated with the

secondary containment differential pressure trip functions. No

credit for operation of these trip functions is taken in any design

basis accidents valuated in the SAR [Safety Analysis Report].

Thesecomponents will be maintained in accordance with the plant

preventive maintenance program. The failure of any of these

components does not result in the occurrence of an accident.

Consequently, there is no increase in the probability of occurrence

of an accident previously evaluated in the SAR.

The Outside Atmosphere to Reactor Enclosure Delta Pressure-Low

and Outside Atmosphere To Refueling Area Delta Pressure-Low trip

functions are not symptomatic of a design basis accident. No credit

for operation of the trip functions is taken in any design basis

accidents evaluated in the SAR. Neither failure of the differential

pressure components nor failure to generate the associated trip

functions affects the consequences of an accident previously

evaluated in the SAR. The appropriate

[[Page 49943]]

accident prevention and mitigation actions are generated from other

plant parameters symptomatic of an accident. Sufficient plant

parameters symptomatic of a design basis accident are monitored to

initiate the appropriate actions as evaluated in the SAR.

Furthermore, all safety-related systems will still be able to

perform all of their design basis safety-related functions.

Consequently, there is no increase in the consequences of an

accident previously evaluated in the SAR.

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 failure of the differential pressure automatic isolation

instrumentation components does not result in the occurrence of an

accident. The failure to generate the associated trip functions does

not result in the occurrence of an accident. This activity does not

involve any changes to the design, function, or operation of any

plant components or safety-related systems. There are no changes to

the separation, redundancy, qualification, quality assurance or fire

protection requirements for the associated components and systems.

These components will be maintained in accordance with the plant

preventative maintenance program. Consequently, there is no

possibility of an accident of a different type than previously

evaluated in the SAR.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The ability of secondary containment to minimize any ground

level release of radioactive material which may result from any

accident is not affected. Surveillance and operability requirements

for secondary containment SGTS [Standby Gas Treatment System] and

RERS [Reactor Enclosure Recirculation System] are not changed by

this activity. Draw down time, leakage factors, secondary

containment system ratings, and secondary containment system

response to a LOCA [Loss of Coolant Accident] or refueling accident

are not affected by this activity. SGTS and RERS initiation will

continue to occur when plant parameters symptomatic of a LOCA or

refueling accident exceed predetermined values. There are no changes

to the inputs for the post-LOCA offsite dose analysis.

Therefore, the proposed TS changes do not involve a significant

reduction in a margin of safety.

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

this 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: July 28, 1995

Description of amendment request: The proposed amendment would

modify Technical Specifications (TS) Surveillance Requirements 4.9.1.1,

4.9.1.2, 4.9.3, 4.9.5, and 4.9.8 to delete specific requirements to

perform surveillances just prior to beginning or resuming core

alterations or control rod withdrawal associated with refueling

activities. This proposed TS change would delete the phrase ``incore

instrumentation'' from the footnote in TS Section 3/4.9.5,

``Communications.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed TS changes do not involve any physical changes to

plant systems or equipment. The proposed TS changes only delete

those Surveillance Requirements (SRs) pertaining to the performance

of tests just prior to beginning or resuming core alterations or

control rod withdrawal, and revises a footnote description to be

consistent with the current TS definition of ``Core Alteration.''

The proposed TS changes do not revise any of the other applicable

periodic SRs, or modify any procedural controls currently in place

governing fuel handling operations. The periodic surveillance test

frequencies provide adequate assurance that the equipment will

remain in an operable condition. The normal periodic surveillance

intervals bound those surveillance intervals for the tests that are

being altered by this proposed TS change. In the event that one of

the periodic surveillances has not been performed within the

specified time interval, entry into the specified condition (i.e.,

performance of core alterations, control rod withdrawal, or handling

of fuel or control rods) is not permitted as required by TS 4.0.4

until the surveillance has been satisfactorily completed.

The consequences of an accident are not increased by the

proposed TS changes, since the changes only involve revising the

frequency of conducting surveillance tests. The method of operation

or performance of plant structures, systems, or plant components are

not affected by the proposed TS changes. The proposed TS changes

will not impact the operation of any fuel handling equipment, and

therefore, the potential for a Fuel Handling Accident as described

in Section 15.7.4 of the LGS [Limerick Generating Station] Updated

Final Safety Analysis Report (UFSAR) is not increased.

In addition, any unexpected reduction of water level in the

reactor cavity or fuel pool at the start of fuel handling or control

rod handling will be immediately apparent to operators by direct

observation. Plant procedures utilized by the refueling personnel

require the suspension of core component transfers in the event of

loss of water inventory.

Therefore, the proposed TS change does 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 only involve changes to the frequency in

which the specified surveillance tests are performed. The proposed

TS changes do not revise any of the other applicable periodic SRs,

or modify any procedural controls currently in place governing fuel

handling operations. The periodic surveillance test frequencies

provide adequate assurance that the equipment will remain in

operable condition. The periodic surveillance intervals bound those

surveillance intervals for the tests that are being altered by this

proposed TS change. The refueling interlock system combined with

strict procedural controls provide multiple barriers to preclude an

inadvertent criticality.

The proposed TS changes do not involve any physical changes to

plant systems or equipment. The proposed TS changes do not alter the

configuration of the plant or the way that the plant is operated.

The associated plant equipment will continue to function as

designed. This equipment is not designed to perform any other

function than it is presently capable of, and therefore, will not

affect the operation of any other plant equipment.

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 changes do not involve a significant

reduction in a margin of safety.

The proposed TS changes do not involve any physical changes to

plant systems or equipment. The reactor will continue to be

maintained subcritical during refueling operations and reactor water

level will be maintained at the required level (i.e., above the

vessel flange). The proposed TS changes do not affect the operation

of other plant systems and equipment essential in maintaining

reactor water temperature during refueling operations, or the

capability in responding to a postulated Fuel Handling Accident.

The proposed changes do not adversely affect reliability of the

refueling interlocks or refuel platform communications equipment.

[[Page 49944]]

Since the proposed changes only impact the frequency in which certain

surveillance tests are performed, and do not change the plant

configuration or setpoints, there is substantial assurance that the

reactor will be maintained subcritical during refueling.

Therefore, the proposed TS 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 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: July 28, 1995

Description of amendment request: The proposed amendment would

revise Technical Specifications Table 4.3.1.1-1, ``Reactor Protection

System Instrumentation Surveillance Requirements'', to reflect changes

to the surveillance test frequency requirements for various Reactor

Protection System [RPS] 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:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

In all of the applicable SAR [Safety Analysis Report] evaluated

events, the IRM [Intermediate Range Monitor] and APRM [Average Range

Power Monitor] instrumentation is credited for performing a

mitigating function (i.e., initiating a scram), to terminate the

transient prior to a safety limit being exceeded. The proposed TS

changes do not alter the RPS configuration, or RPS instrumentation

setpoints, nor do they change the manner in which the IRM and APRM

instrumentation carry out the scram functions. Therefore the

consequences of any potential malfunction of equipment important to

safety will remain unchanged.

In each case where a startup surveillance test requirement is

proposed to be deleted, (i.e., IRM and APRM), the normal

surveillance test frequency specified for the required Operational

Condition remains unchanged (except for the APRM Upscale Setdown

functional test). The startup surveillance requirement is

conservatively bounded by the normal surveillance test interval

which is greater than or equal to any interval associated with the

startup surveillance requirement and ensures that the IRM and APRM

instrumentation reliability is unchanged. This is in accordance with

the Improved Standard Technical Specifications, NUREG-1433, issued

September 28, 1992.

The reliability of the APRM Upscale Setdown scram function will

not be decreased due to changing the functional test frequency from

Weekly (W), to Quarterly (Q), in Operational Conditions 2, 3, and 5

(Startup, Hot Shutdown and Refueling, respectively). Plant

operational data taken from each of the APRM calibration/functional

tests performed since August 1992 until present at LGS Units 1 and

2, shows that setpoint reliability will be maintained if the

functional test frequency is increased to quarterly, as proposed.

Presently, each time an APRM calibration/functional test is

performed, both the Upscale Setdown and the Flow Reference scram

circuits are tested. The results of the quarterly tests confirm that

the APRM Upscale Setdown function already has over 2.5 years of

performance without failure in Operational Condition 1, thus being

extremely reliable.

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 affect only the required surveillance

test intervals, not the RPS configuration or RPS instrumentation

setpoints. The proposed TS changes do not introduce a new failure

mode for the IRM or APRM instrumentation. Plant operating experience

data confirms that at LGS Units 1 and 2, the IRM and APRM

instrumentation will continue to perform their safety function as

currently designed, with the same degree of reliability.

The proposed TS changes do not alter the configuration of the

plant, nor the way the plant is operated.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The following TS Bases were reviewed for potential reduction in

the margin of safety:

B 2.2.1 Reactor Protection System Instrumentation Setpoints

B 3/4.1.4 Control Rod Program Controls

B 3/4.2 Power Distribution Limits

B 3/4.3.1 Reactor Protection System Instrumentation

B 3/4.3.6 Control Rod Block Instrumentation

The surveillance test frequency changes proposed for the RPS

instrumentation section of TS do not adversely affect the IRM or

APRM instrumentation, which will continue to perform the RPS

functions required to maintain the present margin of safety. Changes

to the IRM instrumentation startup surveillance intervals are

already bounded by the existing surveillance requirements, and are

in accordance with the Improved Standard Technical Specifications,

NUREG-1433, issued September 28, 1992. The same statement applies to

the APRM instrumentation, with respect to deletion of the startup

surveillance requirement. The change of the APRM Upscale Setdown

Channel functional test surveillance interval from Weekly to

Quarterly was evaluated to ensure that the APRM instrumentation

would perform that function, with the same degree of reliability as

presently experienced. A review of the plant operating experience

data at LGS Units 1 and 2 shows that APRM instrumentation is

extremely reliable for a quarterly surveillance test interval. The

proposed TS changes do not modify plant configuration, RPS

instrumentation setpoints, or RPS operation. The margin of safety

remains unchanged.

Therefore, the proposed TS 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 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 1, 1995

Description of amendment request: The proposed amendment would

modify Technical Specifications Section 3/4.9.1, ``Reactor Mode

Switch,'' in order to provide alternate actions to allow the

continuation of core alterations in the event certain Reactor Manual

Control System (RMCS) and refueling interlocks are inoperable, while

preserving the intended function of the inoperable interlocks.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant

[[Page 49945]]

increase in the probability or consequences of an accident previously

evaluated.

The refueling and one-rod-out interlocks impose barriers to

preclude an inadvertent criticality during refueling operations.

Section 7.7.2.15.1 of the LGS Updated Final Safety Analysis Report

(UFSAR) clearly delineates the functions of the interlocks and the

criteria used in assessing correct refueling and one-rod-out

interlock operation in the following statement.

In all cases, correct operation of the refueling interlock

prevents either the operation of loaded refueling equipment over the

core when any control rod is withdrawn, or the withdrawal of any

control rod when fuel-loaded refueling equipment is operating over

the core. In addition, when the reactor mode switch is in REFUEL

position, only one rod can be withdrawn, and selection of a second

control rod initiates a rod block.

The proposed TS changes provide operational flexibility while

strictly conforming to, and preserving, the intended function of the

refueling and one-rod-out interlocks. The proposed TS changes that

could affect interlock capabilities are identified below, along with

the appropriate justification to substantiate that the proposed TS

changes will not result in an increase in the probability or

consequences of an accident previously evaluated.

a.TS Section 3.9.1, ACTION Statement b. The proposed change to

this existing TS ACTION will add a verification that all control

rods are fully inserted, and then disabled from being withdrawn as a

suitable alternative to placing the reactor mode switch in the

SHUTDOWN position when the one-rod-out interlock is not operable. In

addition, the proposed change to this TS section includes a caveat

of non-applicability for those control rods already removed in

accordance with requirements stipulated in TS Sections 3.9.10.1 and

3.9.10.2. As indicated in LGS UFSAR which is described in the

statement above, it is expected that the refuel and one-rod-out

interlocks will permit the withdrawal of only one (1) control rod at

a time with the reactor mode switch in the REFUEL position, and no

control rods can be moved when fuel-loaded refueling equipment is

operating over the core. By verifying all control rods are inserted,

then disabling withdraw capabilities of all rods, as requested, the

most limiting requirements for control rod motion will be met. The

potential for having more than one (1) control rod out at a time, or

having any control rod not fully inserted while fuel-loaded

refueling equipment is operating over the core, does not exist when

applying the alternative. Therefore, the intended functions of the

refuel and one-rod-out interlocks are operationally preserved. Since

TS Sections 3.9.10.1 and 3.9.10.2 have specific requirements for

removing surrounding fuel prior to control rod blade removal, the

control rods already removed are no longer required to carry out a

safety function in the defueled cell, and as a result would not

apply for this specific proposed TS change. From a control rod

withdrawal perspective, there is no functional difference between

the proposed TS change and the existing, and still remaining, TS

ACTION of locking the reactor mode switch in SHUTDOWN position.

b. TS Section 3.9.1, ACTION Statement c. This existing TS ACTION

requires that core alterations be suspended in the event that a

refueling interlock is not operable. The proposed TS change to this

TS ACTION leaves this requirement in place, but makes this ACTION

specifically applicable to the refueling platform, and adds three

(3) new additional ACTION alternatives. The wording for changes to

this TS section are such that implementation of any one of the three

(3) new alternatives can be substituted for suspending core

alterations. The proposed wording for these three (3) new

alternatives and justification is provided below.

1) Verify all control rods are fully inserted and disable

withdraw capabilities of all control rods***.

Since this alternative ensures all control rods are, and will

remain fully inserted, all required conditions of the associated

refueling and one-rod-out interlocks are met. The refueling

interlock is satisfied since a fuel-loaded refueling platform

operating over the core would be assured that all control rods are

fully inserted and prevented from being withdrawn. The one-rod-out

interlock is satisfied since control rod withdrawal is disabled for

all control rods, which is an even more conservative requirement

than the one-rod-out interlock itself. While operating in this

configuration, there will be no associated travel or hoist

restrictions for the refueling platform over the core, which is

normal for the current refuel interlock design. The potential for

having any control rod not fully inserted while a fuel-loaded

refueling platform is operating over the core, does not exist when

applying this proposed alternative. Therefore, the intended function

of the refueling platform refuel interlocks are operationally

preserved with the implementation of this proposed alternative, and

there will be no increase in the probability of occurrence of an

accident. This proposed alternative also maintains an exclusion (via

a reference to the proposed *** footnote) for control rods removed

in accordance [with] TS Sections 3.9.10.1 and 3.9.10.2. This

exclusion does not apply to inadvertent criticality concerns, as

previously discussed in Item 1.a above.

2) Verify Refuel Platform is not over core (limit switches not

reached) and disable refuel platform travel over core.

As previously stated above, LGS UFSAR Section 7.7.2.15.1

stipulates that the refueling platform position interlocks initiate

a control rod block whenever a fuel-loaded refueling platform is

over the core, and stop a fuel-loaded refueling platform from moving

over the core if a control rod is already withdrawn. This specific

proposed TS change satisfies both these requirements by precluding

the possibility of the platform from being over the core. If a

control rod is being withdrawn, the platform will not be over the

core, and the withdrawal will be in accordance with the current

design. If a control rod is already withdrawn, disabling platform

travel over the core, before reaching the over-core limit switches,

is performing the same function as the existing refueling platform

reverse and forward motion blocks. Therefore, the potential for

having any control rod not fully inserted while a fuel-loaded

refueling platform is operating over the core, does not exist when

applying this proposed alternative. The intended refueling interlock

functions are operationally preserved with the implementation of

this proposed alternative.

3) Verify that no Refuel Platform hoist is loaded and disable

all Refuel Platform hoists from picking up (grappling) a load.

As previously stated above, UFSAR Section 7.7.2.15.1 stipulates

that blocking control rod withdrawal with a refueling platform over

the core, and restricting refueling platform travel from going over

the core with a control rod already withdrawn, are based on the

refueling platform hoist being fuel-loaded. An unloaded platform

without grappling capabilities poses no threat to erroneous fuel

bundle or control rod removal, and eliminates the potential for

having any control rod not fully inserted while a fuel-loaded

refueling platform is operating over the core. Therefore,

implementing this proposed alternative operationally preserves the

intended interlock functions.

c. TS Section 3.9.1, ACTION Statement d. The proposed TS change

adds this new TS ACTION section to specify the refueling interlock

requirements for the service platform, since the applicability of

ACTION Statement c above is being revised to specifically address

refueling interlocks associated with the refueling platform. The

proposed TS changes for new this TS section retain the existing

requirement to suspend core alterations if the service platform

associated refueling interlock is not operable, unless the service

platform is not installed over vessel. The specific proposed TS

changes add two (2) new additional ACTION alternatives. The proposed

wording for these two (2) new ACTION statements are such that

implementation of any one of the two (2) new alternatives can be

substituted for suspending core alterations. Not enforcing

operability requirements on the service platform refueling

interlocks when the service platform is not over the vessel does not

pose an inadvertent criticality concern since there is no associated

hoist to manipulate fuel bundles or control rods. These two (2) new

alternatives are not applicable unless the service platform is

installed over the vessel, and are described below.

1) Verify all control rods are fully inserted and disable

withdraw capabilities of all control rods***.

This alternative ensures that all control rods are, and will

remain, fully inserted which meets the required conditions for

proper refueling and one-rod-out interlock operation. The refueling

interlock is satisfied since a fuel-loaded service platform hoist

operating over-core is assured that all control rods are fully

inserted and prevented from being withdrawn. The one-rod-out

interlock is satisfied since all control rods are disabled, an even

more conservative requirement than the one-rod-out interlock itself.

While operating in this configuration, there will be no associated

hoist restrictions for the service

[[Page 49946]]

platform, which is normal for the current refuel interlock design. The

potential for having any control rod not fully inserted while a

service platform hoist is fuel-loaded over the core, does not exist

when utilizing this proposed alternative. Therefore, the intended

function of the service platform refuel interlocks are operationally

preserved with the implementation of this proposed alternative. This

proposed alternative also maintains an exclusion (via a reference to

the proposed *** footnote) for control rods removed in accordance

with the requirements of TS Sections 3.9.10.1 and 3.9.10.2. This

exclusion is not applicable to inadvertent criticality concerns as

discussed in Item 1.a above.

2) Verify Service Platform hoist is not loaded and disable

Service Platform hoist from picking up (grappling) a load.

As previously described above, UFSAR Section 7.7.2.15.1

stipulates that blocking control rod withdrawal with the service

platform over the core is based on the service platform hoist being

fuel-loaded. An unloaded hoist without grappling capabilities poses

no threat to erroneous fuel bundle or control rod removal and

eliminates the potential for having any control rod not fully

inserted while a fuel-loaded service platform is operating over the

core. Therefore, implementing this proposed alternative

operationally preserves the intended refueling interlock functions.

As discussed in the LGS UFSAR, the use of the refueling and one-

rod-out interlocks are evaluated from a prevention, not a

mitigation, perspective. A Rod Withdrawal Error (RWE) transient

event during refueling is concerned with an inadvertent criticality,

and assumes the reactor vessel head is off, and the plant is

shutdown (i.e, Operating State A). As described in the LGS UFSAR

under Nuclear Safety Operational Analysis (NSOA) Event 16, it is

assumed that the Reactor Protection System (RPS) terminates the

event should the reactor actually reach Operating State B (i.e.,

head off and not shut down), which is conditional on the reactor

mode switch being in the STARTUP position. The proposed TS changes

only pertain to the refueling and one-rod-out interlocks. Since

these interlocks act only in a preventive mode, the consequences of

an inadvertent criticality accident during refueling remain

unchanged.

Since the proposed TS changes are limited to the one-rod-out and

refueling interlocks, they do not affect the reliability of the

associated equipment. The proposed TS changes specify alternative

actions that can be taken in the event that an interlock is

inoperable. These alternative actions serve to ensure the failed

interlock function is preserved, and do not affect the probability

of malfunction of the interlocks.

The one-rod-out and refueling interlocks, as evaluated in the

LGS UFSAR, are designed to preclude an inadvertent criticality

during refueling operations by placing strict controls on fuel

bundle and control rod manipulations, using the following methods.

a. Preventing operation of a fuel-loaded refueling platform or

service platform hoist while over the core if a control rod is

already withdrawn.

b. Preventing a fuel-loaded refueling platform from traveling

over the core if a control rod is already withdrawn.

c. Preventing any control rod from being withdrawn if a fuel-

loaded refueling platform or service platform is already operating

over the core.

d. Preventing the withdrawal of more than one control rod at a

time with the reactor mode switch in the REFUEL position.

The LGS UFSAR indicates that a single component failure does not

cause an interlock failure and that a single interlock failure does

not cause an accident. The proposed TS changes provide alternative

actions that can be taken in the event of an associated component or

interlock malfunction. Implementing the proposed TS changes will

continue to ensure that the intended interlock functions are

maintained and operationally preserved, as described in the LGS

UFSAR.

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 only pertain to the refueling and one-

rod-out interlocks. The refueling and one-rod-out interlocks impose

barriers to preclude an inadvertent criticality during refueling

operations. The proposed TS changes provide operational flexibility,

while strictly conforming to, and preserving, the intended function

of the refueling and one-rod-out interlocks. There is no other

potential failure mode for these interlocks than has already been

evaluated and described in the LGS UFSAR. Implementation of these

proposed changes will maintain and operationally preserve the

intended interlock functions. Therefore, the malfunction of any

associated component or interlock will not adversely impact the

plant and any other equipment important to safety, directly or

indirectly.

Therefore, the proposed TS changes do not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed TS changes only affect the TS associated with the

one-rod-out and refueling interlocks. The associated TS Bases

Section 3/4.9, ``Refueling Operations,'' states that the one-rod-out

and refueling interlocks maintain conditions during refueling

activities that reinforce refueling procedures and reduce the

potential for the probability of occurrence of each of the following

conditions:

a. Inadvertent criticality,

b. Damage to reactor internals or fuel assemblies, and

c. Exposure of personnel to excessive radioactivity.

The proposed TS changes do not adversely affect the one-rod-out

or refueling interlocks. The associated interlocks will continue to

perform the refueling functions required to maintain the present

margin of safety. The proposed TS changes only contain alternative

actions that can be taken in the event an interlock is inoperable.

These proposed alternative actions ensure that the intent of the

interlocks is preserved, and that there is no reduction in the

ability of the interlocks to maintain adequate refueling conditions.

The proposed TS changes will preserve the intended interlock

functions, and maintain the existing level of protection against

refueling errors that could lead to an inadvertent criticality,

damage to reactor internals or fuel assemblies, or excessive

personnel radiation exposure. The one-rod-out and refueling

interlocks will continue to function with their present degree of

reliability. The proposed TS changes will continue to maintain

strict controls on fuel bundle and control rod manipulations to

avoid inadvertent criticality. The proposed TS changes provide the

same level of assurance regar[d]ing the manipulation of control rods

during refueling operations as that currently described in the LGS

UFSAR, and as discussed below.

a. Preventing operation of a fuel-loaded refueling platform or

service platform hoist while over the core if a control rod is

already withdrawn.

b. Preventing a fuel-loaded refueling platform from traveling

over the core if a control rod is already withdrawn.

c. Preventing any control rod from being withdrawn if fuel-

loaded refueling platform or service platform is already operating

over the core.

d. Preventing the withdrawal of more than one control rod at a

time with the reactor mode switch in the REFUEL position.

Therefore, the proposed TS changes do not involve a reduction in

a margin of safety.

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

this eview, 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

South Carolina Electric & Gas Company (SCE&G), South Carolina

Public Service Authority, Docket No. 50-395, Virgil C. Summer

Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: February 21, 1995, as revised on August

31, 1995

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to reflect changes to 10 CFR

Part 20 (including Appendix B, Table 2 concentrations) and provide

additional administrative corrections.

[[Page 49947]]

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 probability or consequences of an accident previously

evaluated does not involve a significant increase.

The proposed TS changes showing the relocation of the old 10 CFR

20.106 requirements to the new 10 CFR 20.1302, the old 10 CFR

20.203(c)(2) requirements to the new 10 CFR 20.1601(a), and the old

10 CFR 20.407 requirements to the new 10 CFR 20.2206(b) will not

involve a significant increase in the probability or consequences of

an accident previously evaluated because there will be no change in

the types and amounts of effluents that will be released, nor will

there be an increase in individual or cumulative occupational

radiation exposures.

The proposed revision to the liquid and gaseous release rate

limits will not involve a significant increase in the probability or

consequences of an accident previously evaluated because there will

be no change in the types and amounts of effluents that will be

released, nor will there be an increase in individual or cumulative

occupational radiation exposures. This is only a change to the

method of (algorithm) determining release rate limits and will not

change net limits or change the more restrictive 10 CFR 50 Appendix

I dose limits.

The proposed revision to the radioactive material quantity in

the settling pond and its associated TS Bases will not involve a

significant increase in the probability or consequences of an

accident previously evaluated because there will be no change in the

types of effluents that will be released, nor will there be an

increase in individual or cumulative occupational radiation

exposures. This is only a change to the quantity of radioactive

material in the settling pond and will conservatively lower net

limits.

The proposed revision to the TS bases for the liquid holdup tank

activity limit will not involve a significant increase in the

probability or consequences of an accident previously evaluated

because there will be no change in the types and amounts of

effluents that will be released, nor will there be an increase in

individual or cumulative occupational radiation exposures. The curie

limit is not affected, therefore, the change does not represent a

decrease in the level of control previously evaluated.

The proposed revision to the distance at which dose rates are

measured from the radiation source or surface will not involve a

significant increase in the probability or consequences of an

accident previously evaluated because there will be no increase in

the individual or cumulative occupational radiation exposures. The

change in distance is conservative in its effect on worker

protection and is in conformance with 10 CFR 20.1601 requirements.

2. The possibility of a new or different kind of accident from

any previously evaluated is not created.

The proposed TS changes showing the relocation of the old 10 CFR

20.106 requirements to the new 10 CFR 20.1302, relocation of the old

10 CFR 20.203(c)(2) requirements to the new 10 CFR 20.1601(a), and

relocation of the old 10 CFR 20.407 requirements to the new 10 CFR

20.2206(b) will not create the possibility of a new or different

kind of accident from any previously evaluated because the revisions

are administrative and will not change the types and amounts of

effluents that will be released.

The proposed revision to the liquid and gaseous release rate

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

accident from any previously evaluated because the revision is

administrative and will not change the types and amounts of

effluents that will be released.

The proposed revision to the quantity of radioactive material in

the settling pond and its associated TS Bases will not create the

possibility of a new or different kind of accident from any

previously evaluated because the revision will not change the types

of effluents that will be released. This is only a change to the

quantity of radioactive material in the settling pond and will

conservatively lower net limits.

The proposed revision to the TS bases for the liquid holdup tank

activity limit will not create the possibility of a new or different

kind of accident from any previously evaluated because the revision

is administrative and will not change the types and amounts of

effluents that will be released.

Implementation of the more conservative distance at which dose

rates are measured will not create the possibility of a new or

different kind of accident from any previously evaluated.

3. A significant reduction in a margin of safety is not

involved.

The proposed revisions due to the location of requirements will

not reduce a margin of safety because they are administrative in

nature. No equipment or procedural changes are postulated. There is

no impact on any margin of safety.

The proposed revision to liquid and gaseous release rate limits

will not reduce a margin of safety because it is administrative in

nature. These revisions preserve the existing level of effluent

control. No changes to the more restrictive 10 CFR 50 Appendix I

dose limits are made. There are no equipment or operational

procedure changes, therefore, no accidents of any kind will be

created by this change.

The proposed revision to the quantity of radioactive material in

the settling pond and its associated TS Bases will not reduce a

margin of safety because it is conservative in nature and preserves

the existing level of effluent control. There are no equipment or

operational procedure changes required, therefore, no accidents of

any kind will be created by this change.

The proposed revision to the TS bases for the liquid holdup tank

activity limit will not reduce a margin of safety because it is

administrative in nature and preserve[s] the existing level of

effluent control. No equipment or procedural changes are postulated.

There is no impact on any margin of safety.

The change in distance for a High Radiation Area classification

from 18 in.(45 cm) to (30 cm)12 in. from the radiation source or

surface will not reduce the margin of safety because this change

will reduce the worker's stay time in the area and therefore

minimize exposure.

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: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218

NRC Project Director: Frederick J. Hebdon

Southern California Edison Company, et al., Docket Nos. 50-361 and

50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,

San Diego County, California

Date of amendment requests: July 19, 1995

Description of amendment requests: The licensee proposes to revise

technical specifications (TSs) to (1) support modifications to the

containment area radiation monitors, to either upgrade or replace

existing equipment with state-of-the-art equipment, (2) relocate the

setpoint and allowable values for the control room airborne radiation

monitors to be consistent with the containment airborne radiation

monitors TS, and (3) make minor editorial changes to the TS pertaining

to the control room airborne radiation monitors and the containment

airborne radiation monitors. The proposed changes affect TS Tables 3.3-

3, 3.3-4, 3.3-5, 3.3-6, 4.3-2, and 4.3-3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Control Room Airborne Radiation Monitors

The proposed change would permit relocation of the setpoint and

allowable values for the monitors from the Technical Specifications

(TSs) to the administrative control procedures. This change is

consistent with the existing Containment Airborne Radiation Monitor

TSs. This change will not prevent the radiation monitors from

[[Page 49948]]

performing their intended function following a design basis accident.

Therefore, operation of the facility in accordance with this change

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Containment Area Radiation Monitors

The proposed change deletes the existing Containment Area

Radiation Monitors RE-7856-1 and RE-7857-2 and their Engineered

Safety Feature Actuation System (ESFAS) function to initiate

containment purge isolation on high radiation in containment. The

deletion of this ESFAS function does not create a precursor to any

analyzed accident since these monitors are for accident mitigation

only.

Currently, no release of radioactivity is assumed during a Fuel

Handling Accident in containment since the Containment Area

Radiation Monitors detect and isolate containment purge prior to

release. The proposed deletion will cause some release prior to

detection and isolation of purge by the remaining noble gas

containment monitors. The consequences of a Fuel Handling Accident

inside containment were previously re-evaluated, assuming no

containment purge isolation, to resolve inconsistencies in the

original analysis assumptions and methodology. The results of the

calculation indicated off-site doses well within the limits of 10

CFR 100 and Control Room doses that met the limits of 10 CFR 50

Appendix A General Design Criterion 19. Containment purge isolation

on high gaseous activity during a Fuel Handling Accident will still

be available with this proposed change but is not required for the

dose consequences to remain within the dose criteria. Therefore, the

proposed change will not significantly increase the consequences of

a Fuel Handling Accident inside containment.

The Loss of Coolant Accident (LOCA) function of the Containment

Purge Isolation System (CPIS) signal will be essentially unaffected

by this proposed change. Currently, containment purge isolation

(containment minipurge) on high radiation signals is a diverse

signal with Safety Injection Actuation System (SIAS) and Containment

Isolation Actuation System (CIAS). In a LOCA event, containment

purge isolation is expected to occur on either SIAS or CIAS prior to

a CPIS signal on high radiation in containment. While this proposed

change reduces the diversity of radiation monitoring inputs, the

diversity of parameters measured (pressure and radiation) is still

preserved. Therefore, the proposed change will not increase the

consequences of a LOCA.

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

or different kind of accident from any accident previously

evaluated.

Control Room Airborne Radiation Monitors

Relocating the monitor setpoint and allowable values from the

TSs to the administrative procedures would not alter the design and

operational interface between the Control Room Isolation System

instrumentation and existing plant equipment. As such, the monitors

would continue to operate and perform their intended safety function

to isolate the control room following a design basis accident as

before. Therefore, operation of the facility in accordance with this

proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Containment Area Radiation Monitors

The deletion of the Containment Area Radiation Monitors will not

alter the operation of CPIS. The remaining interface between CPIS

and existing plant equipment will continue to perform their intended

safety function to isolate containment purge by closing the

containment purge valves. This function will continue to be

performed by Containment Airborne Radiation Monitors 2(3) RT-7804-1

and 2(3) RT-7807-2. Therefore, operation of the facility in

accordance with this proposed change will not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Control Room Airborne Radiation Monitors

Relocating the monitor setpoint and allowable values to the

administrative procedures would not alter the existing margin of

safety. The relocation would only relinquish control of the setpoint

and allowable values from the TSs to quality-affecting (changes will

require a 10 CFR 50.59 evaluation) procedures. Therefore, operation

of the facility will not involve a significant reduction in a margin

of safety.

Containment Area Radiation Monitors

The proposed change does not affect the margin of safety in

Modes 1 through 4 since the diversity of the parameters measured is

maintained for minipurge isolation. Either SIAS, CIAS, CPIS, or

manual operation will close the containment mini purge valves. The

main purge is sealed closed during Modes 1 through 4 with the purge

valves closed and deactivated.

The diversity of the parameters measured is not maintained for

the containment main purge isolation. The main purge is only

applicable during Modes 5 and 6 and main purge isolation is

initiated only by either CPIS or manual operation. This proposed

change along with the previously submitted PCN-299 reduces the

diversity of radiation sensing in containment for CPIS generation

from four types (gaseous, iodine, particulate, and gamma) to one

type (gaseous activity). Since the consequences of a Fuel Handling

Accident inside containment without purge isolation have been

calculated to be well within 10 CFR 100 dose limits, the loss of

diversity for this accident does not result in a significant

reduction in a margin of safety. Therefore, this 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 requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, P. O. Box 19557, Irvine, Califor

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