Biweekly Notice

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 March 3, 1995, through March 17, 1995. The

last biweekly notice was published on March 15, 1995.

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

[[Page 16182]] of the facility in accordance with the proposed

amendment would not (1) involve a significant increase in the

probability or consequences of an accident previously evaluated; or (2)

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

accident previously evaluated; or (3) involve a significant reduction

in a margin of safety. The basis for this proposed determination for

each amendment request is shown below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

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

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

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

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

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

intervene is discussed below.

By April 28, 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

[[Page 16183]] telephone number, date petition was mailed, plant name,

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

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

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and

to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Date of amendment request: February 24, 1995

Description of amendment request: The proposed change would remove

Section 4.3 from the Technical Specifications (TS) because the primary

system testing following opening is already performed in accordance

with the American Society of Mechanical Engineers Boiler and Pressure

Vessel Code, as implemented in the licensee's inservice inspection

program as required by TS 4.0.1.

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:

This change does not involve a significant hazards consideration

for the following reasons.

1. The requested change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. This requested change will provide consistency between

our Technical Specifications (TS) and 10 CFR 50.55a which requires

testing in accordance with Section XI of the ASME Boiler and

Pressure Vessel Code. The requirements contained in TS Section 4.3

were placed into TS prior to incorporation of Section XI into the

ASME Boiler and Pressure Vessel Code. The NRC and industry have

since recognized the ASME Boiler and Pressure Vessel Code, Section

XI as the appropriate testing program. Adequate assurance of primary

system integrity will be provided since primary system testing will

continue to be controlled and performed in accordance with the rules

for inservice inspections provided by ASME Boiler and Pressure

Vessel Code, Section XI as implemented by our approved In-Service

Inspection (ISI) Program, as required by TS Section 4.0.1.

Therefore, there would be no increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated. The requested change deletes the current TS requirements

for primary system testing by recognizing that we will continue to

perform required testing consistent with 10 CFR 50.55a and ASME

Boiler and Pressure Vessel Code, Section XI, as implemented by our

approved ISI Program, as required by TS Section 4.0.1. This

requested change does not involve the addition or modification of

plant equipment, nor does it alter the design or operation of plant

systems. Therefore, the requested change does not create the

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

previously evaluated.

3. The requested change does not involve a significant reduction

in a margin of safety. The requested change deletes the current TS

Section 4.3 requirements for primary system testing and maintains

the margin of safety by continuing to perform required testing in

accordance with 10 CFR 50.55a and ASME Boiler and Pressure Vessel

Code, Section XI, as implemented by our approved ISI Program, as

required by TS Section 4.0.1. Therefore, the requested change does

not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

NRC Project Director: William H. Bateman

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of amendment request: March 3, 1995

Description of amendment request: The proposed amendment would

eliminate the requirement to perform periodic measurement testing of

the response times for selected pressure and differential pressure

sensors. The requirement that reactor trip and engineered safety

feature response time functions be within their specified limit at

least once per 18 months will be verified instead of demonstrated. The

associated bases section for response time requirements will be changed

to allow the sensor response time portion of the channel response time

to use historical records, testing results, or vendor supplied

engineering specifications. No other changes to response time methods

are included in this change.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment does not result in a condition where the

design, material, or construction standards that were

applicable prior to the change are altered nor does it modify

any system interface. The same Reactor Trip System and Engineered

Safety Features Actuation System instrumentation is being used; the

time response allocations/modeling assumptions in the Final Safety

Analysis Report (FSAR) Chapter 15 analyses are still the same; only

the method of verifying time response is changed. The proposed

activity will not change, degrade, or prevent actions or alter any

assumptions previously made in evaluating the radiological

consequences of an accident described in the FSAR. Therefore, there

would be no increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed amendment does not alter the performance of the

pressure and the differential pressure transmitters used in the

plant protection systems. The sensors will still have response time

verified by test before placing the sensor in operational service

and after any maintenance that could affect response time. Changing

the method of periodically verifying instrument response for certain

sensors (assuring equipemt operable) from time response testing to

calibration and channel checks will not create any new accident

initiators or scenarios. Periodic surveillance of these instruments

will detect significant degradation in the sensor response

characteristic. Therefore, the proposed changes do not create the

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

previously evaluated. [[Page 16184]]

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed amendment to [sic] does not affect the total system

response time assumed in the safety analysis. The periodic system

response time verification method for selected pressure and

differential pressure sensors is modified to allow use of actual

test data or engineering data. The method of verification still

provides assurance that the total system response is within that

defined in the safety analysis, since calibration tests will detect

any degradation which might significantly affect sensor response

time. Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

NRC Project Director: William H. Bateman

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

Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.

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

Will County, Illinois

Date of amendment request: May 20, 1994, as supplemented February

2, 1995

Description of amendment request: The proposed amendment would

permit the licensee to use an alternate repair criteria (ARC),

designated as the F* criteria. Use of the F* criteria would

allow tubes with otherwise pluggable indications, to remain in service

as long as the indications are below the designated minimum distance of

the F* criteria. The F* criteria for Byron and Braidwood

defines a length of 1.7 inches of undegraded expanded tube within the

tubesheet as the minimum distance acceptable for implementing the ARC.

Below the F* length, a circumferential tube defect can exist and

the tube can remain in service. The proposed amendment will change the

plugging limit definition and would exclude plugging steam generator

tubes with indications that satisfy the F* criteria. The F*

criteria maintains the structural integrity of the degraded tube as the

primary pressure boundary and allows the tube to remain in service for

heat transfer and core cooling.

This alternate repair criteria qualification is documented in

Babcock & Wilcox Nuclear Technologies (BWNT) Topical Report BAW-10196 P

Revision 1, ``W-D4 F* Qualification Report,'' which is included as

part of the licensee's submittal. The staff's proposed no significant

hazards consideration determination for the requested change was

published on July 6, 1994 (59 FR 34659). In response to the staff's

request for additional information by letter dated February 2, 1995,

the licensee revised their previous submittal.

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:

A. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The supporting qualification report for subject criteria

demonstrates that the presence of the tubesheet will enhance the

tube integrity in the region of the tube-to-tubesheet roll

expansions by precluding tube deformation beyond its initial

expanded outside diameter. The resistance to a tube rupture is

strengthened by the presence of the tubesheet in that region. The

results of hardrolling of the tube into tubesheet provides a

mechanical leak limiting seal between the tube and the tubesheet. A

tube rupture cannot occur because the contact between the tube and

the tubesheet does not permit sufficient movement of tube material.

The type of degradation for which the F* criteria has been

developed (cracking with a circumferential orientation) can

theoretically lead to a postulated tube rupture event provided that

the postulated through-wall circumferential crack exists near the

top of the tubesheet. An evaluation including analysis and testing

has been done to determine the resistive strength of the expanded

tubes within the tubesheet. This evaluation provides the basis for

the acceptance criteria for tube degradation subject to the F*

criteria. The F* length of roll expansion is sufficient to

preclude tube pullout from tube degradation located below the

F* distance, regardless of the extent of the tube degradation.

The Technical Specification leakage rate requirements and accident

analysis assumptions remain unchanged in the unlikely event that

significant leakage from this region does occur. For consistency

with current offsite dose limits, the site allowable leakage limit

during a MSLB has been conservatively calculated to be 12.8 gpm for

Byron and 9.1 gpm for Braidwood, which includes the accident leakage

from IPC in addition to the accident leakage from F* on the

faulted steam generator and the operational leakage limit. The

operational leakage limit of Specification 3.4.6.2.c in each of the

three remaining intact steam generators shall include the

operational leakage from F*. As a requirement for operation

following application IPC, the projected distribution of crack

indications over the operating period must be verified to result in

primary to secondary accident leakage less than the site allowable

leakage limit. Thus, the consequences of a MSLB remain unchanged.

The tube rupture and pullout is fully bounded by the existing

steam generator tube rupture analysis included in the UFSAR. The

leakage testing of the roll expanded tubes indicates that for tube

expansion lengths approximately equal to the * distance, any

postulated primary to secondary leakage from * tubes would be

insignificant. The proposed alternate repair criteria does not

adversely impact any other previously evaluated design basis

accident.

The leakage from an F* tube would be limited by the tube-

to-tubesheet interface since this leak would occur below the

secondary face of the tubesheet. Qualification testing and previous

experience indicate that normal and faulted leakage is well below

Technical Specification and administrative limits creating no

increase in the consequences associated with tube rupture type

leakages. The UFSAR analyzed accident scenarios are still bounding

since the normal and faulted leak rates are well within the normal

operating limit of 150 gallons per day. This conclusion is

consistent with previous F* programs approved and used at other

operating plants.

All of the design and operating characteristics of the steam

generator and connected systems are preserved since the F*

criteria utilizes the ``as rolled'' tube configuration that exists

as part of the original steam generator design. The F* joint

has been analyzed and tested for design, operating, and faulted

condition loadings in accordance with Regulatory Guide 1.121 safety

factors. The potential for a tube rupture is not increased from the

original submittal as demonstrated in the qualification analyses and

testing completed in the BWNT report.

Therefore, this change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

B. The proposed changes do not create the possibility of a new

or different type of accident from any accident previously

evaluated.

Implementation of the proposed F* criteria does not

introduce any changes to the plant design basis. Use of the criteria

does not provide a mechanism to initiate an accident outside of the

region of the expanded portion of the tube. In the unlikely event

the failed tube severed completely at a point below the F*

region, the remaining F* joint would retain engagement in the

tubesheet due to its length of expanded contact within the tubesheet

bore. This engagement length would prevent any interaction of the

severed tube with neighboring tubes. Any hypothetical accident as a

result of any tube degradation in the expanded region of the tube

would be bounded by the existing tube rupture accident analysis.

Tube bundle structural integrity will be maintained. Tube bundle

[[Page 16185]] leak tightness will be maintained such that any

postulated accident leakage from F* tubes will be negligible

with regard to offsite doses.

Therefore, there is not a potential for creating the possibility

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

evaluated.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

The use of the F* criteria has been demonstrated to

maintain the integrity of the tube bundle commensurate with the

requirements of Regulatory Guide 1.121 and the primary to secondary

pressure boundary under normal and postulated accident conditions.

Acceptable tube degradation for the * criteria is any

degradation indication in the tubesheet region, more than the

F* distance from the secondary face of the tubesheet or the top

of the last hardroll contact point whichever is further into the

tubesheet. The safety factors used in the verification of the

strength of the degraded tube are consistent with the safety factors

in the ASME Boiler and Pressure Vessel Code and Regulatory Guide

1.121 used in steam generator design. The * distance has been

verified by various testing to be greater than the length of the

roll expanded tube-to-tubesheet interface required to preclude both

tube pullout and significant leakage during normal and postulated

accident conditions. The protective boundaries of the steam

generator continue to be maintained with the use of the F*

criteria. A tube with the indication of degradation previously

requiring removal from service can be kept in service through the

F* criteria. Since the joint is contained within the tubesheet

bore, there is no additional risk associated with the previously

analyzed tube rupture event. The leak testing acceptance criteria

are based on the primary to secondary leakage limit in the Technical

Specifications and the leakage assumptions used in the UFSAR

accident analyses.

Implementation of the alternate repair criteria will decrease

the number of tubes which must be taken out of service with tube

plugs or repaired by sleeves. Both plugs and sleeves reduce the RCS

flow margin; thus, implementation of the F* criteria will

maintain the margin of flow that would otherwise be reduced in the

event of increased plugging or sleeving.

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

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

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

requested amendments involve no significant hazards consideration.

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

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

Braidwood, the Wilmington Township Public Library, 201 S. Kankakee

Street, Wilmington, Illinois 60481.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee

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

Date of amendment request: November 22, 1994, as supplemented

January 30, March 2, and March 13, 1995.

Description of amendment request: This request was previously

published in the Federal Register on February 15, 1995 (60 FR 8746). It

is being renoticed to provide clarification to the scope of the

original request. The amendments would revise Technical Specification

(TS) 3.8 to establish restricted loading patterns and associated burnup

criteria for placing fuel in the Oconee spent fuel pools. In addition,

the Design Features sections associated with the reactor and fuel

storage would be revised. These changes are necessary to address two

new fuel designs which have increased initial fuel enrichment and

therefore cannot be stored in the spent fuel pools under existing TS or

loaded into the reactor. An administrative change would be made to TS

6.9.1 to include spent fuel pool boron concentration in the Core

Operating Limits Report. Other administrative changes would be made in

the Design Features section to make the specification consistent with

wording in the standard TS. Finally, the two additional supplements to

the original request are referenced herein.

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:

Standard 1. The proposed amendments will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Each accident analysis addressed in the Oconee Final Safety

Analysis Report (FSAR) has been examined with respect to changes in

Cycle 15 parameters to determine the effect of the Cycle 16 reload

and to ensure that the acceptance criteria of the FSAR safety

analyses remain satisfied. The transient evaluation of Cycle 16 is

considered to be bounded by previously accepted analyses. Section 7

of the Reload Report addresses ``Accident and Transient Analysis''

for this core reload.

There is no increase in the probability or consequences of an

accident due to the spent fuel storage restrictions proposed in this

amendment request. It has been shown that the calculated, worst case

keff for this area is [less than or equal to] 0.95 under all

conditions. There is no increase in the probability of a fuel drop

accident in the SFP [spent fuel pool] since the mass of the new

assemblies is not significantly different from the mass of the old

assemblies. The likelihood of other accidents, previously evaluated

and described in the FSAR, is also not affected by the proposed

changes. In fact, it could be postulated that since the increase in

fuel enrichment will allow for extended fuel cycle lengths, there

will be a decrease in fuel movement and the probability of an

accident may actually be reduced. There is also no increase in the

consequences of a fuel rod drop accident in the SFP since the

fission product inventory of individual fuel assemblies will not

change significantly as a result of increasing the initial

enrichment. In addition, no change to safety related systems is

being made. Therefore, the consequences of a fuel rupture accident

remain unchanged. In addition, it has been shown that Keff all

conditions. Therefore, the consequences of a criticality accident in

the SFP remain unchanged as well. The above analysis ensures that

the proposed reload amendment request will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The analyses performed in support of this reload are in

accordance with the NRC approved methods delineated in Specification

6.9.2. The predicted operating characteristics of Oconee 3 Cycle 16

are similar to previously licensed designs. The Mark B10T and Mark

B11 fuel assembly designs remain mechanically compatible with all

fuel handling equipment. Therefore, no new or different kind of fuel

handling accident is created by the proposed amendment request.

Section 15.11 of the Oconee FSAR states that the refueling boron

concentration is maintained such that a criticality accident during

refueling is not considered credible. The proposed amendment request

continues to assure that a criticality accident in the SFP or during

refueling is not credible. The double contingency principle

discussed in ANSI N-16.1-1975 and the April 1978 NRC letter allows

credit for soluble boron under other abnormal or accident

conditions, since only a single accident need be considered at one

time. Thus, by requiring a minimum boron concentration in the SFP, a

criticality accident caused by violating the SFP storage

restrictions is not considered credible. Therefore, the proposed

amendment request does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The Oconee 3 Cycle 16 design was performed using the NRC

approved methods given in Specification 6.9.2. The safety limits for

Oconee 3 Cycle 16 are unchanged from previous cycles. The limits and

margins summarized in the Oconee 3 Cycle 16 Reload Report are well

within the allowable limits and requirements, and reflect no

reductions to any margins of safety.

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

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

[[Page 16186]] satisfied. Therefore, the NRC staff proposes to

determine that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 14, 1993, as supplemented by

letter dated March 3, 1995.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) by removing the reactor

vessel material specimen withdrawal schedule and by updating the

reactor coolant system pressure-temperature (P-T) curves. The specimen

withdrawal schedule will be relocated to the Updated Final Safety

Analysis Report (UFSAR). The original Notice was published on January

19, 1994 (59 FR 2867).

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:

Although the Reactor Vessel material specimens withdrawal

schedule will be removed from the Technical Specifications, the

Technical Specifications bases will continue to provide background

information on the use of the data obtained from material specimens.

Also, updates to the schedule will continue to be submitted to the

NRC for approval prior to implementation.

Operating the plant in accordance with the new, updated P-T

Curves will assure preserving the structural integrity of the

reactor vessel over the life of the plant. The pressure and

temperature limits were developed in accordance with 10 CFR [Part]

50 Appendix G requirements.

Removing the requirements associated with the previous exemption

to Appendix H (TS 4.4.8.1.2 items a & b) is purely an administrative

change.

Therefore, the proposed changes will not significantly increase

the probability or consequences of any accident previously

evaluated.

Removal of the Reactor Vessel material specimen schedule from

the Technical Specifications has no impact on accidents at the

plant. Updates to the schedule will still be required to be

submitted to the NRC prior to implementation per Section II.B.3 of

Appendix H to 10 CFR Part 50.

Also, updates to the P-T Curves will not create a new or

different type [of] accident. The reactor vessel beltline P-T limits

were revised applying the general guidance of the ASME Code,

Appendix G procedures with the necessary margins of safety for

heatup, cooldown and inservice hydro test conditions.

The change to TS 4.4.8.1.2 items a & b is purely administrative.

Therefore, the proposed changes will not create the possibility

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

evaluated.

Removal of the schedule for Reactor Vessel material specimen

withdrawal from the Technical Specifications does not impact the

margin of safety. The schedule will continue to receive NRC review

and approval prior to implementation of updates to the schedule.

Updates to the P-T Curves are provided to preserve the margin to

[sic] safety to assure that when stressed under operating,

maintenance and testing the boundary behaves in a non-brittle manner

and the probability of rapidly propagating fracture is minimized.

The change to TS 4.4.8.1.2 items a & b is purely administrative.

Therefore, the proposed changes will not result in a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122

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

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

NRC Project Director: William D. Beckner

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: February 22, 1995

Description of amendment request: The proposed changes are

administrative in nature in that reference to an ``automatic''

containment air lock tester will be deleted from TS 4.6.1.3. The

automatic airlock tester is no longer being used.

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 amendment is administrative in nature in that the

revision will eliminate the wording associated with optional use of

the personnel airlock automatic leakage tester. The requirement for

testing the personnel airlock at a pressure greater than or equal to

Pa for at least 15 minutes remains unchanged. The acceptance

criteria of personnel airlock seal leakage less than 0.01 La is

also unchanged. The automatic leakage tester is not an accident

initiator nor a part of the success path(s) which function to

mitigate accidents evaluated in the plant safety analyses. The

proposal does not involve any changes to the configuration or method

of operation of any plant equipment that is used to mitigate the

consequences of an accident, nor does it alter any assumptions or

conditions in the plant safety 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.

(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 to remove the reference to the personnel

airlock automatic tester from the technical specifications will not

introduce any new failure modes or system interactions, nor will it

require the installation of any new or modified equipment. The

requirement to leak test the personnel air locks will not be

changed. Thus, 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 is administrative in nature in that it

eliminates the reference to the personnel airlock automatic leakage

tester but does not alter the surveillance and acceptance criteria

for such testing. Seal leakage testing is performed in accordance

with an approved plant procedure which allows use of either an

automatic tester or a portable testing cart. The automatic leakage

tester is not used to actuate safety related equipment, provide

interlocks, or perform plant control functions. The conditions

evaluated in the plant accident and transient analyses do not

involve this tester. The proposed change does 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

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

Based on the above discussion and the supporting Evaluation of

Technical Specification changes, FPL has determined that the

proposed license amendment involves no significant hazards

consideration.

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

this [[Page 16187]] 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: Harold F. Reis, Esquire, Newman and

Holtzinger, 1615 L Street, NW., Washington, DC 20036

NRC Project Director: David B. Matthews

Florida Power and Light Company, et al., Docket No. 50-389, St.

Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: February 27, 1995

Description of amendment request: The proposed amendment will

change Table 3.3-3 and 3.3-4 to accommodate an improved coincidence

logic and relay replacement for the 4.16 kV Loss of Voltage Relays.

Actions required for certain trip units with the number of operable

channels one less than the total number of channels will also be

changed. In addition, the format used to state the time delay for the

4.16 kV Degraded Voltage trip unit will be revised.

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 change will result in a better overall posture of

the plant under degraded/loss of voltage conditions. The design

upgrade for the 4.16 kV Loss of Voltage system is more reliable, has

inherently higher accuracy, and is easier to maintain and calibrate

in the field. The coincidence logic will eliminate the spurious

plant trip potential from the existing design. Restating the maximum

time delay for the 4.16 kV Degraded Voltage (coincident with SIAS

[safety injection actuation signal]) protective relays in a ``less

than'' format will assure that the transfer of power to the on-site

sources occurs before the level of voltage becomes injurious to the

equipment under accident conditions, and will ensure that stripping

of the emergency power busses and loading of the EDG (s) [emergency

diesel generators] will occur within the time allowed by original

design criteria. The maximum allowed time delay for this function is

not being increased, and the time delay assumed in the accident

analyses for connecting the emergency bus to the diesel generator

will not be exceeded. 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 does not change the operation, function

or modes of plant operation. The ability of the loss of power and

degraded grid voltage protection scheme to properly transfer from

the off-site to the on-site power sources is being maintained. The

relays in the improved design of the 4.16 kV Loss of Voltage

function are of the type presently being used in identical

applications at both St. Lucie plant units. No new hazards are

created or postulated which may cause an accident different from any

accident previously analyzed. The modifications will result in a

more sensitive protection scheme allowing continuous operation

without unnecessary challenges to the safety systems, and will

continue to provide adequate protection to all the safety equipment.

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 capability of the loss of power and degraded grid voltage

protection scheme is enhanced by the changes being proposed and is

confirmed by the existing surveillance requirements. The planned

modifications to the 4.16 kV Loss of Voltage function will result in

a more sensitive undervoltage detection system and reduce the

possibility of spurious actuation. The maximum time assumed in the

safety analyses for connecting each Emergency Bus to its dedicated

Emergency Diesel Generator is not being changed, and assurance that

separation from a degraded off-site power source will occur before

this time interval is exceeded during accident conditions will be

maintained by the proposed amendment. Accordingly, the margin of

safety is not affected. Therefore, operation of the facility in

accordance with the proposed amendment would not involve a

significant reduction in a margin of safety.

Based on the discussion presented above and on the supporting

Evaluation of Proposed TS [Technical Specifications] Changes, FPL

has concluded that this proposed license amendment involves no

significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

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

Holtzinger, 1615 L Street, NW., Washington, DC 20036

NRC Project Director: David B. Matthews

Florida Power and Light Company, et al., Docket No. 50-389, St.

Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: February 27, 1995

Description of amendment request: The proposed amendment will

modify surveillance requirement (SR) 4.9.8.1 and 4.9.8.2 to allow a

reduction in the required minimum shutdown cooling flow rate under

certain conditions during operational MODE 6. In addition, the format

of the SR will be changed to clarify the intent of the stated

surveillances.

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.

Operation of the SDCS [shutdown cooling system] is not an

accident initiator and, therefore, does not significantly increase

the probability of an accident previously evaluated.

The proposed change will allow a plant configuration needed to

perform maintenance activities on LPSI [low-pressure safety

injection]/SDCS headers by isolating one injection flow line for an

operable SDCS train during certain MODE 6 conditions. In the event

of a failure or unavailability of the alternate SDCS train, this

configuration could result in the proposed minimum flow rate.

The proposed change only modifies the minimum required flow

rate, and does not affect the probability of this event. FPL has

evaluated the proposed value of reactor coolant flow and has shown

that the bases for the existing LCO [limiting condition for

operation] will continue to be satisfied. Therefore, there are no

significant increases in the consequences of any event from the

proposed change. No other system interactions are involved related

to previously evaluated accidents, and the proposed change has no

adverse effect on any other system performance.

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 change does not affect the normal operation of the

plant. No new [[Page 16188]] systems are introduced and there is no

adverse effect on any other system configuration or performance. The

change will, however, allow isolation of one SDCS injection flow

path for maintenance activities in MODE 6 under controlled

conditions. The failure of the alternate SDCS train does not create

a new accident and has been further evaluated in the reduced flow

configuration, and shown to meet all the TS bases requirements.

Therefore, operation of the facility in accordance with the proposed

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

of accident from any accident previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The safety considerations related to the proposed change are

described in the bases to TS [Technical Specification] 3/4.9.8. FPL

has evaluated the proposed reduction in SDCS flow requirement, under

stated conditions, and has shown that the proposed flow rate meets

all the TS bases requirements involving decay heat removal, boron

dilution, and stratification. Established acceptance criteria

providing margins of safety are not being changed by the proposed

amendment. Therefore, operation of the facility in accordance with

the proposed amendment would not involve a significant reduction in

a margin of safety.

Based on the discussion presented above and on the supporting

Evaluation of Proposed TS Changes, FPL has concluded that this

proposed license amendment involves no significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location:Indian River Junior College

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

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

Holtzinger, 1615 L Street, NW., Washington, DC 20036

NRC Project Director: David B. Matthews

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket No.

50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County,

Georgia

Date of amendment request: March 14, 1995

Description of amendment request: Georgia Power Company (GPC or the

licensee) has proposed a temporary change to Hatch Unit 2 Technical

Specification (TS) Required Action 3.3.6.1.F.1, and associated Bases.

The proposed change would add a note to the Primary Containment

Isolation Instrumentation actions to permit the drywell and wetwell

purge valves which are isolated by the drywell radiation monitor signal

to be opened with one inoperable drywell radiation monitor. The note

will expire prior to startup from the Hatch Unit 2 refueling/

maintenance outage scheduled in the fall of 1995, at which time the

radiation monitor can be repaired or replaced. Should the unit be

forced into a cold shutdown of sufficient duration (i.e., drywell de-

inerted), the inoperable radiation monitor will be repaired at that

time. The TS containment sections allow these valves to be opened for

inerting, de-inerting, and pressure control. However, with radiation

monitor 2D11-K621B inoperable, the primary containment isolation

instrumentation TS require the valves be closed until the unit achieves

a cold shutdown condition. Without the ability to open these valves

until cold shutdown, pressure control and de-inerting are difficult.

The purpose of the high drywell radiation primary containment

isolation signal is to limit fission product release following a

postulated loss-of-coolant accident (LOCA) with significant fuel

damage. It is one of several signals which isolate the primary

containment vent and purge valves. A high drywell pressure signal will

not only shut down the reactor and generate a LOCA signal, it will also

isolate these valves.

High drywell radiation indicates possible gross failure of the fuel

cladding. The generation of this isolation signal is not credited in

any accident or transient analysis. Chapter 15 of the Hatch Unit 2

Final Safety Analysis Report (FSAR) discusses the radiological

consequences of a postulated large break LOCA with fuel failure to show

conformance to 10 CFR Part 100 and 10 CFR Part 50, Appendix A. This

analysis is not affected.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

The change does not involve a significant hazards consideration

for the following reasons:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Opening the containment purge and vent valves with an inoperable

drywell radiation monitor will not increase the probability of any

previously evaluated accident. The fact that the monitor cannot send

an automatic isolation signal will not significantly affect the

consequences of an accident. The function of the primary containment

isolation signal is to detect and limit release of fission products

following significant fuel damage. The generation of this isolation

signal is not credited in any accident or transient analysis.

Chapter 15 of the Unit 2 FSAR evaluates the radiological

consequences of a postulated design basis LOCA with non-mechanistic

fuel damage. This licensing evaluation shows conformance to the

radiological limits presented in 10 CFR 100 and 10 CFR 50, Appendix

A. The results of this analysis are not affected since the valves

are otherwise operable and receive isolation signals from other

instrumentation.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change does not involve the installation of any new

equipment, or the modification of any equipment designed to prevent

or mitigate the consequences of accidents or transients. Therefore,

the change has no effect on any accident initiator, and no new or

different type of accidents are postulated to occur.

3. The proposed amendment does not result in a significant

reduction in the margin of safety.

As discussed in Item 1 above, the assumptions and results of the

licensing evaluations remain unchanged. Therefore, the margin of

safety is not significantly affected.

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:Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,

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

NRC Project Director: Herbert N. Berkow

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: February 28, 1995

Description of amendment request: Technical Specification (TS)

Section 6.5.1.12 would be revised to delete the requirement to render

determinations in writing with regard to whether or not activities

listed in TS Sections 6.5.1.2 and 6.5.1.5 constitute an unreviewed

safety question.

Basis for proposed no significant hazards consideration

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

provided its analysis of the [[Page 16189]] issue of no significant

hazards consideration, which is presented below:

. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The proposed change removes the requirement to

render determinations in writing with regard to whether or not

opposed changes to the Technical Specifications and investigations

of violations of Technical Specifications constitute an unreviewed

safety question. This change is considered an administrative change

to remove a requirement which is not relevant to these activities

and which is also consistent with the BWR Revised Standard Technical

Specifications (NUREG 1433). Existing requirements to perform

Technical and Independent Safety Reviews of these activities are not

affected. Therefore, the proposed amendment does not significantly

increase the probability of occurrence or the 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 change is considered administrative since it removes a

requirement which is not relevant to the affected activities, and

which is also consistent with the BWR Revised Standard Technical

Specifications Administrative Controls for Review and Audit.

Existing requirements to perform Technical and Independent Safety

Reviews for the affected activities are not changed. Therefore, this

change has no effect on the possibility of creating a new or

different king 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 change removes a requirement which is not

relevant to the affected activities. Existing Technical

Specification requirements to perform Technical and Independent

Safety Reviews for the affected activities are not changed and

therefore, will continue to ensure that such activities properly

address nuclear safety and safe plant operation. Therefore, it is

concluded that operation of the facility in accordance with the

proposed amendment does not involve a reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location:Ocean County Library, Reference

Department, 101 Washington Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,

Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Phillip F. McKee

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: February 15, 1995

Description of amendment request: The proposed amendment would

modify (by relocation to the Technical Requirements Manual) Technical

Specification (TS) 3/4.3.3.7, Chemical Detection Systems, and TS 3/

4.8.4.1, Electrical Equipment Protective Devices - Containment

Penetration Conductor Overcurrent Protective Devices, and the

associated Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed change to Technical Specification 3.3.3.7, Chemical

Detection Systems and 3.8.4.1, Electrical Equipment Protective

Devices-Containment Penetration Conductor Overcurrent Protective

Devices, is of an administrative nature in that the listed Technical

Specifications and Bases will be relocated in entirety to the

Technical Requirements Manual (TRM). Any future changes to the

relocated requirements will be in accordance with 10CFR 50.59 and

approved station procedures. Whether the listed Technical

Specifications and Bases are located in Technical Specifications or

the Technical Requirements Manual has no effect on the probability

or consequences of any accident previously evaluated.

The proposed change does not alter the assumptions previously

made in the listed Technical Specifications. The proposed change

allows the Commission and South Texas more effective use of

personnel resources to control requirements that meet the four

Criteria in the Final Policy Statement. The proposed change will not

change the dose to workers.

Since the probability of a [sic] accident is unaffected by the

administrative relocation of the listed Technical Specifications,

and the doses are not affected and do not exceed acceptance limits,

the proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed change to Technical Specification 3.3.3.7, Chemical

Detection Systems and 3.8.4.1, Electrical Equipment Protective

Devices-Containment Penetration Conductor Overcurrent Protective

Devices, is of an administrative nature in that the listed Technical

Specifications and Bases will be relocated in entirety to the

Technical Requirements Manual (TRM). Any future changes to the

relocated requirements will be in accordance with 10CFR 50.59 and

approved station procedures. Whether the listed Technical

Specifications and Bases are located in Technical Specifications or

the Technical Requirements Manual has no effect on any previously

evaluated accident. It does not represent a change in the

configuration or operation of the plant and, therefore, does not

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

any accident previously evaluated.

3. Does the proposed change involve a significant reduction in

the margin of safety?

The proposed change to Technical Specification 3.3.3.7, Chemical

Detection Systems and 3.8.4.1, Electrical Equipment Protective

Devices-Containment Penetration Conductor Overcurrent Protective

Devices, is of an administrative nature in that the listed Technical

Specifications and Bases will be relocated in entirety to the

Technical Requirements Manual (TRM). Any future changes to the

relocated requirements will be in accordance with 10CFR 50.59 and

approved station procedures. The margin of safety is not reduced

when the requirements are relocated to a Licensee-controlled

document because the requirements to change a License Basis Document

via the 10CFR 50.59 process ensures the same questions concerning

the margin to safety required for a License Amendment are asked. The

major difference is the time and expense required for the License

Amendments. Therefore, this proposed change does not significantly

reduce the margin of safety.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, N.W., Washington, D.C. 20036

NRC Project Director: William D. Beckner

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: February 15, 1995 [[Page 16190]]

Description of amendment request: The proposed amendment would

modify Technical Specification 4.6.2.3.a.2 (and associated Bases) to

reflect the reactor containment fan cooler flow rate assumed in the

accident analyses and to specify that this flow is provided by the

component cooling water system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed change to Technical Specification 4.6.2.3.a.2 is to

reflect the cooling water temperature assumed in the accident

analyses. The revised Technical Specification surveillance

requirement will change the cooling water flow rate requirement to

each Reactor Containment Fan Cooler from greater than or equal to

550 gallons per minute to greater than or equal to 1800 gallons per

minute.

The proposed change, which will result in an increased

acceptance criteria for the flow to the Reactor Containment Fan

Coolers, is not indicative of accident initiators. The change will

ensure that the surveillance requirement reflects the flow rate

value assumed in the South Texas Project accident analyses and that

the design and operability requirements of equipment important to

safety are ensured.

The accident mitigation features of the plant are not affected

by the proposed change since the change reflects the original

assumptions made in the design of the accident mitigation features

of the South Texas Project. Therefore, the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Does the proposed change create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed change does not create the possibility of a new or

different kind of accident previously evaluated in the Safety

Analysis Report because all the accidents were analyzed with a flow

rate of 1800 gallons per minute to the Reactor Containment Fan

Cooler.

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

margin of safety?

There will be no adverse affects on margins of safety since a

more stringent surveillance requirement will be applied to the

Reactor Containment Fan Cooler. The Technical Specification

operability and surveillance requirements are not reduced but rather

made more restrictive by this proposed change. The change ensures

that the margin of safety originally intended for the Reactor

Containment Fan Coolers is maintained.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location:Wharton County Junior College,

J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, N.W., Washington, D.C. 20036

NRC Project Director: William D. Beckner

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

Linn County, Iowa

Date of amendment request: February 13, 1995

Description of amendment request: The proposed amendment would

delete the audit frequency requirements from the Duane Arnold Energy

Center Technical Specifications (TS) and add them to the Quality

Assurance Program Description located in the Updated Final Safety

Analysis Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed amendment does not involve a change in the

probability or consequences of an accident previously evaluated. No

physical changes will occur as a result of this amendment. The

change is administrative in nature and does not impact the operation

of the plant or the plant's response to any accident. Because it

will allow management the flexibility to adjust the audit

frequencies based upon the performance of the program or

organization being audited, the overall performance of the

organization will be improved.

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

new or different kind of accident from any accident previously

evaluated. No physical changes will occur as a result of this

amendment. The change is administrative in nature and does not

affect the operation or design of the plant; therefore, it does not

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

any accident previously evaluated. The audits will continue to be

performed to provide assurance of conformance to the applicable

requirements.

(3) The proposed amendment will not reduce the margin of safety.

No physical changes will occur as a result of this amendment. The

change is administrative in nature and does not affect the operation

or design of the plant. Safety limits and limiting safety system

settings are not affected by this proposed amendment. The amendment

removes requirements for frequency of audits from the TS, thus

permitting more effective scheduling of audits based on performance

and the status of the activities audited. This should result in a

more effective audit program that will contribute to an improvement

in the overall performance of the organization.

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, N. W., Washington, D. C. 20036-5869NRC Acting

Project Director: John N. Hannon

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,

Illinois

Date of amendment request: February 10, 1995

Description of amendment request: The proposed amendment would

modify Technical Specification 3.3.2.1, ``Control Rod Block

Instrumentation,'' to revise two surveillance requirements and their

associated notes for the Rod Withdrawal Limiter (RWL) mode of the Rod

Pattern Control System. These changes will conform these requirements

to their original bases and eliminate the potential for unnecessary

power reductions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

(1) The proposed changes are consistent with the Rod Withdrawal

Error (RWE) analysis presented in Clinton Power Station (CPS)

Updated Safety Analysis Report (USAR) Section 15.4.2. The proposed

changes do not result in any change to plant equipment or operation;

only the plant conditions for which the Rod Withdrawal Limiter (RWL)

function(s) are required to be tested are being revised. The

proposed changes continue to ensure that the RWL is OPERABLE and

tested to ensure that continuous control rod withdrawals remain

within the assumptions of the RWE analyses. The proposed changes

have no impact on the probability of occurrence of a RWE event.

Therefore, the proposed changes do not result in a significant

increase in the probability or consequences of any accident

previously evaluated.

(2) The proposed changes do not result in any changes to plant

equipment or operation; only the plant conditions for which the RWL

[[Page 16191]] function(s) are required to be OPERABLE and tested

are being revised. The proposed changes continue to ensure that the

RWL is OPERABLE and tested to ensure that continuous control rod

withdrawals remain within the assumptions of the RWE analyses. As a

result, no new failure modes are introduced. The proposed changes

are clearly within the limits of plant operation as described in the

USAR and the RWE analyses. Therefore, the proposed changes cannot

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

any accident previously evaluated.

(3) The proposed changes revise the testing requirements to be

consistent with the testing required prior to Amendment No. 95. The

proposed changes ensure that the RWL is OPERABLE and tested to

ensure that continuous control rod withdrawals remain within the

assumptions of the RWE analyses. The proposed changes are clearly

within the limits of plant operation as described in the USAR and

the RWE analyses. Therefore, the proposed changes do not involve a

reduction in a margin of safety.

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

this review, it appears that the three standards of 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:Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th St., Decatur,

Illinois 62525.

NRC Acting Project Director: John N. Hannon

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine

YankeeAtomic Power Station, Lincoln County, Maine

Date of amendment request: February 14, 1995

Description of amendment request: The proposed amendment would

change responsibility for audits of the emergency and security plans

and their implementing procedures. Audit responsibility would change

from the licensee's Nuclear Safety Audit and Review (NSAR) Committee

and the Plant Operation Review Committee (PORC), to the respective

emergency and security plans. The proposed amendment is consistent with

the guidance of NRC Generic Letter 93-07, Modification of the Technical

Specification Administrative Control Requirements for Emergency and

Security Plans, dated December 28, 1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the Standards of 10 CFR 50.92(c). A summary of the licensee's

analysis is presented below:

1. The proposed amendment would not involve a significant

increase in the probability or consequences of an accident

previously analyzed.

The proposed changes do not have a direct effect on the physical

plant or the maintenance of the physical plant, but would improve the

safe operation of the plant by reducing the administrative burden of

PORC and NSAR. This change would allow a better focus of management

resources to the operational safety oversight of plant activities. The

requirement to review, audit, document, control, and submit for

regulatory review, the Emergency Plan and the Security Plan and their

implementing procedures, is defined by regulation and remains

unchanged. The proposed changes will not, of themselves, result in any

reduction in the effectiveness of either the Emergency Plan or the

Security Plan to protect the health and safety of the public. The

proposed changes, therefore, will not increase the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any previously evaluated.

This change is administrative in nature and does not change or

modify the physical plant or maintenance of the physical plant.

Applicable regulations continue to enforce the requirement for review

and audit by individuals not responsible for implementation of the

existing programs. Consequently, independent oversight of the programs

and procedures is not compromised by these proposed changes. The

possibility of a new or different accident from any previously

evaluated as a result of future changes in the implementation of the

Security or Emergency Plans is not created.

3. The proposed amendment would not involve a significant

reduction in a margin of safety.

The proposed changes will revise the administrative

responsibilities of the PORC and NSAR committees allowing a better

focus of resources on operational safety reviews. The requirements

of the applicable Federal and State regulations ensure the continued

effective oversight of the implementation of Security and Emergency

Plans. Consequently, the adoption of the proposed changes would not

involve a significant reduction in a margin of safety.

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: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, Maine 04578

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road, Brunswick, Maine 04011

NRC Project Director: Walter R. Butler

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

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

Connecticut

Date of amendment request: October 18, 1994, as supplemented

February 21, 1995.

Description of amendment request: The following changes requested

in the October 18, 1994, submittal were published in Federal Register

on November 9, 1994 (59 FR 35876). The proposed amendment would require

three Type A overall Integrated Containment Leakage Tests be conducted

at approximately equal intervals during shutdowns during each 10 year

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

period, it would be conducted during the thirteenth refueling outage

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

thirteenth refueling outage. The amendment would also change the

Containment Leakage Bases by reflecting the conditions of a proposed

exemption to 10CFR50, Appendix J, that would remove the requirement

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

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

By letter dated February 21, 1995, the licensee withdrew the action

related to conducting the third Type A test for the second 10-year

period during the thirteenth refueling outage and the reference to a

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

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

conducted when the plant is shutdown for the 10-year plant inservice

inspection. The following basis for the proposed no significant hazards

consideration determination relates to the February 21, 1995, request.

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

[[Page 16192]] consideration, which is presented below:

The proposed change does not involve a SHC because the change

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

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

containment does not exceed the maximum allowable primary

containment leakage rate at the design pressure. This assures

compliance with the dose limits of 10CFR100.

The proposed change to Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 2 Technical Specifications will increase the

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

maximum allowable leakage rate at the design containment pressure,

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

make any physical or operational changes to existing plant

structures, systems, or components.

Historically, Type A tests have a relatively low failure rate

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

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

failures of Type B or C components (containment penetrations and

isolation valves). Type B and C components are tested per

Surveillance Requirement 4.6.1.2.d of the Millstone Unit No. 2

Technical Specifications. These tests are required to be conducted

at intervals no greater than 24 months. These local leakage rate

tests provide assurance that containment integrity is maintained.

The Type B and C tests will continue to be performed in accordance

with the requirements of Surveillance Requirement 4.6.1.2.d.

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

Unit No. 2 has maintained control of containment integrity by

maintaining a conservative margin between the acceptance criterion

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

the Millstone Unit No. 2 containment is considered to be in sound

condition.

Based on the above, the proposed change to Surveillance

Requirement 4.6.1.2.a of the Millstone Unit No. 2 Technical

Specifications does not involve a significant increase in the

probability or consequences of an accident previously analyzed.

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

from any previously analyzed.

The proposed change to Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 2 Technical Specifications will increase the

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

physical or operational changes to existing plant structures,

systems, or components. In addition, the proposed change does not

modify the acceptance criteria for the Type A tests. Maintaining the

leakage through the containment boundary to the atmosphere within a

specific value ensures that the plant complies with the requirements

of 10CFR100. The containment boundary serves as an accident

mitigator; it is not an accident initiator. Therefore, the proposed

change to Surveillance Requirement 4.6.1.2.a does not create the

possibility of a new or different kind of accident from any

previously analyzed.

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

The proposed change to Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 2 Technical Specifications will increase the

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

maximum allowable leakage rate at the design containment pressure,

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

make any physical or operational changes to existing plant

structures, systems, or components.

Based on the above, the proposed change does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

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

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

06141-0270.

NRC Project Director: Phillip F. McKee

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: February 1, 1995

Description of amendment request: This Technical Specification (TS)

change would modify the applicable operational conditions for the

secondary containment isolation radiation monitors located on the

refueling floor and for the radiation monitor located in the railroad

access shaft. Specifically, for the refueling floor exhaust duct and

wall exhaust duct radiation monitors, the proposed change would modify

the applicable operational condition during specific control rod

testing evolutions which are core alterations and would indicate that

the operability requirement does not apply during shutdown margin

demonstrations. For the railroad access shaft exhaust duct radiation

monitor, the change to the TS would modify the applicable operational

condition to address plant evolutions involving irradiated fuel

transfer within the railroad accessshaft and above the access shaft

with the equipment hatch open.

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:

I. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

a. The proposed change to the applicable operational condition

for the refueling floor process radiation monitors does not affect

the probability of the design basis accidents. The monitors function

in response to an airborne radioactivity concentration in the

unfiltered air from the Zone III exhaust system and provide

isolation signals which limit offsite doses to within regulatory

limits. As such, there is no correlation between monitor operability

and accident probability. The monitors act to mitigate the offsite

effects of airborne contamination producing accidents, they are not

potential accident initiators.

The proposed change does not result in a significant increase in

the consequence of the design basis accidents. The postulated event

associated with control rod related CORE ALTERATIONS which could

result in increased Zone III airborne radioactivity concentrations

is criticality resulting from a single control rod withdrawal,

resulting in release of fission products. The probability of an

unintended criticality from a single control rod withdrawal is low,

and the potential for this criticality to result in fuel failure

under shutdown conditions is even more remote. Withdrawal of a

single control rod is an analyzed evolution during which time

adequate design and operating controls exist to preclude

criticality. However, in the unlikely event criticality should

occur, the potential offsite effects would not be significant.

Localized criticality involving a leaking rod, or criticality

induced fuel failure, are the postulated mechanisms by which an

increase in Zone III airborne radioactivity could be attained.

Neither of these postulated, but very unlikely events, will result

in radioactive release in excess of 10CFR100 limits. Any release

would be monitored by instrumentation in the Reactor Building vent

stack required to be OPERABLE at all times. In addition, Area

Radiation Monitors are installed on the refueling floor to

supplement the refueling floor process radiation monitors by

providing radiological information to plant operators. Operators can

use the vent stack and/or ARM information to manually initiate

secondary containment isolation if radiological conditions warrant

this action. Emergency Operating Procedures direct operator action

in the event of higher than normal radiation readings.

b. The proposed change to the applicable operational condition

for the railroad access shaft process radiation monitor does not

affect the probability of the design basis accidents. The monitor

functions in response to an airborne radioactivity concentration in

the unfiltered air from the Zone III exhaust [[Page 16193]] system

and provides isolation signals which limit offsite doses to within

regulatory limits. As such, there is no correlation between monitor

operability and accident probability. The monitor acts to mitigate

the offsite effects of airborne contamination producing accidents,

it is not a potential accident initiator.

The proposed change does not result in a significant increase in

the consequence of the design basis accidents. The design intent of

the railroad access shaft process radiation monitor is to monitor

radiation in the unfiltered air from the Zone III railroad access

shaft exhaust system, and provide signals which automatically

isolate the Zone III portion of the secondary containment, start the

Standby Gas Treatment System, and start the Recirculation System

(Zone III) on a high radiation condition within the access shaft.

This function is intended to limit the consequences of a fuel

handling accident in the railroad access shaft. The monitor has no

significant capability to react to a CORE ALTERATION related

transient, or one resulting from operations with the potential to

drain the reactor vessel. The design intent of the monitor is

maintained under the proposed change, as the proposed change focuses

monitor operability on conditions when irradiated fuel is in the

railroad access shaft or above it with the railroad access shaft

cover open.

For the above stated reasons, the applicable operational

condition for Refueling Floor Exhaust Duct High Radiation Monitors,

Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft

Exhaust Duct Radiation Monitor can be modified without significantly

increasing the probability or consequences of an accident previously

evaluated.

II. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The Refueling Floor Exhaust Duct High Radiation Monitors, Wall

Exhaust Duct Radiation Monitors, and the Railroad Access

ShaftExhaust Duct Radiation Monitor function in response to an

airborne radioactivity concentration in the unfiltered air from the

Zone III exhaust system and provide isolation signals which limit

offsite doses to within regulatory limits. As such, there is no

correlation between monitor operability and the potential for

creating new or different accident scenarios. The monitors act to

mitigate the offsite effects of airborne contamination producing

accidents, they are not potential accident initiators.

For the above stated reasons, the applicable operational

condition for Refueling Floor Exhaust Duct High Radiation Monitors,

Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft

Exhaust Duct Radiation Monitor can be modified without creating the

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

previously evaluated.

III. This change does not involve a significant reduction in a

margin of safety.

a. The proposed change to the applicable operational condition

for the refueling floor process radiation monitors does not involve

a significant reduction in the margin of safety. The postulated

event associated with control rod related CORE ALTERATIONS which

could result in increased Zone III airborne radioactivity

concentrations is criticality resulting from a single control rod

withdrawal under shutdown conditions. There are multiple barriers to

protect against the postulated event of criticality from a single

rod withdrawal. Technical Specifications, plant operating

procedures, and plant design control the withdrawal of control rods

to minimize the potential for an inadvertent criticality event

during shutdown. In addition, a fuel loading verification is

performed, per procedure, on the as loaded core configuration to

ensure that the fuel is loaded correctly. Each reload core is

designed such that there is at least a 99.9% probability with a 95%

confidence that the core will not be critical as a result of a

single control rod withdrawal. The safety margin associated with a

potential criticality event from a single control rod withdrawal,

under shutdown conditions, is not impacted by the proposed change.

In the unlikely event that control rod manipulations resulted in

reactor criticality, adequate protective measures are provided by

core monitoring instrumentation required to be operable in OPCON 5.

Under this scenario, assuming the inadvertent control rod withdrawal

resulted in a significant reactivity addition, the Reactor

Protection System (RPS) would respond by inserting all control rods

via the Scram function. The RPS monitors for recriticality during

OPCON 5 with SRMs (per Technical Specification Section 3.9.2), and

IRMs. The safety margin associated with RPS response to a

criticality event, under shutdown conditions, is not impacted by the

proposed change.

Assuming that a criticality did occur as a result of a single

control rod withdrawal, any increase in Zone III airborne

radioactivity from a previously failed assembly located in the

vicinity of the withdrawn control rod or a fuel rod failure

associated with the control rod withdrawal would not result in an

offsite dose exceeding regulatory limits. Assuming that criticality

occurs following core loading and verification (i.e. 20

days after shutdown), the offsite dose as a result of the release of

fission products from a single failed fuel rod would be much less

than 1% of the applicable site boundary limits. In addition, the

failure of four complete fuel assemblies (i.e. nearly equal to 300

fuel rods in the bundles surrounding the withdrawn control rod)

would not result in offsite dose exceeding the applicable regulatory

limits. Failure of more than four complete fuel assemblies due to

the withdrawal of a single control rod in OPCON 5 is not considered

credible. In fact, given the initial conditions of this event (i.e.

cold, zero power, subcritical) and the reactivity characteristics of

the fuel (i.e. negative fuel temperature reactivity coefficient) it

is very unlikely that a criticality of this nature would result in

failure of any fuel rods. Although the refueling floor process

radiation monitors would not be OPERABLE, Zone III airborne

radioactivity concentrations can be independently detected with Area

Radiation Monitors (ARMs) which are located on the refueling floor.

These monitors provide control room indication, and would alert

operators to changing radiological conditions on the refueling

floor. In addition to providing personnel notification, the ARMs act

as a supplement to the process radiation monitors in detecting

abnormal migrations of radioactive material in or from the process

streams. Operators can manually initiate secondary containment

isolation based on ARM input. The Emergency Operating Procedures

require the operators to take appropriate actions on higher than

normal radiation readings. Moreover, any airborne radioactivity

leakage from Zone III would be monitored via instrumentation in the

Reactor Building vent stack required to be OPERABLE at all times;

local alarms, remote recording, and main control room and Technical

Support Center alarms are provided. Operators can manually initiate

secondary containment isolation based on exhaust sample readings.

Due to the bounding regulatory limits and the redundant monitoring

and operator response capabilities, the safety margin associated

with the potential for offsite airborne radioactive release, under

shutdown conditions, is not significantly impacted by the proposed

change.

b. The elimination of operability requirements associated with

CORE ALTERATIONS, operations with the potential to drain the reactor

vessel, and other irradiated fuel moves not associated with the

railroad access shaft, do not affect the ability of the railroad

access shaft process radiation monitor to implement its design

function. As such, the current operability requirements for the

monitor which involve evolutions in areas other than the railroad

access shaft do not contribute to the margin of plant safety; thus

eliminating these operability requirements will not reduce the

margin of plant safety.

For the above stated reasons, the applicable operational

condition for Refueling Floor Exhaust Duct High Radiation Monitors,

Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft

Exhaust Duct Radiation Monitor can be modified without 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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz [[Page 16194]]

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: February 2, 1995

Description of amendment request: This amendment would change the

Technical Specifications for the units to increase the licensed

discharge fuel assembly exposure for SPC 9X9-2 fuel from 40 to 45 GWD/

MTU.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not:

I. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

PP&Ls technical basis for increasing the licensed discharge

exposure limit as proposed is documented in PL-NF-94-005-P-A. The

technical basis includes onsite fuel inspections, fuel design

analyses and evaluations, and an in-reactor fuel assembly extended

exposure demonstration. In response to NRC concerns on fuel failures

at higher exposures, very conservative analyses were performed for

the CRDA [control rod drive assembly] assuming very low failure

thresholds, and offsite dose calculation results were shown to be

well within regulatory limits, even at a failure threshold of 30

cal/gm. The NRC has previously reviewed and approved all of the

above information, and inspection results have met all approved

criteria.

An evaluation of FSAR [Final Safety Analysis Report] design

basis events was performed to determine the impact of the proposed

increase in fuel exposure. The LOCA [loss-of-coolant accident]

analysis performed in support of PP&Ls Power Uprate efforts

incorporated the effects of higher exposure and LHGR [linear heat

generation rate]. From a radiological release perspective, the Power

Uprate evaluations of LOCA, MSLB [main steam line break], CRDA, and

refueling accidents each bound the potential impacts of extended

exposure fuel.

Those reload analyses deemed necessary to confirm that the above

conclusions remain valid will be performed on a cycle-specific

basis.

Based on the above, the proposed action will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

II. Create the possibility of a new or different kind of

accident from any accident previously evaluated.

The proposed action will increase the residence time of fuel

within the Susquehanna reactors. The potential consequences of this

action remain solely with the fuels ability to perform within

specified limits during the increased duty, and were reviewed in I

above. All required evaluations involving fuel impacts have been

previously evaluated.

Based on the above, the proposed action cannot create the

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

previously evaluated.

III. Involve a significant reduction in a margin of safety.

The proposed action will allow increasing the licensed discharge

fuel assembly exposure limit, resulting in increases in the fuel rod

LHGR and LHGR for APRM [average power range monitor] Setpoints,

which are controlled via the Technical Specifications and the Core

Operating Limits Report.

The discussion in I. above delineates the evaluations performed

to support this action. It concludes that neither the probability

nor the consequences of events previously evaluated will be

affected. Operator performance will not be affected, because the

operators only monitor the ratio of the fuel LHGR to the fuel design

limit. No other potentially impacted safety margins have been

identified.

Based on the above, the proposed change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location:Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: February 10, 1995

Description of amendment request: The proposed amendment would

modify the Susquehanna Steam Electric Station, Unit 1 and 2 Technical

Specifications (TS) to (1) extend the allowable out-of-service times

(AOTs) for maintenance and repair and the surveillance test intervals

(STIs) between channel functional tests for the following groups of

instruments: reactor protection systems instrumentation (TS 3.3.1),

isolation actuation instrumentation (TS 3.3.2), emergency core cooling

system actuation instrumentation (TS 3.3.3), ATWS (anticipated

transient without scram) recirculation pump trip system instrumentation

(TS 3.3.4.1), end-of-cycle recirculation pump trip system

instrumentation (TS 3.3.4.2), reactor core isolation cooling system

(RCIC) actuation instrumentation (TS 3.3.5), control rod block

instrumentation (TS 3.3.6), radiation monitoring instrumentation (TS

3.3.7.1), and feedwater/main turbine trip system actuation

instrumentation (TS 3.3.90); (2) change the required actions and AOTs

for the instruments listed above to make requirements consistent with

supporting analysis in General Electric topical reports and change

additional actions required to prevent extended AOTs from resulting in

extended loss of instrument function; (3) change the required actions

and AOTs for the instruments listed above for instrumentation

associated with the ADS (automatic depressurization system),

recirculation pump trip, and pump suction lineup for HPCI (high

pressure core injection) and RCIC; (4) change applicability

requirements and required actions for the reactor vessel water level-

low, level 3 function that isolates the RHR (residual heat removal)

system shutdown cooling system so that the function is required to be

operable in operational conditions 3, 4, and 5 to prevent inadvertent

loss of reactor coolant via the RHR shutdown cooling system; (5) remove

notes in Table 3.3.2-1, 3.3.2-2, and 4.3.1-1 related to maintenance on

leak detection temperature detectors and remove the note to TS 3.3.6

for Unit 1 related to a previous relief from TS 3.0.4; and reformat,

renumber, and/or reword existing requirements to incorporate the

changes listed above.

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:

I. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed TS changes increase the AOTs and STIs for actuation

instrumentation intended to detect or mitigate accidents; establish

required actions consistent with NUREG-1433 for some instruments

that are more specific but equivalent to existing required actions;

establish new requirements to prevent inadvertent loss of reactor

coolant via the RHR Shutdown Cooling System during OPERATIONAL

CONDITIONS 3, 4 and 5; and, eliminate notes that were intended to

provide one time only exemptions from certain requirements. The

proposed changes affect only those Technical Specification

requirements that govern operability, required actions and routine

testing of plant instruments that detect or mitigate accidents. The

proposed changes do [[Page 16195]] not affect any equipment or

requirements that are assumed to be initiators of any analyzed

events. Therefore, the proposed changes will not involve an increase

in the probability of occurrence of an accident previously

evaluated.

The proposed changes will not increase the consequences of an

accident previously evaluated because the changes will not involve

any physical changes to plant systems, structures, or components

(SSC), or the manner in which these SSC are operated, maintained,

modified, tested or inspected. The proposed changes will not alter

the operation of equipment assumed to be available for the

mitigation of accidents or transients by the plant safety analysis

or licensing basis. The proposed changes extend the intervals

between required performances of routine instrument testing. The

proposed changes also modify time limits allowed for operation with

inoperable instrument channels in situations when an inoperable

instrument channel would not prevent actuation of the associated

equipment. These changes are based on the demonstrated reliability

of these instruments and are justified by the analysis in References

1 through 8 [See February 10, 1995 application]. The small increases

in the probability that the proposed changes will result in an

equipment actuation failure has been determined in References 1

through 8 [See February 10, 1995 application] to be offset by safety

benefits such as a reduction in the number of inadvertent

actuations, a reduction in wear due to excessive testing, and better

utilization of plant personnel and resources. These changes will not

allow continuous plant operation with plant conditions such that a

single failure will result in a loss of any safety function.

Proposed changes to required actions and completion times for

instrumentation associated with the ADS initiation, Recirculation

Pump Trip, and pump suction lineup for HPCI and RCIC make the

required actions and completion times consistent with NUREG-1433,

Standard Technical Specifications for General Electric Plants, BWR/

4, Revision 0 (Reference 12). These changes are also consistent with

the assumptions used in References 1 through 8 [See February 10,

1995 application]. Therefore, these changes establish or maintain

adequate assurance that components are operable when necessary for

the prevention or mitigation of accidents or transients and that

plant variables are maintained within limits necessary to satisfy

the assumptions for initial conditions in the safety analysis. In

addition, the proposed change provides the benefit of avoiding an

unnecessary shutdown transient when appropriate measures are

available to compensate for the inoperable instrumentation.

Therefore, the proposed changes will not increase the consequences

of an accident previously evaluated.

There is no significant increase in the probability or

consequences of an accident previously evaluated resulting from

changes that reformat, renumber, and/or reword existing requirements

to incorporate the changes above or from the removal of notes that

were intended for one time only use and are no longer applicable.

II. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

This proposed change will not involve any physical changes to

plant systems, structures, or components (SSC), or the manner in

which these SSC are operated, maintained, modified, tested, or

inspected. The changes in normal plant operation are consistent with

the current safety analysis assumptions. Therefore, this change will

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

from any accident previously evaluated.

III. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed TS changes: increase the AOTs and STIs for

actuation instrumentation intended to detect or mitigate accidents;

establish required actions consistent with NUREG-1433 for some

instruments that are more specific but equivalent to existing

required actions; establish new requirements to prevent inadvertent

loss of reactor coolant via the RHR Shutdown Cooling System during

Operational Conditions 3, 4 and 5; and, eliminate notes that were

intended to provide one time only exemptions from certain

requirements.

There is no significant reduction in the margin of safety

resulting from changes to the minimum surveillance test intervals

(STIs) and allowable out-of-service times (AOTs) for the testing

and/or repair of instrumentation. This conclusion is based on the

demonstrated reliability of these instruments and is justified by

the analysis in References 1 through 8 [See February 10 1995

application]. The small increases in the probability that the

proposed changes will result in an equipment actuation failure has

been determined in References 1 through 8 [See February 10, 1995

application] to be offset by safety benefits such as a reduction in

the number of inadvertent actuations, a reduction in wear due to

excessive testing.

These changes will not allow continuous plant operation with

plant conditions such that a single failure will result in a loss of

any safety function.

There is no significant reduction in the margin of safety

resulting from changes to required actions and completion times for

instrumentation associated with the ADS initiation, Recirculation

Pump Trip, and pump suction lineup for HPCI and RCIC. These changes

make the required actions and completion times consistent with

NUREG-1433, Standard Technical Specifications for General Electric

Plants, BWR/4. These changes are also consistent with the

assumptions used in References 1 through 8 [See February 10, 1995

application]. Therefore, these changes establish or maintain

adequate assurance that components are operable when necessary for

the prevention or mitigation of accidents or transients and that

plant variables are maintained within limits necessary to satisfy

the assumptions for initial conditions in the safety analysis. In

addition, the proposed change provides the benefit of avoiding an

unnecessary shutdown transient when appropriate measures are

available to compensate for the inoperable instrumentation.

Additionally, the proposed required actions ensure that actions to

mitigate loss of single failure tolerance are initiated within 24

hours (12 hours for RPS) in accordance with the results of the

analyses in References 1 through 8 [See February 10, 1995

application] and action to mitigate a loss of instrument function is

initiated within 1 hour. Therefore, these changes will not allow

continuous plant operation with plant conditions such that a single

failure will result in a loss of any safety function. The

Pennsylvania Power & Light Company performed reviews that confirmed

the analyses in References 1 through 8 [See February 10, 1995

application] are applicable to SSES and that there would be no

effect on the identification of excessive instrument setpoint drift

as a result of increasing the minimum interval between instrument

functional tests from monthly to quarterly.

There is no significant reduction in the margin of safety

resulting from changes that reformat, renumber, and/or reword

existing requirements to incorporate the changes above or from the

removal of notes that were intended for one time only use and are no

longer applicable.

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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Public Service Electric and Gas

Company,Delmarva Power and Light Company, and Atlantic City

Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom

Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: January 13, 1995

Description of amendment request: The proposed changes concern a

revision to the frequency of calibration for the Local Power Range

Monitor (LPRM) signals from every 6 weeks to every 2000 Megawatt Days

per Standard Ton (MWD/ST).

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: [[Page 16196]]

1. The proposed change does not involve a significant increase

in the probability of consequences of an accident previously

evaluated.

This change does not affect the operation of any equipment. The

change does not affect the fundamental method by which the LPRMs are

calibrated. The increased time between required LPRM calibrations

does not affect either the initiator of any accident previously

evaluated or any equipment required to mitigate the consequences of

an accident, or the isotopic inventory in the fuel. Thus, the change

does not increase either the probability or the radiological

consequences of an accident previously evaluated.

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

or different kind of accident from any previously evaluated.

The proposed change does not introduce a new mode of plant

operation and does not involve the installation of any new equipment

or modifications to the plant. Therefore, it does not create the

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The GETAB determination of the Maximum Critical Power Ratio

(MCPR) Safety Limit allows a maximum total nodal uncertainty of the

TIP readings (of which the LPRM Update uncertainty is a part) of

8.7%. The change in LPRM calibration frequency results in an LPRM

Update uncertainty of 4.2% nodal power. This, combined with the

other uncertainties which comprise the total TIP readings

uncertainty, yields a total TIP readings uncertainty of less than

the allowed 8.7%. Thus the change in LPRM calibration frequency will

not affect the MCPR Safety Limit.

The LPRMs are utilized as input to the APRM and RBM systems. The

primary safety function of the APRM system is to initiate a scram

during core-wide neutron flux transients before the actual core-wide

neutron flux level exceeds the safety analysis design basis. This

prevents fuel damage from single operator errors or equipment

malfunctions. The APRMs are calibrated at least twice per week to

the plant heat balance, utilize a radially and axially diverse group

of LPRMs as input and are utilized to detect changes in average, not

local, power changes. Therefore, the effects of decreasing the LPRM

calibration frequency on the APRM system responses will be minimal

due to any individual LPRM drift being practically canceled out (due

to diversity of input) and/or due to the frequent recalibration of

the APRMs to an independent power calculation (the heat balance).

Thus, decreasing the LPRM calibration frequency will not

significantly impact the performance of the APRM system's scram

function, and there is no impact on transient delta-CPRs.

The RBM system is utilized in the mitigation of a Rod Withdrawal

Error (RWE). The RBM system is designed to prevent the operator from

increasing the local power significantly when withdrawing a control

rod. On each selection of a control rod, the average of the

assigned, unbypassed LPRMs is adjusted to equal a 100% reference

signal for each of the two RBM channels. Each RBM channel

automatically limits the local thermal margin changes by limiting

the allowable change in local average neutron flux to the RBM

setpoint. If the local average neutron flux change is greater than

that allowed by the RBM setpoint, within either RBM channel, the rod

withdrawal permissive is removed preventing further movement. Since

the change in local neutron flux is calculated from the change in

the average of the LPRM readings, and calibrated on every rod

selection to the reference signal, offsets in individual LPRM

readings due to calibration differences are effectively eliminated

for a given RBM setpoint. Therefore, the constraints on the

withdrawal of any given rod are unchanged and there will not be any

increase in RWE delta-CPR.

Since the MCPR Safety Limit is unaffected and the delta-CPR

values are unchanged, the cycle CPR limits are unchanged. Therefore,

the change in the frequency of LPRM calibration does not result in a

reduction in a margin of safet