Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person. This biweekly notice includes

all notices of amendments issued, or proposed to be issued from May 23,

1997, through June 6, 1997. The last biweekly notice was published on

June 4, 1997 (62 FR 30629).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission

[[Page 33118]]

take this action, it will publish in the Federal Register a notice of

issuance and provide for opportunity for a hearing after issuance. The

Commission expects that the need to take this action will occur very

infrequently.

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

Review and Directives Branch, Division of Freedom of Information and

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

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

publication date and page number of this Federal Register notice.

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

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

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

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

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

petitions for leave to intervene is discussed below.

By July 18, 1997, 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 20555-0001, Attention: Docketing and

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

Date of amendment request: May 16, 1997

Description of amendment request: The modification involves

replacing the service water (SRW) heat exchangers with new plate and

frame heat exchangers having increased thermal performance capability.

The saltwater (SW) and SRW piping configuration will be modified as

necessary to allow proper fit-up to the new components. A flow control

scheme to throttle saltwater flow to the heat exchangers and the

associated bypass lines will be added.

[[Page 33119]]

Saltwater strainers with an automatic flushing arrangement will be

added upstream of each heat exchanger. The majority of the physical

work associated with this modification is restricted to the SRW pump

room.

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. Would not involve significant increase in the probability or

consequences of an accident previously evaluated.

None of the systems associated with the proposed modification

are accident initiators. The SW and SRW Systems are used to mitigate

the effects of accidents analyzed in the UFSAR [Updated Final Safety

Analysis Report]. The SW and SRW Systems provide cooling to safety-

related equipment following an accident. They support accident

mitigation functions; therefore, the proposed modification does not

increase the probability of an accident previously evaluated.

The proposed modification will increase the heat removal

capacity of the SRW System. The design provided under this activity

ensures that the safety features provided by the SW and SRW are

maintained, and in some instances enhanced; i.e., the availability

of important-to-safety equipment required to mitigate the

radiological consequences of an accident described in the UFSAR is

enhanced by the flexibility and increased thermal margin provided

with this design.

The redundant cooling capacity of the SW and SRW Systems have

not been altered. Furthermore, the proposed activity will not

change, degrade, or prevent actions described or assumed in any

accident described in the UFSAR. The proposed activity will not

alter any assumptions previously made in evaluating the radiological

consequences of any accident described in the UFSAR. Therefore, the

consequences of an accident previously evaluated in the UFSAR have

not increased.

Therefore, the proposed modification does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The proposed activity involves modifying the SW and SRW System

components necessary to support the installation of new SRW heat

exchangers. None of the systems associated with this modification

are identified as accident initiators in the UFSAR. The SW and SRW

Systems are used to mitigate the effects of accidents analyzed in

the UFSAR. None of the functions required of the SRW or SW System

have been changed by this modification. This activity does not

modify any system, structure, or component such that it could become

accident initiator, as opposed to its current role as an accident

mitigator.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

safety.

The safety design basis for the SW and SRW Systems is the

availability of sufficient cooling capacity to ensure continued

operation of equipment during normal and accident conditions. The

redundant cooling capacity of these systems, assuming a single

failure, is consistent with assumptions used in the accident

analysis.

The design, procurement, installation, and testing of the

equipment associated with the proposed modification are consistent

with the applicable codes and standards governing the original

systems, structures, and components. The design of instruments and

associated cabling ensures that physical and electrical separation

of the two subsystems is maintained. Common-mode failure is not

introduced by this activity. The equipment is qualified for the

service conditions stipulated for that environment. New cable and

raceways for this design will be installed in accordance with

seismic design requirements. The additional electrical load has been

reviewed to ensure the load limits for the vital 1E buses are not

exceeded. The circuits and components related to the control valves

control loops are safety-related, and are similar to those used for

the other safety-related flow control functions. The proposed

modification will not have any adverse effects on the safety-related

functions of the SW and SRW Systems.

For the above reasons, the existing safety bases have not been

altered by the proposed modification. This activity will not reduce

the margin of safety as it exists now. In fact, the margin of safety

has been increased by this activity due to the increase in the

thermal capacity of the dual train design and the increased

availability of safety-related components.

Therefore, this proposed modification 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 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: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

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

NRC Project Director: Alexander W. Dromerick, Acting Director.

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: April 23, 1997

Description of amendment request: The proposed changes would revise

surveillances 4.3.2.1.1.a, 4.3.2.1.4.b, 4.3.2.1.6.g, 4.3.2.1.10a,

4.3.2.1.10.b, and 4.7.3.b.3 to provide enhanced descriptions of the

tests being performed and the tested components.

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 clarification 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.

The components affected by the proposed changes are not

initiators of any accident previously evaluated. The proposed

changes to specification 4.3.2.1 items affect only the description

of the testing and make no changes in actual operation or testing.

The sample heat exchanger valves isolate on receipt of a Safety

Injection signal and that feature is unaffected by the additional

testing in the proposed change. Therefore, there is no increase in

the probability or consequence of a previously analyzed accident.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes to the surveillance frequencies do not

involve physical alterations or additions to plant equipment or

alter the manner in which safety-related systems function or are

normally operated. The additional testing proposed for the sample

heat exchanger valves demonstrates the proper operation of a design

feature but does not operate the valve in any new way. Therefore,

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

different kind of accident from any accident previously evaluated.

(3) The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed changes to specification 4.3.2.1 clarify existing

testing. The additional testing for the CCW [component cooling

water] surge tank level instrumentation adds two components to the

surveillance documentation. Therefore, there is no 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.

[[Page 33120]]

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart, Acting.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: April 14, 1997

Description of amendment request: The proposed amendments would

revise TS 3/4.3.8, ``Feedwater/Main Turbine Trip System Actuation

Instrumentation'' by changing the minimum channels required from 3 to

4. This change reflects a modification that is being installed to

correct a design deficiency that could have resulted in a failure to

trip the feedwater pumps and main turbine on high water level due to

the loss of one of the two instrument lines. The modification adds an

auxiliary contact to the trip system logic resulting in an additional

channel. The licensee is also proposing to modify the TS action

statements for inoperable channels to be similar to TS 3.3.1, ``Reactor

Protection 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) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The proposed Technical Specification (TS) change will resolve

the common instrument line failure (break) from preventing reactor

high water level trip of Feedwater Pumps and Main Turbine. It will

not change the probability of occurrence of any accidents, because

this instrumentation is not an accident initiator. This

instrumentation resolves a potential concern regarding the results

of an instrument line break in conjunction with a Feedwater

Controller Failure Maximum Demand, which has been postulated and

analyzed separately, but are not required to be analyzed in

combination, as is described in Chapter 15 of the LaSalle UFSAR.

There will not be any increase in probability of feedwater transient

(postulated feedwater controller failure with assumed simultaneous

failure of one high level trip channel of Feedwater/Main Turbine

Trip Actuation Instrumentation), nor an instrument line break. The

design change associated with this TS change will prevent the

failure of the level 8 trip of Feedwater Pumps and Main Turbine due

to loss of common variable water leg of level instrument channels

``B'' and ``C''. Thus there is a slight increase [in] the

reliability of the high level trip by assuring that a single

instrument failure, including a failure of a sensing line, will not

prevent a level 8 trip. The Feedwater/Main Turbine Trip on Reactor

Vessel Water Level-High, Level 8, mitigates the consequences of the

transient, Feedwater Controller Failure Maximum Demand, due to the

main turbine trip with subsequent Turbine Stop Valve closure scram

and Reactor Recirculation Pump Trip. This limits the neutron flux

peak and fuel thermal transients so that no fuel damage occurs. MCPR

remains at or above the operating limit and peak centerline fuel

temperature increase is small. The consequences of an accident will

not increase, because the redundancy of the instrumentation portion

of the Trip Function is somewhat increased.

TS 3.3.8 limiting Condition for Operation (LCO) Actions b and c

are proposed to be changed to be similar to the LCO for TS 3.3.1,

Reactor Protection System Action b.1 to assure trip capability,

while being consistent with the allowed outage times of current TS

3.3.8. Also, the proposed action statements and allowed outage times

are consistent with LCO 3.3.2.2, ``Feedwater and Main Turbine High

Water Level Trip Instrumentation'', of NUREG 1433, Revision 1,

Standard Technical Specifications, General Electric Plants, BWR4,

dated April 1995. The limit on continued plant operation of 72 hours

in current Action c.1, is overly restrictive, since with one

inoperable channel tripped and one Operable channel, the Trip

Function is restored to the same status as current Action b.1 (one

more instrument failure will cause a failure to actuate on high

reactor water level). Therefore, although the proposed Actions are

increasing the allowed outage time for the case with only one

remaining Operable channel, from 72 hours to 7 days, the level of

protection for automatic trip capability is maintained except for a

2 hour period during which trip capability may not exist. In

addition, like current Action b.1, the proposed Actions assure that

the longest time that automatic trip capability failure due to

another instrument failure will exist is 7 days. Therefore, the

potential for failure of the Feedwater/Main Turbine trip on reactor

vessel high water level may be slightly increased, but is not

significant considering the non-safety-related Feedwater Pump and

Main Turbine trips are not and are not required to be single-failure

proof.

Based on the above, the proposed amendments will not increase

the probability or consequences of any accident previously

evaluated.

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

accident from any accident previously evaluated because:

The Feedwater/Main Turbine trip is a non-safety function in the

non-safety-related feedwater system. The high water level trip is an

equipment protective action preventing main steam carry over in the

main steam from damaging the main turbine and preventing high

pressure liquid discharge through the safety relief valve discharge

lines in case of a feedwater transient due to a controller failure

to maximum demand. The trip system is not designed to any applicable

standards or regulatory guides or 10CFR50 Appendix A General Design

Criteria per UFSAR Table 7.1-2. The trip system is not designed nor

required to meet the single failure criteria. This is a non-safety/

non-divisional trip actuation required in Operating Condition 1, Run

Mode, such that high integrity of the trip is maintained. The

feedwater system is not required to mitigate the consequences of

accidents.

The design change associated with this TS change will increase

the reliability of the trip logic. This is accomplished by assuring

that a failure of a sensing line will not prevent or cause a level 8

trip. The failure of Feedwater/Main Turbine channel ``C'' trip

channel will not have any impact on the RCIC system nor Feedwater/

Main Turbine channels ``A'' & ``B'', because the added signal is

isolated by a safety-related relay. The 2 out of 3 logic for the

trip is maintained.

In addition, the changes to the action statements of the

specification do not allow a condition that could cause the

actuation instrumentation to fail in a different manner.

Based on the above, the proposed change will not create the

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

accident or transient previously evaluated.

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

because:

The proposed TS change will not prevent tripping of Feedwater/

Main Turbine or cause false trips. The existing 2 out of 3 logic

trip is maintained and does not affect existing failure modes or

introduce new failure modes. This change will prevent failure of

level 8 trip of Feedwater Pumps and Main Turbine upon loss of common

variable water leg for Reactor Vessel Water Level-High, Level 8,

instrument channels ``B'' & ``C'' and will slightly increase

reliability of the trip logic. Failure of the non-safety-related

trip logic will not impact any safety-related system, structure, or

component.

The changes to the TS LCO Action statements is consistent with

the existing actions, while minimizing the time that automatic trip

capability is not maintained. The change from 72 hours allowed

operation with one channel Operable and only one channel tripped to

7 days is consistent with the current allowed outage time for only

one channel inoperable and not tripped, so any change to the margin

of safety provided by the current action requirements is minor.

Based on the above, the proposed TS 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

[[Page 33121]]

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

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

Date of amendment request: May 1, 1997

Description of amendment request: This request changes Technical

Specification (TS) Surveillance Requirement (SR) 4.9.A.8.b by

clarifying the load value for the emergency diesel generator to be

equal to or greater than the largest single load and revise the

frequency and voltage requirements during performance of the test.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because of the

following:

The proposed changes represent a clarification of the intent of

the performance of the largest single emergency load rejection

surveillance for the diesel generator. These changes allow for

simulated testing that will more closely duplicate actual emergency

loading conditions. By removing the specific load value requirement

from the surveillance, the test can be performed using the actual

largest load in the same plant configuration that would exist during

an actual accident scenario. Verification of the steady-state

voltage and frequency within the required time limits provides

confidence that the diesel generator can successfully recover from

this transient. This provides greater assurance that the diesel

generator is capable of performing its intended design function

during an accident and the subsequent recovery. The changes to the

surveillance requirement will not significantly increase the

consequences of an accident previously evaluated.

The diesel generator's design function is to mitigate the

consequences of an accident by providing an independent onsite

source of alternate AC power with the capacity for operation of

systems required to shutdown the reactor and maintain it in a safe

shutdown condition until offsite power is restored. The diesel

generator and its associated subsystems are not assumed in any

safety analysis to initiate any accident sequence for Quad Cities

Station; therefore, the probability of an accident previously

evaluated is not increased by the proposed amendment.

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

accident from any accident previously evaluated because:

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

different kind of accident previously evaluated for Quad Cities

Station. The changes revise the largest single emergency load

rejection surveillance test acceptance criteria for the diesel

generator. This load rejection transient for the diesel generator is

bounded by a previously performed accident analysis. This analysis

assumes the loss of one diesel generator due to loss of 125 VDC

control power for the duration of a LOCA combined with a LOOP. The

diesel generator's design function is to mitigate the consequences

of an accident by providing an independent onsite source of

alternate AC power with the capacity for operation of systems

required to shutdown the reactor and maintain it in a safe shutdown

condition until offsite power is restored. Only one diesel generator

is required to perform this function per unit. Performance of the

Surveillance Requirement as proposed provides greater assurance that

the diesel generator is capable of performing its intended design

function during an accident and the subsequent recovery. No

significant changes to existing testing or new modes of facility

operation are proposed by this change. The proposed changes maintain

at least the present level of operability. Therefore, the proposed

changes do not create the possibility of a new or different kind of

accident from any previously evaluated.

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

because:

The proposed amendment is required to ensure the diesel

generator is tested in accordance with the design basis

requirements. The changes represent a revision to the test

acceptance criteria for performance of the largest single emergency

load rejection surveillance for the diesel generator. This is a

possible transient for the diesel generator that is bounded by a

previously performed accident analysis. The proposed changes do not

adversely affect the capability of the diesel generator to perform

its design function. This function is to mitigate the consequences

of an accident by providing an independent onsite source of

alternate AC power with the capacity for operation of systems

required to shutdown the reactor and maintain it in a safe shutdown

condition until offsite power is restored. Performance of the

Surveillance Requirement as proposed provides greater assurance that

the diesel generator is capable of performing its intended design

function during an accident and the subsequent recovery. Existing

plant safety margins or the reliability of the equipment assumed to

operate in the safety analysis are not changed. The proposed changes

have been evaluated at Quad Cities and found to be acceptable for

use based on system design, safety analysis requirements and

operational performance. Since the changes maintain the necessary

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

significant reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

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

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

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

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

Date of amendment request: May 27, 1997.

Description of amendment request: The proposed amendments would

delete from the Technical Specifications (TS) of each unit the

specified minimum volume of borated water available to the Standby

Makeup Pump; the minimum volume is already specified in other parts of

the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. This amendment to the Catawba TS maintains the necessary

minimum volume of borated water available to mitigate a design basis

SSS [standby shutdown system] event through a 72 hour period.

Eliminating TS Surveillance 4.7.13.3a.2 does not increase the

probability or consequences of any previously evaluated accident,

since an adequate borated water source for the SMP [standby makeup

pump] is continued to be required by other existing TS.

(2) Will the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

No. This amendment to the Catawba TS continues to ensure that

the necessary minimum volume of borated water is available to

mitigate an SSS event. The SSS is required to mitigate certain

previously evaluated design basis fire, security, and other events.

This amendment does not create the possibility of a new or different

kind of accident from any accident previously evaluated. This

amendment changes the TS applicable to an accident mitigating

function and does not impact any accident initiator, either new,

different, or previously evaluated.

(3) Will the change involve a significant reduction in a margin

of safety?

No. This amendment continues to ensure that the necessary

minimum volume of borated water is available to mitigate an SSS

design basis event. The available minimum volume is maintained well

above the design basis requirement. Since the source of borated

water that is available to supply the SMP continues to be controlled

by existing TS (TS 3.7.13.3a.1 and 3.9.10), which both envelope the

current 112,320 gallons, sufficient volume has been and will

continue

[[Page 33122]]

to be present to meet design basis requirements. Therefore, no

reduction in a margin of safety will result from the changes

proposed in this amendment.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730.

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

(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North

Carolina 28242-0001.

NRC Project Director: Herbert N. Berkow.

Duke Power Company, et al., Docket No. 50-414, Catawba Nuclear Station,

Unit 2, York County, South Carolina

Date of amendment request: May 27, 1997

Description of amendment request: The proposed amendment would

delete from the Technical Specification of Unit 2 requirements

regarding steam generator tube sleeving and repair. These requirements

are not applicable to the Westinghouse Model D5 steam generators used

by Unit 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. This amendment to the Catawba Unit 2 Technical

Specifications will have no impact on operation of the facility

since the change will delete steam generator repair methods that are

not applicable to the Catawba Unit 2 steam generators and have not

been used to repair the Catawba Unit 2 steam generators.

(2) Will the change create the possibility of a new or different

type of accident from any accident previously evaluated?

No. This amendment will delete steam generator repair methods

that are not applicable and have not been used. Therefore, the

proposed changes will not create the possibility of a new or

different accident.

(3) Will the change involve a significant reduction in the

margin of safety?

No. This amendment will delete steam generator repair methods

that are not applicable and have not been used. There will be no

impact on safety margins as a result of these changes.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730.

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

(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North

Carolina 28242.

NRC Project Director: Herbert N. Berkow.

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

Station, Unit 1, Claiborne County, Mississippi.

Date of amendment request: May 7, 1997.

Description of amendment request: The amendment request would

eliminate selected response time testing (RTT) surveillance

requirements (SRs) from the Technical Specifications (TSs) for certain

components of the following systems: reactor protection system (SR

3.3.1.1.15), primary containment and drywell isolation instrumentation

(SR 3.3.6.1.8), and emergency core cooling system (SRs 3.5.1.8 and

3.5.2.7).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. No significant increase in the probability or consequences of

an accident previously evaluated results from this change.

The purpose of the proposed Technical Specification (TS) change

is to eliminate response time testing (RTT) requirements for

selected components in the Reactor Protection System (RPS), Primary

Containment and Drywell Isolation Instrumentation, and Emergency

Core Cooling System (ECCS) actuation instrumentation. The Boiling

Water Reactor Owners' Group (BWROG) has completed an evaluation

which demonstrates that [RTT] is redundant to the other TS-required

testing. These other tests, in conjunction with actions taken in

response to NRC Bulletin 90-01, ``Loss of Fill-Oil in Transmitters

Manufactured by Rosemount,'' and Supplement 1 [to the bulletin], are

sufficient to identify failure modes or degradations in instrument

response time and ensure operation of the associated systems within

acceptable limits. There are no known failure modes that can be

detected by [RTT] that cannot also be detected by the other TS-

required testing. This evaluation was documented in NEDO-32291-A,

``System Analyses for Elimination of Selected Response Time Testing

Requirements,'' October 1995. EOI [The licensee] has confirmed the

applicability of this evaluation to Grand Gulf Nuclear Power Station

(GGNS). In addition, EOI will complete the actions identified in the

NRC staff's Safety Evaluation of NEDO-32291-A.

Elimination of [ECCS] RTT during MODES 4 and 5 [(i.e., cold

shutdown and refueling, respectively)] is acceptable since there are

no design basis accidents in MODES 4 and 5 for which the ECCS High

Pressure Core Spray (HPCS) system is required to initiate within a

specified period of time. The requirement to maintain [ECCS]

OPERABLE during Modes 4 and 5 is preserved in the affected Technical

Specification. The ECCS RTT required by SR 3.5.1.8 (applicable

during MODES 1, 2, and 3, [or power operation, startup, and hot

shutdown, respectively]) is adequate to identify any operability

problems with the ECCS HPCS system. In addition, during MODES 4 and

5, the probability and consequences of accidents are reduced due to

the pressure and temperature limitations of these MODES.

Because of the continued application of other TS-required tests

such as channel calibrations, channel checks, channel functional

tests, and logic system functional tests, the response time of these

systems [listed in the first paragraph] will be maintained within

the acceptance limits assumed in the plant [(GGNS)] safety analyses

and required for successful mitigation of an initiating event. The

proposed changes do not affect the capability of the associated

systems to perform their intended function within their required

response time, nor do the proposed changes themselves affect the

operation of any equipment.

As a result, EOI has concluded that the proposed changes do not

involve a significant increase in the probability or the

consequences of an accident previously evaluated.

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

different kind of accident from any [accident] previously evaluated.

The proposed changes only apply to the testing requirements for

the components [in the systems] identified above and do not result

in any physical change to these or other components [in other

systems] or their operation. As a result, no new failure modes are

introduced. Therefore, the proposed changes do not create the

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

previously evaluated.

3. This change would not involve a significant reduction in a

margin of safety.

The current TS-required response times are based on the minimum

allowable values assumed in the plant [(GGNS)] safety analyses.

These analyses conservatively establish the margin of safety. As

described above, the proposed changes do not affect the capability

of the associated systems to perform their intended function within

the allowable response time used as the basis for the plant safety

analyses. The potential failure modes for the components within the

scope of this request were evaluated for

[[Page 33123]]

impact on instrument response time. This evaluation confirmed that,

with the exception of loss of fill-oil of Rosemount transmitters,

the remaining TS-required testing is sufficient to identify failure

modes or degradations in instrument response times and ensure

operation of the instrumentation within the scope of this request is

within acceptable limits. The actions taken in response to NRC

Bulletin 90-01 and Supplement 1 [to the bulletin] are adequate to

identify loss of fill-oil failures of Rosemount transmitters. As a

result, it has been concluded that plant and system response to an

initiating event will remain in compliance with the assumptions of

the [GGNS] safety analysis. Elimination of RTT for ECCS HPCS system

in MODES 4 and 5 does not reduce the margin of safety since there

are no design basis events in MODES 4 and 5 requiring this system to

respond in [a] specified period of time from onset of the event.

Response time testing required by SR 3.5.1.8 (applicable during

MODES 1, 2, and 3) is adequate to identify any equipment or

operability concerns).

Further, although not explicitly evaluated, the proposed changes

will provide an improvement to plant safety and operation by

reducing the time safety systems are unavailable, reducing the

potential for inadvertent safety system actuation, reducing plant

shutdown risk, limiting radiation exposure to plant personnel [that

would be due to the RTT], and eliminating the diversion of key

personnel resources to conduct unnecessary testing. Therefore, EOI

concluded that this request will result in an overall increase in

the margin of safety.

[Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.]

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

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

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

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

NRC Project Director: William D. Beckner.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: May 24, 1997.

Description of amendment request: The proposed amendment will

modify Technical Specification (TS) 3/4.7.4, Ultimate Heat Sink (UHS),

Table 3.7-3, by incorporating more restrictive dry cooling tower (DCT)

fan requirements, and it will change the wet cooling tower water

consumption in the TS Bases. This proposed amendment seeks to modify

the TS to be consistent with revised design basis calculations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The proposed change modifies the UHS TS by not allowing

operation with less than 12 DCT fans per DCT. This change is

necessary to adequately preserve the assumptions and limits of the

revised UHS design basis calculations. These calculations conclude

that the UHS is capable of dissipating the maximum peak heat load

resulting from the limiting design bases accident (i.e., large break

LOCA [large break loss of coolant accident]). The proposed change

does not directly affect any material condition of the plant that

could directly contribute to causing an accident or that could

contribute to the consequences of an accident. The proposed change

ensures that the mitigating effects of the UHS will be consistent

with the design basis analysis. Therefore, the proposed change will

not involve a significant increase in the probability or

consequences of any accident previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

The proposed change modifies the UHS TS to be consistent with

revised design basis calculations. The UHS TS is being modified to

eliminate operation with less than 12 DCT fans per DCT. The proposed

change will not alter the operation of the plant or the manner in

which the plant is operated. Therefore, the proposed change will not

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

any accident previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The proposed change modifies the UHS TS by not allowing

operation with less than 12 DCT fans per DCT. The proposed change

preserves the margin of safety by ensuring that the UHS will be

capable of dissipating the maximum design basis accident heat load

with adequate margin. Therefore, the proposed change will not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: May 24, 1997

Description of amendment request: The proposed amendment will

modify Technical Specifications (TS) 3.1.1.1, 3.1.1.2, 3.10.1 and

Figure 3.1-1 by removing the cycle dependent boron concentration and

boration flow rate from the Action Statements and removing the ``RWSP

at 1720 ppm'' curve from the figure. A change to TS Bases 3/4.1.1.1 and

3/4.1.1.2 has been included to support 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. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The Shutdown Margin requirements are determined by the reload

analysis performed every cycle. The Cycle 9 reload analysis has

determined that the current Shutdown Margin requirements are

acceptable. The proposed change eliminates the reference to 1720 ppm

in the Action Statement because 1720 is not adequate to ensure that

the Shutdown Margin requirements are met at the beginning of cycle.

The proposed Action Statement will continue to ensure that in the

event the Shutdown Margin requirements are not met, boration will be

immediately initiated to restore the Shutdown Margin to within

limits.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

The proposed change does not change the design or configuration

of the plant nor does it change how boration systems are operated

during normal or accident conditions. It

[[Page 33124]]

ensures that the Shutdown Margin requirements for accidents already

evaluated are promptly restored in the event that the requirements

are not met.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The proposed change has not decreased the amount of Shutdown

Margin required. The current Shutdown Margin requirements have been

validated by the Reload Analysis for Cycle 9 and are adequate to

ensure that the reactor can be made subcritical from all operating

conditions, transients, and design basis events. The proposed change

ensures that the Shutdown Margin requirements are promptly restored

in the event that they are not met. As such, the proposed change

ensures that the current margin of safety is maintained.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 3, 1997

Description of amendment request: The proposed amendment requests a

change to the ACTION Requirements for Technical Specification 3/4.3.2

for the Safety Injection System Sump Recirculation Actuation Signal

(RAS). The proposed change will revise the allowed outage time for a

channel of RAS to be in the tripped condition from ``prior to entry

into the applicable MODE(S) following the next COLD SHUTDOWN'' to the

more restrictive time limit of 48 hours and adds a shutdown

requirement. Additionally, the 3.0.4 exemption is being removed from

the ACTION for the tripped condition. A change to the Technical

Specification Basis Section 3/4.3.2 has also been included.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The proposed revision to the TS changes the allowed outage time

that a channel of RAS can be in the tripped condition from a maximum

of approximately 18 months when one channel is inoperable and 92

days when two channels are inoperable to 48 hours. If a channel were

in the tripped condition and a single failure occurred (that of one

other channel of RAS), a premature [refueling water storage pool]

RWSP low level signal would be generated. During a Design Basis

Accident with a containment high pressure condition causing the RWSP

outlet check valves to seat, this single failure would prevent the

contents of the RWSP from being injected into the reactor coolant

system and possibly resulting in failure of both trains of

[Emergency Core Cooling System] ECCS and [Containment Spray] CS.

Additionally, this would cause the [Low Pressure Safety Injection]

LPSI pumps to stop. Reducing the time that a channel of RAS can be

placed in the tripped condition will reduce the probability of this

scenario occurring during a Design Basis Accident. Since the allowed

outage time for a channel of RAS is being limited to 48 hours, this

is considered an off-normal operation and a single failure is not

required to be postulated during a Design Basis Accident in the

accident analysis. Reducing the time the channel can be placed in

the tripped condition and thus, the exposure time to this scenario,

would not be an accident initiator. The proposed change of being

more conservative in the time and condition limits in the TS will

not affect the assumptions, design parameters, or results of any

accident previously evaluated.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

The proposed change does not change the design or configuration

of the plant. The proposed change provides a more conservative

allowed outage time for the channel to be in the tripped condition.

There has been no physical change to plant systems, structures or

components nor will the proposed change reduce the ability of any of

the safety-related equipment required to mitigate Anticipated

Operational Occurrences or accidents. In fact, this change will

potentially increase the ability of safety related equipment to

perform its functions. The configuration required by the proposed

specification is permitted by the existing specification.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The proposed change provides a more conservative allowed outage

time for the channel to be in the tripped condition. By reducing the

allowed outage time, the probability is reduced that a single

failure (that of a failure of one channel of RAS with one channel in

the tripped condition) would occur that would cause the suction to

be prematurely supplied by the Safety Injection System Sump,

potentially disabling the [High Pressure Safety Injection] HPSI and

CS pumps, and stopping of the LPSI pumps. Therefore, the only change

to the margin of safety would be an increase. Since the allowed

outage time for a channel of RAS is being limited to 48 hours, this

is considered an off-normal operation and a single failure is not

required to be postulated during a Design Basis Accident in the

accident analysis. The proposed changes do not affect the limiting

conditions for operation or their bases.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner.

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

County, Iowa

Date of amendment request: May 9, 1997

Description of amendment request: The proposed amendment would

revise the definitions of Limiting Safety System Setting (LSSS) and

Instrument/Channel Calibration to reference a new program being added

to the Technical Specification (TS) (Section 6.13) for the control of

instrument setpoints.

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 33125]]

consideration, which is presented below:

1. The proposed TS amendment will not significantly increase the

probability or consequences of any previously-evaluated accidents.

The proposed changes will not result in any direct hardware

changes. The change only adds a program to the TS for the

establishment and control of instrumentation setpoints that is

consistent with current DAEC [Duane Arnold Energy Center] practice.

The Instrument Setpoint Control Program is based upon a methodology

for the calculation of instrument setpoints that conforms to the

guidelines of Regulatory Guide 1.105, Rev. 2. The methodology

ensures that adequate margin exists between the normal plant

operating conditions and actual instrument setpoints to preclude

spurious plant/equipment trips. As a result, the proposed program

establishes the criteria for changes in instrument setpoints to

ensure that such changes will not result in unnecessary plant

transients. Consequently, the probability of any previously-analyzed

event is not increased by this change.

The role of the instrumentation and their associated setpoints

is in detecting and mitigating plant events and thereby limiting the

consequences of any previously-analyzed event. The LSSS[NTSP] and

corresponding LTPO[AV] have been developed in accordance with the

DAEC Instrument Setpoint Control Program criteria to ensure that the

instrumentation remains capable of mitigating events as described in

the safety analyses and that the results and consequences described

in the safety analyses remain bounding. Therefore, these changes do

not involve a significant increase in the consequences of an

accident previously evaluated.

2. The proposed changes will not create a new or different kind

of accident from those previously evaluated.

The proposed changes will not change the method or manner of

plant operation, in particular, calibration of TS-required

instrumentation. The use of the proposed TS program for the control

of changes to instrument setpoints does not impact safe operation of

the DAEC in that the design and safety analysis limits will continue

to be satisfied. The proposed TS program involves no system

additions or physical modifications, other than setpoint changes.

Any setpoint changes must conform to the criteria set forth in the

TS Instrument Setpoint Control Program. The instrument setpoints are

developed using a methodology that conforms to the guidelines

contained in Regulatory Guide 1.105, Rev. 2 to ensure the affected

instrumentation remains capable of mitigating accidents and

transients. Since operational methods remain unchanged and the

instrument setpoints have been evaluated to maintain the plant

within existing design basis criteria, no new or different type of

accident is created.

3. The proposed change will not result in a significant

reduction in any margin of safety.

The proposed TS program establishes the DAEC Instrument Setpoint

Control Program, which is based upon an NRC-approved methodology.

The program establishes the controls and criteria used to establish

and revise instrument setpoints. The setpoint calculations use the

uncertainties associated with the DAEC instrumentation and actual

DAEC physical data and operating practices to ensure the validity of

the resulting LTPO[AV] and LSSS[NTSP]. The methodology is based upon

combining the uncertainties of the associated channels and takes

into account calibration accuracy, instrument uncertainties, drift,

etc. The use of this methodology for establishing these setpoints

ensures that the design and/or safety analysis limits are not

exceeded in any transient or accident. Therefore, 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: Cedar Rapids Public Library,

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

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

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

NRC Project Director: Gail H. Marcus.

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

County, Iowa

Date of amendment request: May 9, 1997

Description of amendment request: The proposed amendment would

revise the definition of Limiting Condition for Operation (LCO) to

address the situation when systems, components, etc., are removed from

service or otherwise made inoperable during secondary modes of

operation, without requiring entry into the LCO actions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed TS amendment will not significantly increase the

probability or consequences of any previously evaluated accidents.

The proposed change merely adds criteria to the TS that are

consistent with the original design and licensing basis assumptions.

Operation in secondary modes of operation (such as surveillance

testing, torus cooling mode (test line-up) or Residual Heat Removal

system, and use of High Pressure Coolant Injection system or Reactor

Core Isolation Cooling system in test line-up for reactor pressure

control during transients) is assumed in the safety analysis report

(Ref. UFSAR Section 6.3.4.2.1 and 7.3.4.2). Because no changes in

actual equipment operation or testing are being made as part of this

change, the probability of any event which could be induced by such

operation or testing is not increased. Also, the change will ensure

that the time such equipment is removed from service is kept very

short in duration, either through existing TS Allowed Outage Time

(AOT) notes or administratively by procedures. This is consistent

with the assumption that the time in such secondary modes of

operation (i.e., safe test interval) is much shorter than the

allowable repair time (i.e., LCO time). Therefore, the proposed

change will not significantly increase the probability of any

previously evaluated accident.

The uniform application of the new TS criteria will further

ensure that the plant remains within the original design and

licensing basis assumptions for equipment removed from service

during secondary modes of operation. In particular, in the special

case where testing also removes the redundant system, train,

component, etc., from service, these criteria ensure that both

affected systems, trains, etc., are properly controlled. This is

acceptable because the time in such secondary modes of operation is

very short in duration, such that the impact on the overall

availability/reliability is insignificant. Therefore, the

consequences of any previously analyzed accident are not

significantly increased by this change.

2. The proposed changes will not create a new or different kind

of accident from those previously evaluated.

The proposed changes will not add a new or different kind of

accident because the plant will not be operated in a different way.

Operation in secondary modes has been previously evaluated and found

to be acceptable (Ref. General Electric reports APED-5736: Guideline

for Determining Safe Test Intervals and Repair Times for Engineered

Safeguards, and NEDO-10739: Methods for Calculating Safe Test

Intervals and Allowable Repair Times for Engineered Safeguard

Systems). The proposed change merely adds criteria to the TS that

are consistent with the assumptions contained within these

evaluations. Consequently, no new or different accidents are

postulated as a result of this proposed change.

3. The proposed change will not result in a significant

reduction in any margin of safety.

Because the criteria being added to the TS enforce the

assumptions of the evaluations that form the basis of the existing

TS (Ref. TS Bases 4.1, 4.2, and 3.5), the proposed change will not

result in a significant reduction in any 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: Cedar Rapids Public Library,

[[Page 33126]]

00 First Street, SE., Cedar Rapids, Iowa 52401.

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

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

NRC Project Director: Gail H. Marcus.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment requests: December 20, 1996

Description of amendment requests: The proposed amendments would

reduce the frequency and scope of reactor coolant pump flywheel

inspections.

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:

We have evaluated the proposed T/S changes and have determined

they do not represent a significant hazards consideration based on

the criteria established in 10 CFR 50.92(c). Operation of Cook

Nuclear Plant in accordance with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

This change will reduce the frequency and scope of the

surveillance testing on the reactor coolant pump flywheels.

Operating power plants have been inspecting their flywheels for over

20 years with no flaws identified which affect flywheel integrity.

Past examinations performed to satisfy T/S 4.4.10.1 have not

revealed any cracking of flywheel plates at Cook Nuclear Plant.

Crack extension over a 60 year service life is negligible.

Structural reliability studies have shown that eliminating

inspections after 10 years of plant life will not significantly

change the probability of failure. Most flaws which could lead to

failure would be detected during preservice inspection or, at worst,

early in plant life, and crack growth over plant life is negligible.

As stated in the SER associated with WCAP-14535, assuming an initial

crack of 10% of the distance from the keyway to the flywheel outer

radius and a maximum fatigue crack growth, ASME margins would be

maintained during the 10-year inspection period. Therefore, the

change in test frequency will not endanger public health or safety.

For these reasons, it is our belief the proposed changes do not

involve a significant increase in the probability or consequences of

a previously evaluated accident.

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

from any accident previously evaluated.

The changes will not introduce any new modes of plant operation,

nor will any physical changes to the plant be required. Thus, the

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

of accident from any accident previously analyzed or evaluated.

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

This change will reduce the frequency and scope of the

surveillance testing on the reactor coolant pump flywheels.

Operating power plants have been inspecting their flywheels for over

20 years with no flaws identified which affect flywheel integrity.

Past examinations performed to satisfy T/S 4.4.10.1 have not

revealed any cracking of flywheel plates at Cook Nuclear Plant.

Crack extension over a 60 year service life is negligible.

Structural reliability studies have shown that eliminating

inspections after 10 years of plant life will not significantly

change the probability of failure. Most flaws which could lead to

failure would be detected during preservice inspection or at worst

early in plant life, and crack growth over plant life is negligible.

As stated in the SER associated with WCAP-14535, assuming an initial

crack of 10% of the distance from the keyway to the flywheel outer

radius and a maximum fatigue crack growth, ASME margins would be

maintained during the 10-year inspection period. For these reasons,

it is our belief the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Gail H. Marcus.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: April 30, 1997

Description of amendment request: The proposed amendment would

remove Technical Specifications (TSs) regarding meteorological

monitoring instrumentation in accordance with NRC Generic Letter (GL)

95-10, ``Relocation of Selected Technical Specification Requirements

Related to Instrumentation.'' Specifically, the amendment would delete

TS 3/4.3.7.3, ``Meteorological Monitoring Instrumentation,'' including

associated TS Tables 3/4.3.7.3-1, and TS Bases 3/4.3.7.3. The TS Index

would be revised to show these deletions. The deletion of TS 3.3.7.3

would also eliminate the requirement that a Special Report to be

submitted to the NRC pursuant to TS 6.9.2 when one or more

meteorological monitoring instrumentation channels is inoperable for

more than 7 days. The licensee states that the deleted requirements

would be relocated to the Updated Safety Analysis Report (USAR), except

that the special reporting requirement would be discontinued as the

licensee would continue to evaluate future inoperability of

meteorological instrumentation for reportability in accordance with 10

CFR 50.72 and 10 CFR 50.73. The licensee will also insert the word

``nominal'' in the relocated tables in the USAR to indicate that the

meteorological instrumentation elevations of 30 and 200 feet are

nominal elevations (this change would be made because, as the licensee

reported in LER 96-14, the actual locations of the air temperature

monitoring instruments are 26.8 feet and 194.8 feet and the actual

locations of the wind indicator (speed and direction) monitoring

instruments are 30.9 feet and 199.4 feet). As stated in GL 95-10, the

NRC staff has determined that meteorological monitoring instrumentation

does not serve such a primary protective function as to warrant

inclusion in the TS in accordance with 10 CFR 50.36 criteria. Thus, in

GL 95-10, the NRC staff established that relocation of the

meteorological instrumentation requirements to the USAR (whereby

changes are controlled by the licensee pursuant to 10 CFR 50.59) is

acceptable.

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 operation of Nine Mile Point Unit 2 [NMP2], in accordance

with the proposed amendment, will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The NMP2 meteorological monitoring instrumentation is used to

provide data for use in radioactive dose assessment with respect to

routine or accidental releases of radioactive materials to the

atmosphere. The deletion of the special reporting requirements is an

administrative change. The subject special reporting requirements

serve no nuclear related protective function. The relocation of the

meteorological monitoring instrumentation requirements from the TSs

to the USAR, and the addition of the word nominal to the USAR and

tables, will not increase the probability of an accident since the

specification applies only to monitoring instrumentation. This also

is an administrative change and does not reduce the effectiveness of

the current instrumentation requirements. The meteorological

monitoring instrumentation

[[Page 33127]]

requirements are not precursors to any accident previously

evaluated. According to the NRC Staff (GL 95-10), the meteorological

monitoring instrumentation does not serve to ensure the plant is

operated within the bounds of initial conditions assumed in any

design basis accidents or transients previously evaluated, or that

the plant will be operated to preclude transients or accidents. In

addition, the meteorological monitoring instrumentation does not

function as part of the primary success path of a safety sequence

analysis used to demonstrate that the consequences of these events

are within the appropriate acceptance criteria. Therefore, the

proposed changes do not significantly increase the probability or

consequences of an accident previously evaluated.

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

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

different kind of accident from any previously evaluated.

The proposed deletion of the special reporting requirements is

an administrative change. The subject special reporting requirements

serve no nuclear related protective function. The proposed change

also removes meteorological monitoring instrumentation

specifications from the NMP2 TSs. This also is an administrative

change and does not reduce the effectiveness of the current

instrumentation requirements. The relocation of the meteorological

instrumentation requirements to the USAR, and the addition of the

word nominal to the USAR and tables, will not create the possibility

of a new or different kind of accident since the specification only

applies to monitoring instrumentation. The NRC Staff has concluded

in GL 95-10 that the provisions of the meteorological monitoring

instrumentation specifications are not related to dominant

contributors to plant risk. The NMP2 meteorological instrumentation

is used to provide data for use in radioactive dose assessment with

respect to routine or accidental releases of radioactive materials

to the atmosphere. Since no physical modification to the plant is

being performed, and no changes to actual plant operations are

required by the change, removal of the specifications from the NMP2

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

accident from any previously evaluated.

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

the proposed amendment, will not involve a significant reduction in

a margin of safety.

The proposed deletion of the special reporting requirements is

an administrative change. The subject special reporting requirements

serve no nuclear related protective function. The proposed removal

of the instrumentation requirements from the NMP2 TSs is also an

administrative change and does not reduce the effectiveness of the

current instrumentation requirements. The relocation of the

meteorological instrumentation requirements to the USAR, and the

addition of the word nominal to the USAR and tables, will not

involve a reduction in a margin of safety since the specification

only applies to monitoring instrumentation. The instrumentation will

continue to meet the requirements of Regulatory Guide 1.23, and the

offsite dose calculations will continue to use the actual measured

elevation differences. In GL 95-10, the NRC Staff concluded (1) That

the meteorological monitoring instrumentation does not function as

part of the primary success path of a safety sequence analysis, and

(2) that the meteorological monitoring instrumentation

specifications are not related to dominant contributors to plant

risk. Therefore, the removal of the meteorological monitoring

instrumentation specifications from the NMP2 TSs will not result in

a significant reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Alexander W. Dromerick, Acting Director.

Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone

Nuclear Power Station, Unit No. 1, New London County, Connecticut

Date of amendment request: May 15, 1997

Description of amendment request: The proposed amendment would

revise Technical Specification Sections 3.1 and 4.1 ``Reactor

Protection System'' and the associated Bases to remove run mode

intermediate range monitor high flux/inoperative with the associated

average power range monitor downscale scram trip function and

incorporate editorial revisions.

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 operation of Millstone Nuclear Power Station, Unit No. 1,

in accordance with the proposed amendment, will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

No physical change is being made to any systems or components

that are credited in the safety analysis, therefore there is no

change in the probability or consequences of any accident analyzed

in the UFSAR [Updated Final Safety Analysis Report].

The design basis accident applicable to the startup power region

is the Control Rod Drop Accident (CRDA). The UFSAR does not credit

the RUN Mode IRM [intermediate range monitor] High Flux/Inoperative

with the associated APRM [average power range monitor] downscale

scram Trip Function (IRM RUN Mode SCRAM) in the termination of this

accident. Accident mitigation is provided by the APRM 120% power

scram. Therefore, elimination of the IRM RUN Mode SCRAM function has

no adverse affect on previously evaluated accidents.

The Continuous Control Rod Withdrawal Error (CWE) transient is

terminated by the Rod Block Monitor (RBM) in the RUN Mode. The APRM

Reduced High Flux Scram provides the primary STARTUP Mode protection

in conjunction with the IRMs and limits the consequences of this

transient. Therefore, elimination of the IRM RUN Mode SCRAM function

has no effect on the consequences of this transient.

Clarification of the LCO [limiting condition for operation] RPS

[reactor protection system] Table aligns requirements with Limiting

Safety System Settings. Further revisions to LCO 3.1 Reactor

Protection System Table 3.1.1 and associated TS [technical

specification] bases to clarify APRM Trip Functions do not alter the

required trip functions. Deletion of RUN requirement and associated

Action B for Reduced High Flux fixes an editorial error introduced

in a previous amendment. This trip function is not effective with

the mode switch in the RUN position and removal does not alter the

neutron monitoring requirements credited in the accident analyses.

Adding a new surveillance to verify SRM [source range monitor]/

IRM/APRM overlap will enhance neutron monitoring during startups and

shutdowns and does not have an adverse affect on previously

evaluated accidents.

None of the proposed changes will affect any of the rod blocks

or other precursor events to either the CRDA or CWE. Therefore,

there is no change in the probability of any accident previously

analyzed.

2. The operation of Millstone Nuclear Power Station, Unit No. 1,

in accordance with the proposed amendment, will not create the

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

previously evaluated.

The proposed changes affect only the operations of neutron

monitoring and protective systems (IRM and APRM) which provide

indication and mitigation actions only. Operation of these systems

does not create the possibility for new precursors (such as

reactivity) which would introduce a new or different kind of

accident from any accident previously evaluated.

Additionally, the proposed changes do not affect the ability of

those systems required to mitigate previously evaluated accidents

during the modes they are credited.

3. The operation of Millstone Nuclear Power Station, Unit No. 1,

in accordance with the proposed amendment, will not involve a

significant reduction in a margin of safety.

The only scram function that the UFSAR takes credit for in the

mitigation of the limiting accident (control rod drop accident) is

the APRM 120% power scram which is not

[[Page 33128]]

affected by this change. Only the IRM RUN Mode SCRAM, for which the

UFSAR takes no credit in the termination of any analyzed event, is

removed by this change. Removal of the IRM RUN Mode SCRAM will avoid

the need to operate the plant in a ``half scram'' condition with the

potential for an inadvertent plant transient. For these reasons, the

change does not involve a significant reduction in a margin of

safety.

The Continuous Control Rod Withdrawal Error (CWE) transient is

terminated by the Rod Block Monitor (RBM) in the RUN Mode. When

initiated from the STARTUP Mode, the consequences of a CWE are

limited by the APRM Reduced High Flux scram in conjunction with the

IRM scram function. Therefore eliminating the TS requirement for the

IRM RUN Mode SCRAM will not reduce the margin of safety for this

transient.

Clarification of the LCO RPS Table aligns requirements with

Limiting Safety System Settings. Further revisions to LCO 3.1

Reactor Protection System Table 3.1.1 and associated TS bases to

clarify APRM Trip Functions do not alter the required trip

functions. Deletion of the RUN requirement and associated Action B

for Reduced High Flux corrects an editorial error introduced in a

previous amendment. This trip function is not effective with the

mode switch in the RUN position and removal does not alter the

neutron monitoring requirements credited in the accident analyses.

Adding a new surveillance to verify SRM/IRM/APRM overlap will

enhance neutron monitoring during startups and shutdowns and

consequently 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: Learning Resources Center,

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

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385.

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

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

CT 06141-0270.

NRC Deputy Director: Phillip F. McKee.

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: May 20, 1997

Description of amendment request: This submittal supersedes the

January 22, 1996, submittal which was previously noticed on February

28, 1996 (61 FR 7554). The proposed change would relocate the

containment isolation valve (CIV) list, Table 3.6-2, from the Technical

Specifications to the Technical Requirements Manual (TRM). This change

would affect Technical Specification Sections 1.8.1.b, 4.6.1.1.a,

3.6.3.1, 4.6.3.1.1, and 4.6.3.1.2, and Basis Section 3/4.6.3. A note at

the bottom of Table 3.6-2 regarding the CIVs that are subject to

administrative controls is retained in the Technical Specifications by

relocating it to Sections 1.8.1.b and 3.6.3.1. This change is being

performed in accordance with Generic Letter 91-08, which provides

guidance for removal of component lists from the Technical

Specifications.

Additionally, a change to provide relief in the surveillance

requirement in Section 4.6.1.1.a is included. The change allows valves,

blind flanges, and deactivated automatic valves located inside the

containment and are locked, sealed, or otherwise secured in the closed

position to be verified closed prior to entering Mode 4 from Mode 5, if

not performed within the previous 92 days. The current requirements

check the valve position once per 31 days.

TS Bases Section 3/4.6.3 is updated to reflect the removal and

relocation of the CIV list to the TRM. Also, details of the

administrative controls for operating CIVs while in Modes 1 through 4

are added to Bases Section 3/4.6.3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to relocate the containment isolation valve

(CIV) list will not result in any hardware or equipment operating

changes. The proposed change is based on Generic Letter (GL) 91-08

and merely relocates the CIV table and removes all references to the

table. The relocation of the CIV table from the Technical

Specifications does not affect the operability requirements of any

of the listed valves. Technical Specifications will still continue

to require the CIVs to be operable. The LCO [limiting condition for

operation] and surveillance requirements for the valves will remain

in Technical Specifications. The CIV table will be relocated to the

Millstone Unit No. 2 Technical Requirements Manual (TRM), which is

controlled in accordance with 10 CFR 50.59. This change does not

alter the design, function, or operation of the valves involved.

Thus, there is no significant affect on the possibility or

consequences of any previously evaluated accident.

The change to Surveillance Requirement (SR) 4.6.1.1.a will allow

the valves, blind flanges and deactivated automatic valves located

inside the containment that are locked, sealed, or otherwise secured

in the closed position to be verified closed prior to entering Mode

4 from Mode 5, if not performed within the previous 92 days, instead

of the current 31 day requirement. This means that the surveillance

interval could be as long as the entire operating cycle, depending

on whether entry into Mode 5 is required during the cycle. The

change in the surveillance frequency (increase in time from 31 days

to not less than 92 days and only prior to entering Mode 4 from Mode

5) recognizes that these valves are operated under administrative

controls and probability of misalignment is low. This provides

adequate assurance that the containment function assumed in the

accident analysis will be maintained. Therefore, there is no

significant affect on the probability or consequences of any

previously evaluated accident. This proposed change to SR 4.6.1.1.a

is consistent with NUREG-1432 Standard Technical Specifications for

Combustion Engineering Pressurized Water Reactors Revision 1 (SR

3.6.3.4).

The information added to the Bases will provide additional

guidance to ensure the plant is operated correctly. This information

will not result in any new approaches to plant operation. Therefore,

there is not significant affect on the probability or consequences

of any previously evaluated accident.

These proposed changes do not alter the design, function, or

operation of the valves involved. Therefore, there is no significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The change to relocate the CIV list from the Technical

Specifications to the TRM will not impose any different operational

or surveillance requirements, nor will the change remove any such

requirements. Adequate control will be maintained. Furthermore, as

stated above, the proposed change does not alter the design,

function, or operation of the valves involved, and therefore does

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

from any accident previously evaluated.

The change to SR 4.6.1.1.a reduces the surveillance frequency

for valves, blind flanges and deactivated automatic valves located

inside the containment. It does not alter the design, function, or

operation of the valves. Therefore, it does not create the

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

previously evaluated.

The information added to the Bases will provide additional

guidance to ensure the plant is operated correctly. This information

does not alter the design, function, or operation of the valves

involved. Therefore, it does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

The proposed changes will not reduce the margin of safety since

they have no impact on any safety analysis assumption. The proposed

changes do not decrease the scope

[[Page 33129]]

of equipment currently required to be operable or subject to

surveillance testing, nor do the proposed changes affect any

instrument setpoints or equipment safety functions.

The effectiveness of Technical Specifications will be maintained

since the change will not alter function or operability requirements

for any CIV. In addition, the relocation of the valve list is

consistent with the guidance provided in GL 91-08, and the change to

the surveillance interval is consistent with NUREG-0212 Standard

Technical Specifications for Combustion Engineering Pressurized

Water Reactors Revision 2 (LCO 3.6.1.1) and NUREG-1432 Standard

Technical Specifications for Combustion Engineering Pressurized

Water Reactors Revision 1 (LCO 3.6.3).

The information added to the Bases is consistent with the

guidance provided in GL 91-08 and NUREG-1432 Standard Technical

Specifications for Combustion Engineering Pressurized Water Reactors

Revision 1. The intent of the Technical Specifications will be met

since this information will not result in any new approaches to

plant operation.

Therefore, there is no 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: Learning Resources Center,

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

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385.

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

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

CT 06141-0270.

NRC Deputy Director: Phillip F. McKee.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,

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

Connecticut

Date of amendment request: May 9, 1997

Description of amendment request: The proposed amendment would

revise the shutdown margin requirements and add Technical Specification

3/4.3.5 to provide the limiting condition for operation (LCO) and

surveillance requirements for the shutdown margin monitors. The

proposed amendment would also make administrative changes and revise

the associated Bases section.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed changes in accordance with 10

CFR 50.92 and has concluded that the change does not involve a

significant hazards consideration (SHC). The bases for this

conclusion is that the three criteria of 10 CFR 50.92(c) are not

satisfied. The proposed changes do not involve [an] SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed Technical Specification changes will revise the

current shutdown margin requirements for Modes 3, 4 and 5 in Figures

3.1-1, 3.1-2, 3.1-3, 3.1-4 and 3.1-5 and allow for additional

boration of the RCS [reactor coolant system] as directed by

Specification 3.3.5. The new Shutdown Margin requirements are based

on re-analyses of the Boron Dilution Event provided by Westinghouse.

In the re-analyses, assumptions were modified in order to justify

the operability of the Shutdown Margin Monitor for count rates which

are lower than currently allowed. The proposed Shutdown Margin

requirements for Modes 3, 4 and 5 will continue to assure that the

operator has a minimum of 15 minutes from the alarm to loss of

shutdown margin during an assumed Boron Dilution Event.

The proposed change also adds Technical Specification 3/4.3.5 to

provide the LCO and Surveillance Requirements for the Shutdown

Margin Monitors. LCO 3.3.5 refers to the Core Operating Limits

Report (COLR) which will specify the minimum count rate/alarm ratio

requirements in order to consider the Shutdown Margin Monitors

operable. The LCO also directs the additional boration of the RCS in

order to allow the Shutdown Margin Monitors to be considered

operable for lower count rates. Also, a footnote (**) is included in

Specification 3/4.3.5 to make the Specification treatment of the

valves consistent with the Mode 6 and Mode 5-loops drained

requirements.

Due to the addition of Technical Specification 3/4.3.5, the

related Bases information is added as BASES Section 3/4.3.5.

Additionally, the Bases information for the Shutdown Margin Monitors

which is currently in BASES Section 3/4.3.1 is moved to the added

BASES Section 3/4.3.5. This Bases information is also revised to be

consistent with the added Specification 3/4.3.5.

Also, due to the addition of Technical Specification 3/4.3.5,

the guidance related to the Shutdown Margin Monitor in Tables 3.3-1

and 4.3-1 is deleted to avoid redundancy.

Additionally, Section 3/4.1.2 of the Bases is revised so that it

refers to Figure 3.1-4 (Shutdown Margin for Mode 5/filled) instead

of Figure 3.1-5 (Shutdown Margin for Mode 5/drained). This change

will make the Bases consistent with the ACTION statement

requirements of Technical Specifications 3.1.2.2 and 3.1.2.6.

Finally, Reference 12 (NUSCO-152, Addendum 4) is added to the

list of references in Section 6.9.1.6.b. The addition of this

reference is considered administrative and is not related to or

required by the changes proposed for the Shutdown Margin

requirements or Shutdown Margin Monitors.

The new requirements for increased Shutdown Margin (Figures 3.1-

1 to 3.1-5) and additional boration (LCO 3.3.5) continue to assure

that the operator will have a response time of at least 15 minutes

to mitigate the consequences of a Boron Dilution Event. The

implementation of the new requirements does not alter the alignment

of any plant equipment and therefore, the change cannot increase the

probability or consequences of any previously analyzed accident.

The proposed changes will not adversely affect the assumptions

or results of other FSAR [Final Safety Analysis Report] accident

analysis and it is concluded that this change is safe. The changes

do not adversely affect any equipment credited in the safety

analysis.

Based upon the re-analyses of the boron dilution event, revised

plant operating requirements (shutdown margin) are generated to

maintain the required operator action time. Therefore, there is no

effect on the probability of occurrence or consequences of

previously evaluated accidents.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed Shutdown Margin requirements for Modes 3, 4 and 5

(Figures 3.1-1 to 3.1-5 and additional boration as per Specification

3.3.5) will continue to assure that the operator has a minimum of 15

minutes from the alarm to loss of shutdown margin during an assumed

Boron Dilution Event. Additionally, the use of these revised

requirements allows the Shutdown Margin Monitor to be considered

operable for count rates which are lower than currently allowed.

The changes do not introduce any new failure modes or

malfunctions since the changes implement revised, more conservative

plant operating requirements (shutdown margin) which are based on

re-analyses of the Boron Dilution Event. Also, the changes do not

eliminate any existing requirements.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

The proposed Shutdown Margin requirements for Modes 3, 4 and 5

(Figures 3.1-1 to 3.1-5 and additional boration as per Specification

3.3.5) will continue to assure that the operator has a minimum of 15

minutes from the alarm to loss of shutdown margin during an assumed

Boron Dilution Event. Additionally, the use of these revised

requirements allows the Shutdown Margin Monitor to be considered

operable for count rates...which are lower than currently allowed.

The re-analyses of the Boron Dilution Event demonstrated that

the required

[[Page 33130]]

operator action time is maintained. As such, the re-analyses will

become the ``analysis of record'' for the Boron Dilution Event in

Modes 3, 4 and 5. The Boron Dilution Event analysis is documented in

FSAR Chapter 15.4.6.

The re-analyses of the Boron Dilution Event and the proposed

revisions to the Technical Specifications do not adversely affect

the results of the current FSAR accident analysis and therefore, it

is concluded that this change is safe. Additionally, the change does

not adversely affect any equipment credited in the safety analysis.

The changes do not have an adverse impact on the protective

boundaries and there is no reduction in the margin of safety as

specified in the Technical Specifications. Thus, this proposed

change does not involve a significant reduction in the margin of

safety.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed changes do not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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

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

CT 06141-0270.

NRC Deputy Director: Phillip F. McKee.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,

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

Connecticut

Date of amendment request: May 14, 1997

Description of amendment request: Technical Specification

Surveillance Requirement 4.8.2.1.c.4 requires that each battery charger

be tested to verify that it can supply a specified current at 125

volts. The proposed amendment would increase the required test voltage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with

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

significant hazards consideration (SHC). The basis for this

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

satisfied. The proposed revision does not involve [an] SHC because

the revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed changes to Technical Specification Surveillance

4.8.2.1.c.4 to increase the required test voltage for the battery

chargers from 125 volts to greater than or equal to 132 volts is

consistent with the design criteria of the chargers and performing

battery charger surveillance testing does not significantly increase

the probability of an accident previously evaluated. The proposed

changes to increase the required test voltage for the battery

chargers provides the necessary assurance that the battery chargers

will function as required in previous evaluations and does not

significantly increase the consequence of an accident previously

evaluated.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes to Technical Specification Surveillance

4.8.2.1.c.4 to increase the required test voltage for the battery

chargers from 125 volts to greater than or equal to 132 volts does

not change the operation of the battery chargers during normal or

accident evaluations.

Therefore, the proposed revision does not create the possibility

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

evaluated.

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

The proposed change to Technical Specification Surveillance

4.8.2.1.c.4 to increase the required test voltage for the battery

chargers from 125 volts to greater than or equal to 132 volts

provides assurance that the battery chargers are capable of

supplying the largest combined demands of the various steady state

loads, plus the current required to recharge its battery, which has

undergone a duty cycle discharge, to its fully charged condition in

less than 24 hours.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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

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

CT 06141-0270.

NRC Deputy Director: Phillip F. McKee.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek

Generating Station, Salem County, New Jersey

Date of amendment request: March 3, 1997 as supplemented by letter

dated May 5, 1997. The May 5, 1997, supplement revised the proposed no

significant hazards consideration entirely

Description of amendment request: The proposed changes to the Hope

Creek (HC) Technical Specifications (TSs) would: (1) Change TS 3/4.3.1,

``Reactor Protection System Instrumentation,'' TS 3/4.3.2, ``Isolation

Actuation Instrumentation,'' and TS 3/4.3.3, ``Emergency Core Cooling

System Actuation Instrumentation'' to include additional information

concerning response time testing; (2) Change TS 4.0.5 to reference

inservice inspection and test requirements; (3) Change TS 3/4.6.1,

``Primary Containment,'' and associated Bases to reflect a design

modification; (4) Change TS 3/4.7.7, ``Main Turbine Bypass System,'' to

specify a new operability requirement; and (5) Change the Bases for TS

3/4.8, ``Electrical Power Systems.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes for the TS related to response time testing

reflect testing methodologies that were approved by the NRC in

Amendment No. 85 to the Hope Creek TS. These proposed TS revisions

involve: (1) no hardware changes; (2) no significant changes to the

operation of any systems or components in normal or accident

operating conditions; and (3) no changes to existing structures,

systems or components. Therefore, these changes will not increase

the probability of an accident previously evaluated. Since the plant

systems associated with these proposed changes will still be capable

of: (1) meeting all applicable design

[[Page 33131]]

basis requirements; and (2) retain the capability to mitigate the

consequences of accidents described in the HC [Updated Final Safety

Analysis Report] UFSAR, the proposed changes were determined to be

justified. As a result, these changes will not involve a significant

increase in the consequences of an accident previously evaluated.

The proposed changes to Surveillance Requirement 4.0.5 do not

alter the current requirements for the Hope Creek inservice

inspection and inservice testing programs and are considered to be

editorial in nature. These proposed TS revisions involve: (1) no

hardware changes; (2) no significant changes to the operation of any

systems or components in normal or accident operating conditions;

and (3) no changes to existing structures, systems or components.

Therefore, these changes will not increase the probability of an

accident previously evaluated. Since the plant systems associated

with these proposed changes will still be capable of: (1) Meeting

all applicable design basis requirements; and (2) retain the

capability to mitigate the consequences of accidents described in

the HC UFSAR, the proposed changes were determined to be justified.

As a result, these changes will not involve a significant increase

in the consequences of an accident previously evaluated.

The proposed changes to the drywell and suppression chamber

purge system are being made to justify design modifications to that

system. As discussed in NRC Notice of Violation 50-354/96-10-01,

this design modification replaced isolation valves containing

resilient material seals with metal seated valves under 10CFR50.59.

As a result of this modification, a 24 month frequency has been

implemented to perform Type C tests on these new metal seated

valves. PSE&G has concluded that the 24 month frequency is

appropriate for the new valves since: (1) This frequency is imposed

by Surveillance Requirement 4.6.1.2.d, which is applicable to

similar containment isolation valves in Table 3.6.3-1 that penetrate

the primary containment; and (2) concerns raised about severe

environment-induced degradation and frequent use for the previously

installed resilient seal material valves are not applicable to the

replacement metal seat valves. PSE&G has concluded that the valve

modification was an enhancement to the Hope Creek design that did

not impact the isolation capability of the drywell and suppression

chamber purge system. No significant changes were made to the

operation of these valves in normal or accident operating

conditions. As a result, these changes will not increase the

probability of an accident previously evaluated. Since the plant

systems associated with these proposed changes will still be capable

of: (1) Meeting all applicable design basis requirements; and (2)

retain the capability to mitigate the consequences of accidents

described in the HC UFSAR, the proposed changes were determined to

be justified. As a result, these changes will not involve a

significant increase in the consequences of an accident previously

evaluated.

The proposed changes to [Limiting Condition for Operation] LCO

3.7.7 establish consistent and appropriate requirements for main

turbine bypass valve operability requirements. These changes do not

impact the assumptions contained in these UFSAR analyses since they

do not change the manner in which Hope Creek is currently permitted

to operate. Since the ACTION Statement for LCO 3.7.7 already allows

indefinite continued operation below 25% of RATED THERMAL POWER with

an inoperable main turbine bypass valve system, the proposed

modification to the APPLICABILITY statement for this LCO does not

involve: (1) Hardware changes; (2) significant changes to the

operation of any systems or components in normal or accident

operating conditions; or (3) changes to existing structures, systems

or components. Therefore these changes will not increase the

probability of an accident previously evaluated. Since the plant

systems associated with these proposed changes will still be capable

of: (1) meeting all applicable design basis requirements; and (2)

retain the capability to mitigate the consequences of accidents

described in the HC UFSAR, the proposed changes were determined to

be justified. As a result, these changes will not involve a

significant increase in the consequences of an accident previously

evaluated.

The proposed changes to the HC emergency diesel generator (EDG)

TS Bases [Change 5--Bases for TS 3/4.8, ``Electrical Power

Systems''] include information contained in the Safety Evaluation

Report for Technical Specification Amendment No. 75. This

information concerns the bases for the allowed-outage-time (AOT) for

the C and D EDGs. Concerning the revisions to planned C and D EDG

outages, PSE&G believes that implementation of 10CFR50.65

requirements to monitor EDG unavailability will provide an

acceptable and more clearly defined method for maintaining EDG

availability within acceptable limits. As stated in PSE&G's letter

LR-N97167, dated March 21, 1997, Hope Creek will not plan C or D EDG

outages that exceed 72 hours if the total unavailability of the EDG

will be greater than 720 hours on a 12 month rolling basis. The

proposed TS revisions involve: (1) no hardware changes; (2) no

significant changes to the operation of any systems or components in

normal or accident operating conditions; and (3) no changes to

existing structures, systems or components. Therefore these changes

will not increase the probability of an accident previously

evaluated. Since the plant systems associated with these proposed

changes will still be capable of: (1) Meeting all applicable design

basis requirements; and (2) retain the capability to mitigate the

consequences of accidents described in the HC UFSAR, the proposed

changes were determined to be justified. As a result, these changes

will not involve a significant increase in the consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes for the TS related to response time testing

reflect testing methodologies that were approved by the NRC in

Amendment No. 85 to the Hope Creek TS and are being made to clarify

the licensing basis for performing response time testing. The

proposed changes will not adversely impact the operation of any

safety related component or equipment. Since the proposed changes

involve: (1) No hardware changes; (2) no significant changes to the

operation of any systems or components; and (3) no changes to

existing structures, systems or components, there can be no impact

on the occurrence of an accident previously evaluated. Furthermore,

there is no change in plant testing proposed in this change request

that could initiate an event. Therefore, these changes will not

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

any accident previously evaluated.

The proposed changes to Surveillance Requirement 4.0.5 do not

alter the current requirements for the Hope Creek inservice

inspection and inservice testing programs and are considered to be

editorial in nature. The proposed changes will not adversely impact

the operation of any safety related component or equipment. Since

the proposed changes involve: (1) No hardware changes; (2) no

changes to the operation of any systems or components; and (3) no

changes to existing structures, systems or components, there can be

no impact on the occurrence of an accident previously evaluated.

Furthermore, there is no change in plant testing proposed in this

change request that could initiate an event. Therefore, these

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

of accident from any accident previously evaluated.

The proposed changes to the drywell and suppression chamber

purge system are being made to justify design modifications to that

system. As discussed in NRC Notice of Violation 50-354/96-10-01,

this design modification replaced isolation valves containing

resilient material seals with metal seated valves under 10 CFR

50.59. As a result of this modification, a 24 month frequency has

been implemented to perform Type C tests on these new metal seated

valves. PSE&G has concluded that the 24 month frequency is

appropriate for the new valves since: (1) This frequency is imposed

by Surveillance Requirement 4.6.1.2.d, which is applicable to

similar containment isolation valves in Table 3.6.3-1 that penetrate

the primary containment; and (2) concerns raised about severe

environment-induced degradation and frequent use for the previously

installed resilient seal material valves are not applicable to the

replacement metal seat valves. PSE&G has concluded that the valve

modification was an enhancement to the Hope Creek design that did

not impact the isolation capability of the drywell and suppression

chamber purge system. Since the proposed changes will not adversely

impact the operation of any safety related component or equipment,

there can be no impact on the occurrence of any accident.

Furthermore, there is no change in plant testing proposed in this

change request that could initiate an event. Therefore, these

changes will not create the possibility of a

[[Page 33132]]

new or different kind of accident from any accident previously

evaluated.

The proposed changes to LCO 3.7.7 establish consistent and

appropriate requirements for main turbine bypass valve operability

requirements. These changes do not impact the assumptions contained

in these UFSAR analyses since they do not change the manner in which

Hope Creek is currently permitted to operate. Since the ACTION

Statement for LCO 3.7.7 already allows indefinite continued

operation below 25% of RATED THERMAL POWER with an inoperable main

turbine bypass valve system, the proposed modification to the

APPLICABILITY statement for this LCO does not involve: (1) hardware

changes; (2) significant changes to the operation of any systems or

components in normal or accident operating conditions; or (3)

changes to existing structures, systems or components. The proposed

changes will not adversely impact the operation of any safety

related component or equipment. Since the proposed changes involve:

(1) no significant hardware changes; (2) no significant changes to

the operation of any systems or components; and (3) no changes to

existing structures, systems or components, there can be no impact

on the occurrence of any accident. Furthermore, there is no change

in plant testing proposed in this change request that could initiate

an event. Therefore, these changes will not create the possibility

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

evaluated.

The proposed changes to the HC emergency diesel generator (EDG)

TS Bases [Change 5--Bases for TS \3/4\.8, ``Electrical Power

Systems''] include information contained in the Safety Evaluation

Report for Technical Specification Amendment No. 75. This

information concerns the bases for the allowed-outage-time (AOT) for

the C and D EDGs. Concerning the revisions to planned C and D EDG

outages, PSE&G believes that implementation of 10CFR50.65

requirements to monitor EDG unavailability will provide an

acceptable and more clearly defined method for maintaining EDG

availability within acceptable limits. As stated in PSE&G's letter

LR-N97167, dated March 21, 1997, Hope Creek will not plan C or D EDG

outages that exceed 72 hours if the total unavailability of the EDG

will be greater than 720 hours on a 12 month rolling basis. The

proposed changes will not adversely impact the operation of any

safety related component or equipment. Since the proposed changes

involve: (1) No hardware changes; (2) no significant changes to the

operation of any systems or components; and (3) no changes to

existing structures, systems or components, there can be no impact

on the occurrence of any accident. Furthermore, there is no change

in plant testing proposed in this change request which could

initiate an event. Therefore, these changes will not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes for the TS related to response time testing

reflect testing methodologies that were approved by the NRC in

Amendment No. 85 to the Hope Creek TS. No changes are being made to

methodologies with this proposal. Therefore, the changes contained

in this request do not result in a significant reduction in a margin

of safety.

The proposed changes to Surveillance Requirement 4.0.5 do not

alter the current requirements for the Hope Creek inservice

inspection and inservice testing programs and are considered to be

editorial in nature. Therefore, the changes contained in this

request do not result in a significant reduction in a margin of

safety.

The proposed changes to the drywell and suppression chamber

purge system are being made to reflect design modifications that

have been installed. This design modification replaced isolation

valves containing resilient material seals with metal seated valves

under 10 CFR 50.59. PSE&G has concluded that the 24 month frequency

is appropriate for the new valves since: (1) this frequency is

imposed by Surveillance Requirement 4.6.1.2.d, which is applicable

to other containment isolation valves in Table 3.6.3-1 that

penetrate the primary containment; and (2) concerns raised about

severe environment-induced degradation and frequent use for the

previously installed resilient seal material valves are not

applicable to the replacement metal seat valves. The valve

modification was an enhancement to the Hope Creek design that did

not impact the isolation capability of the drywell and suppression

chamber purge system, and does not result in a significant reduction

in a margin of safety.

The proposed changes to LCO 3.7.7 establish consistent and

appropriate requirements for main turbine bypass valve operability

requirements. These changes do not impact the assumptions contained

in these UFSAR analyses since they do not change the manner in which

Hope Creek is currently permitted to operate. Since the ACTION

Statement for LCO 3.7.7 already allows indefinite continued

operation below 25% of RATED THERMAL POWER with an inoperable main

turbine bypass valve system, the proposed modification to the

APPLICABILITY statement for this LCO would be editorial in nature.

Therefore, the changes contained in this request do not result in a

significant reduction in a margin of safety.

The HC TS Bases [Change 5--Bases for TS \3/4\.8, ``Electrical

Power Systems''] will be revised to include information contained in

the Safety Evaluation Report for Technical Specification Amendment

No. 75. This information concerns the bases for the allowed-outage-

time (AOT) for the C and D emergency diesel generators (EDGs). PSE&G

believes that implementation of 10 CFR 50.65 requirements to monitor

EDG unavailability limits will provide an acceptable and more

clearly defined method for maintaining EDG availability within

acceptable limits and not result in a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070.

Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and

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

NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek

Generating Station, Salem County, New Jersey

Date of amendment request: May 19, 1997

Description of amendment request: The proposed amendment would change

Technical Specification (TS) 3.7.1.3, ``Ultimate Heat Sink'' to reflect

that continued plant operation depends upon the association of ultimate

heat sink (UHS) temperature and safety system availability. The

requirements of TS 3.7.1.1, ``Safety Auxiliaries Cooling System

(SACS)'', TS 3.7.1.2, ``Station Service Water System (SSWS)'' and TS

3.8.1.1, ``Electrical Power Systems'' would be revised to reflect the

revised TS 3.7.1.3. In addition, the Bases for \3/4\.7.1, ``Service

Water Systems'' would be appropriately 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. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed TS revisions related to SSWS/SACS and the emergency

diesel generators (EDGs) [TS 3.7.1.1, TS 3.7.1.2, and TS 3.8.1.1]

involve no hardware changes and no changes to existing structures,

systems or components. The additional system configuration limits and

changes to the operation of SSWS/SACS/EDGs are being made to ensure

that SSWS/SACS can remove required heat loads during design basis

accidents and transients with the proposed UHS river water temperature

and level limits. The link to the UHS LCO in the proposed SSWS/SACS/EDG

TS ACTION Statements and the proposed revisions to the SACS ACTION

Statement for one inoperable SACS subsystem ensure that the plant is

directed to enter a safe shutdown condition whenever the capability to

[[Page 33133]]

mitigate design basis accidents and transients is lost. Since the SSWS/

SACS/EDGs will still remain capable of meeting all applicable design

basis requirements and retaining the capability to mitigate the

consequences of accidents described in the HC UFSAR, the proposed

changes were determined to be justified. As a result, these changes

will not increase the probability of an accident previously evaluated

nor significantly increase in the consequences of an accident

previously evaluated.

The proposed TS revisions related to UHS [TS 3.7.1.3] involve no

hardware changes and no changes to existing structures, systems or

components. The additional system configuration limits and changes

to the operation of UHS supported systems are being made to ensure

that the UHS can remove required heat loads during design basis

accidents and transients with the proposed UHS river water

temperature and level limits. The proposed UHS TS ACTION Statements

ensure that the plant is directed to enter a safe shutdown condition

whenever the capability to mitigate design basis accidents and

transients is lost. The proposed changes to the UHS TS surveillance

requirements to increase monitoring of the river water temperature

at 82 deg.F adequately ensures that the actions required when river

temperatures exceed 85 deg.F are taken as appropriate. Since the UHS

will still remain capable of meeting all applicable design basis

requirements and retaining the capability to mitigate the

consequences of accidents described in the HC UFSAR, the proposed

changes were determined to be justified. As a result, these changes

will not increase the probability of an accident previously

evaluated nor significantly increase in the consequences of an

accident previously evaluated.

With the approval of the proposed changes to the SSWS/SACS/EDG/

UHS TS, the proposed TS Bases changes are considered to be editorial

in nature. As a result, the proposed Bases changes will not increase

the probability of an accident previously evaluated nor

significantly increase in the consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes to the SSWS/SACS/EDG TS contained in this

submittal will not adversely impact the operation of any safety

related component or equipment. Since the proposed changes involve

no hardware changes and no changes to existing structures, systems

or components, there can be no impact on the potential occurrence of

any accident due to new equipment failure modes. The additional

system configuration limits and changes to the operation of SSWS/

SACS/EDGs imposed by the proposed changes ensure that SSWS/SACS and

the UHS can remove required heat loads during design basis accidents

and transients with the proposed UHS river water temperature and

level limits. Furthermore, there is no change in plant testing

proposed in this change request which could initiate an event.

Therefore, these changes will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes to the UHS TS contained in this submittal

will not adversely impact the operation of any safety related

component or equipment. Since the proposed changes involve no

hardware changes and no changes to existing structures, systems or

components, there can be no impact on the potential occurrence of

any accident due to new equipment failure modes. The additional

system configuration limits imposed by the proposed UHS LCO ensure

that supported systems can remove required heat loads during design

basis accidents and transients with the proposed UHS river water

temperature and level limits. Furthermore, there is no change in

plant testing proposed in this change request which could initiate

an event. The proposed changes to the UHS TS surveillance

requirements to increase monitoring of the river water temperature

at 82 deg.F adequately ensures that the actions required when river

temperatures exceed 85 deg.F are taken as appropriate. Therefore,

these changes will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

With the approval of the proposed changes to the SSWS/SACS/EDG

UHS TS, the proposed TS Bases changes are considered to be editorial

in nature. As a result, the proposed Bases changes will not create

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

accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes for the TS related to the SSWS/SACS/EDGs

establish consistent and appropriate requirements for SSWS/SACS/EDG

and UHS operability requirements. The additional system

configuration limits and changes to the operation of SSWS/SACS/EDG

are being made to ensure that SSWS/SACS can remove required heat

loads during design basis accidents and transients with the proposed

UHS river water temperature and level limits. The link to the UHS

LCO in the proposed SSWS/SACS/EDG TS ACTION Statements and the

revision to the SACS ACTION Statement for one inoperable SACS

subsystem ensure that the plant is directed to: (1) enter a safe

shutdown condition whenever the capability to mitigate design basis

accidents and transients is lost; or (2) enter a conservatively

short period of continued operation when system redundancy is

reduced. Since the SSWS/SACS/EDG will still remain capable of

meeting all applicable design basis requirements and retaining the

capability to mitigate the consequences of accidents described in

the HC UFSAR, the proposed changes contained in this submittal were

determined to not result in a significant reduction in a margin of

safety.

The proposed changes for the TS related to the UHS ensure

continued capability of the UHS to mitigate the consequences of

design basis accidents and transients. The additional SSWS/SACS

configuration limits and changes to the operating limits of the UHS

ensure that the UHS can remove required heat loads during design

basis accidents and transients with the proposed river water

temperature and level limits. The proposed UHS TS ACTION Statements

ensure that the plant is directed to: (1) enter a safe shutdown

condition whenever the capability to mitigate design basis accidents

and transients is lost; or (2) enter a conservatively short period

of continued operation when supported system redundancy is reduced.

Since the UHS will still remain capable of meeting all applicable

design basis requirements and retaining the capability to mitigate

the consequences of accidents described in the HC UFSAR, the

proposed changes contained were determined to not result in a

significant reduction in a margin of safety.

With the approval of the proposed changes to the SSWS/SACS/UHS

TS, the proposed TS Bases changes are considered to be editorial in

nature. As a result, the proposed bases changes will not result in a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and

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

NRC Project Director: John F. Stolz.

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

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

Unit No. 1, Fairfield County, South Carolina

Date of amendment request: May 21, 1997

Description of amendment request: The proposed amendment would

revise the Virgil C. Summer Nuclear Station Technical Specifications

(TS), Surveillance Requirements (SRs), to change the methodology for

testing the charcoal adsorbers in (1) the control room normal and

emergency air handling system (TS 3/4.7.6), and (2) the spent fuel pool

ventilation system (TS 3/4.9.11), by reference to the methodology of

ASTM D 3803-1989 from the ANSI STD N509-1980.

The proposed reference testing methodology to ASTM D 3803-1989 for

the control room is at a relative humidity of 70% and 30 degrees C with

methyl iodide penetration of avg greater than 350 deg.F'' and 3/4.5.3, ``ECCS

Subsystems--Tavg less than 350 deg.F''), and delete cycle-

specific guidance concerning manual emergency engineered safety feature

function input checks.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated in

the FSAR [Final Safety Analysis Report]. The purposes for

repositioning the breakers/disconnects for MOVs [motor-operated

valves] 8706A and 8706B are to ensure that the ECCS [Emergency Core

Cooling System] System is aligned properly such that the assumptions

used in the safety analyses are met and to prevent possible

overpressurization of the charging pump suction line piping. The

likelihood of a severe transient occurring in this time frame is

very small and has to be weighed against the possibility of over

pressurizing the CVCS [Chemical and Volume Control System] charging

pump suction piping. The allowance of a 4 hour time period to

perform the required alignment appropriately weighs this risk.

Changing the applicability of the requirement to have indication

from a Source Range Nuclear Instrument available to agree with the

design of the plant does not change the physical design of the plant

or affect any assumptions used in accident analyses and, therefore,

has no effect on the probability or consequences of an accident

previously evaluated in the FSAR. The allowance of 1 hour to perform

the Source Range Channel Check upon reaching P-6 from Mode 2 is

consistent with the current basis for a source range channel

inoperable. Therefore, these changes do not involve a significant

increase in the consequences of an accident previously evaluated.

(2) The proposed changes to the Technical Specifications do not

increase the possibility of a new or different kind of accident than

any accident already evaluated in the FSAR. No new limiting single

failure or accident scenario has been created or identified due to

the proposed changes. Safety-related systems will continue to

perform as designed. Therefore, the proposed changes do not create

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

previously evaluated.

(3) The proposed changes do not involve a significant reduction

in the margin of safety. The margin of safety is not significantly

reduced due to the proposed changes to the breaker/disconnect

positioning requirements of TS [Technical Specifications] 3/4.5.2

and 3/4.5.3 when transitioning between Modes 3 and 4. The likelihood

of either a severe transient occurring in Mode 3 or the possible

overpressurization of the CVCS charging pump suction line by the RHR

[residual heat removal] system in Mode 4 is very small. Changing the

Applicability of the requirement to have indication from a Source

Range Nuclear Instrument available to agree with the design of the

plant does not change the physical design of the plant or affect any

assumptions used in accident analyses and, therefore, has no effect

on the margin of safety. These proposed changes are technically

consistent with the requirements and standard format of NUREG-1431,

Revision 1. Performing the source range channel check within 1 hour

upon reaching P-6 from Mode 2 does not change the physical design of

the plant or affect any assumptions used in accident analyses and,

therefore, also does not [a]ffect the margin of safety. Thus, 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

[[Page 33135]]

proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201.

NRC Project Director: Herbert N. Berkow.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Date of amendments request: May 28, 1997

Description of amendments request: The proposed amendments would

insert a footnote in Technical Specification (TS) Surveillance

Requirement 4.8.1.1.2.e, to clarify that load rejection testing of the

shared emergency diesel generator set on either unit may be used to

satisfy TS 4.8.1.1.2.e surveillance requirements for both units.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes clarify that load rejection testing of the

shared emergency diesel generator set is only required once per five

years, and that testing of the shared EDG [emergency diesel

generator] set on one unit may be used to satisfy SR [Surveillance

Requirement] 4.8.1.1.2.e requirements for both units. These changes

do not affect the probability or consequences of an accident. There

are no changes being made to the emergency diesel generator testing

program. These changes simply clarify the existing test program and

the intent of the test requirements.

Therefore, the proposed TS changes do 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 proposed changes clarify that load rejection testing of the

shared emergency diesel generator set is only required once per five

years, and that testing of the shared EDG set on one unit may be

used to satisfy SR 4.8.1.1.2.e requirements for both units. No new

testing configuration is being proposed that could create the

possibility of any new or different kind of accident from any

accident previously evaluated. There are no changes being made to

the emergency diesel generator testing program. These changes simply

clarify the existing test program and the intent of the test

requirements.

Therefore, the proposed changes do 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 proposed changes clarify that load rejection testing of the

shared emergency diesel generator set is only required once per five

years, and that testing of the shared EDG set on one unit may be

used to satisfy SR 4.8.1.1.2.e requirements for both units. A

similar technical specification change has been previously approved

by the NRC for Hatch Nuclear Plant. The technical specification

bases and the Final Safety Analysis Report have been reviewed.

Clarification of the testing requirements has no effect on the

margin of plant safety since no reduction in the test program is

involved.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201.

NRC Project Director: Herbert N. Berkow.

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit 1, Lake County, Ohio

Date of amendment request: May 2, 1997.

Description of amendment request: The proposed change would

continue to allow entry into Operational Conditions 1, 2, and 3 with

the inboard main steam isolation valve (MSIV) leakage control subsystem

inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This License Amendment application proposes a revision to the

exception to Limiting Condition for Operation (LCO) 3.0.4 as it

applies to the Technical Specification (TS) for the MSIV Leakage

Control System (LCS). This revision is proposed to permit completion

of activities necessary to implement the most appropriate permanent

resolution for the issues that resulted from the elimination of the

secondary containment bypass leakage path through the Main Steam

Line drains. In addition, the revision clarifies that the exception

only applies to the Inboard MSIV LCS subsystem. The drains will

remain in their current configuration, which seals off the secondary

containment bypass leakage path. The sealed drain path results in a

temporary inoperability of the Inboard MSIV LCS subsystem when the

plant is operated below 50 percent rated thermal power (RTP), due to

condensate build-up in the bottom of the steam lines between the

MSIVs. The requested 3.0.4 exception is necessary to permit plant

startups with this temporary inoperability. The exception to LCO

3.0.4 simply permits use of the existing Action statement (Condition

A of LCO 3.6.1.9) during MODE changes.

The probability of occurrence of a previously evaluated accident

is not affected by the proposed revision of the LCO 3.0.4 exception

since no change to the plant or to the manner in which the plant is

operated is involved. The existing plant configuration will be

maintained, and possible concerns resulting from that configuration

have been analyzed. The extra weight of the water pooled between the

MSIVs was analyzed with respect to piping supports and seismic

considerations and was found to be acceptable, and condensate that

is carried past the outboard MSIVs will be drained to the condenser

by drain connections downstream of the outboard MSIVs before it can

reach the turbine. The temporary inoperability of the Inboard MSIV

LCS subsystem when below 50 percent RTP has no impact on accident

initiation probability, since the MSIV LCS does not serve to prevent

accidents, but is only used in mitigating the consequences of Loss

of Coolant Accidents (LOCAs) that have already occurred.

The consequences of an accident are not affected in that the

Outboard MSIV LCS subsystem will be available to perform the MSIV

LCS function by mitigating the consequences of a LOCA during the

temporary period in which the Inboard MSIV LCS subsystem is

unavailable. Condensate that is carried past the outboard MSIVs will

be drained to the condenser by drain connections downstream of the

outboard MSIVs; therefore, no impairment of the Outboard MSIV LCS

subsystem will result from condensed water. The Required Action and

Completion Time for one inoperable MSIV LCS subsystem remains the

same, and limits plant operation to the previously established 30-

day Allowable Outage Time. The Required Action if both subsystems of

MSIV LCS were to become inoperable also remains the same. The MSIV

function of isolating the Main Steam Lines is also unaffected by the

existing plant

[[Page 33136]]

configuration, since MSIV performance will not be affected by the

existence of accumulated water in the bottom of the steam lines

between the MSIVs during plant operation below 50 percent RTP.

Therefore, if necessary, the Main Steam Lines will be isolated, and

leakage past the MSIVs will be routed for filtration as in the

design-basis radiological analyses, and the safety and radiological

consequences of previously evaluated accidents will remain

unaffected.

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

or different kind of accident from any accident previously

evaluated.

The proposed change to permit inoperability of the Inboard MSIV

LCS subsystem during periods of startup and power ascension to 50

percent RTP and during shutdown below 50 percent RTP does not create

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

previously evaluated. The Inboard MSIV LCS subsystem is only

credited during a large-break LOCA wherein Reactor Coolant System

depressurization occurs. The temporary unavailability of the Inboard

MSIV LCS subsystem can be mitigated by operation of the Outboard

MSIV LCS subsystem. The amendment to the TS is an administrative

change that does not involve change to the current plant design or

methods of operation. No new plant equipment failure modes or

accident initiators are introduced by the LCO 3.0.4 exception.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The response to a large-break LOCA will not be affected since

the Outboard MSIV LCS subsystem can be assumed to be available

during the limited period of time that the Technical Specifications

permit the Inboard subsystem to be unavailable. Allowing entry into

MODES 1, 2, and 3 while utilizing the existing Condition A and

Required Action A.1 does not reduce the margin of safety since the

Completion Time allowed for that Condition is not increased. The

proposed change will have no adverse impact on the reactor coolant

system pressure boundary nor will other system protective boundaries

or safety limits be affected.

The NRC staff has reviewed the licensees' 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

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

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

NRC Project Director: Gail H. Marcus.

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit 1, Lake County, Ohio

Date of amendment request: May 2, 1997

Description of amendment request: The proposed change would allow

the leakage rate of one or more main steam lines to be up to 35

standard cubic feet per hour (scfh), as long as the total leakage rate

through all four main steam lines is less than or equal to 100 scfh.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change involves the deletion of the portion of

Technical Specification Surveillance Requirement (SR) 3.6.1.3.10

that states the increased leakage rate of less than or equal to 35

scfh for an individual main steam line is only acceptable for

Operating Cycle 6, and a deletion of the restriction that a main

steam line leakage rate of less than or equal to 35 scfh is

acceptable for only one main steam line. The overall main steam line

leakage limit of less than or equal to 100 scfh for all four main

steam lines is not being revised.

The MSIV [main steam isolation valve] leakage is not an

initiator of an accident, including the steam line rupture accident.

Therefore, the probability of an accident previously evaluated has

not changed.

The consequences of interest are the radiological dose

consequences following a large-break Loss of Coolant Accident

(LOCA). This is the event which the regulatory guidance requires to

be evaluated using the extremely conservative source term

assumptions of Regulatory Guide 1.3, ``Assumptions Used for

Evaluating the Potential Radiological Consequences of a Loss of

Coolant Accident for Boiling Water Reactors.'' Since the overall

main steam line leakage rate of less than or equal to 100 scfh for

all four main steam lines is not being revised, the radiological

consequences of an accident previously evaluated has not changed.

Therefore, the probability or consequences of an accident

previously evaluated have not significantly increased.

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

or different kind of accident from any accident previously

evaluated.

This proposed change does not physically alter the plant or

systems or equipment in the plant, or the method for operation of

the plant. Therefore, the proposed change does not create the

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

previously evaluated.

3. The proposed change will not involve a significant reduction

in the margin of safety.

The proposed change does not revise the overall combined leakage

rate of less than or equal to 100 scfh for all four main steam lines

that is permitted in the present Specification. It is the combined

main steam line penetration leakage rate that is assumed in the

radiological accident analyses. Thus, although individual steam line

leakage rates may be less than or equal to 35 scfh, as long as

overall leakage of the four main steam lines is maintained at its

current value of less than or equal to 100 scfh, the proposed change

does not reduce the margin of safety.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the licensees' 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

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

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

NRC Project Director: Gail H. Marcus.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: November 9, 1987, as supplemented March 31,

1988, June 8, 1992 and February 4, 1997

Description of amendment request: The proposed changes would revise

the Technical Specifications (TS) for the North Anna Power Station (NA

1&2). The changes would reformat the operability and surveillance

requirements for the intermediate range (IR) channels to be consistent

with NUREG-0452, Revision 4, ``Standard Technical Specifications (STS)

for Westinghouse Pressurized Water Reactors'' (Fall 1981), which is

applicable to NA 1&2. Also, the proposed changes would revise the

nominal IR high flux trip setpoint. The IR nuclear flux trips provide

backup reactor core protection during reactor startup. There is no

operating condition under which the IR trip provides sole overpower

protection. It is a backup trip only, and no credit is taken for the

trip in the NA 1&2 Updated Final Safety Analysis Report (UFSAR).

Operating experience at NA 1&2 has shown the IR channel response to be

sensitive to core loading patterns, varying core burnups, and control

rod positions, and the variability in the channel response had made it

difficult to maintain the channels in proper calibration. Therefore,

the proposed change would

[[Page 33137]]

elevate the nominal IR high flux trip setpoint from a current

equivalent to 25% of rated thermal power to a current equivalent to 35%

of rated thermal power.

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 would not:]

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. There is no

adverse impact on the safety analysis (since no credit is taken for

the trips in the existing analyses), and no degradation of the

protection system redundancy or reliability. This latter conclusion

is based on sensitivity studies which show that the effectiveness of

the flux trip system in protecting against the low power reactivity

excursions examined in the FSAR is not sensitive to realistic

variations in the actual flux trip setpoint.

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

from any accident previously identified, since the severity of the

analyzed accidents is unchanged, and since only a change to a

setpoint and the associated surveillance requirements for the

reactor protection system is involved.

3. Involve a significant reduction in a margin of safety, since

none of the safety analysis input or assumptions are changed, nor

are the probability nor the consequences of any previously analyzed

accidents increased.

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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: Brenda Mozafari (Acting).

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: March 31, 1997

Brief description of amendment: The proposed amendment woul

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