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

Federal RegisterOct 8, 1997

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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

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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 September 15, 1997, through September 26,

1997. The last biweekly notice was published on September 24, 1997 (62

FR 50000).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of 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 November 7, 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.

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If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station,

Units Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendments request: March 18, 1997, as supplemented by

letters dated July 28, 1997 and September 9, 1997

Description of amendments request: The amendments would revise the

operating licenses for Palo Verde Units 1, 2 and 3 to reflect approval

of Amendment 42 to the Palo Verde Nuclear Generating Station (PVNGS)

Physical Security Plan. Amendment 42 would revise the methods used to

search materials, packages and personnel prior to their entry into the

protected area, as described within the security plan.

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 changes do not significantly increase the

probability or consequences of an accident previously evaluated. The

``accident'' as it relates to the Security Plan would have to be an

impact to the Design Basis Threat (DBT) postulated for PVNGS. This

change does not decrease the overall security systems (as described

in paragraph's (b) through (h) of 10 CFR 73.55) ability to protect

PVNGS with the objective of high assurance against the DBT of

radiological sabotage as stated in 73.1(a). This change does not

delete or contradict any regulatory requirements.

The applicable design basis threat is described in 10 CFR 73.1.

Based on that threat, the probability of an external determined

violent assault by stealth, or deceptive actions, of several persons

is unaffected by the requested changes to the search requirements.

Similarly, an internal threat of an insider, including an employee

(in any position) is no more likely to occur as a result of the

search techniques. The probability of an attack with a four-wheel

drive land vehicle bomb is unaffected. Theft or diversion of formula

quantities of strategic special nuclear material is a threat of

removal from the inside of the protected area, which is not within

the scope of this change that only affects searches of material

entering the protected area.

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

or different kind of accident from any accident previously

evaluated.

The possibility of an accident of a new or different kind has

not been created because the DBT (as described in the Security Plan

and 10 CFR 73.1) would not be changed as a result of these changes.

The changes supplement regulatory requirements and commitments

already described in the PVNGS Physical Security Plan.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not involve a significant reduction in a

margin of safety. These changes to the personnel, material and

package search criteria are not specifically considered in the basis

for any margin of safety. The DBT considers inside assistance by a

knowledgeable individual, however, these changes would not assist

this individual in either sabotage or theft of nuclear material.

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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: William H. Bateman

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

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

Date of amendment request: September 15, 1997

Description of amendment request: The proposed license amendments

would revise the Technical Specifications (TS) to:

1. Revise the reactor coolant system heatup limitation curves in

Figure 3.4-2, which are applicable only to the first 10 effective full-

power years (EFPYs). The revised curves would be (a) applicable to the

first 15 EFPYs; (b) include the latest radiation surveillance capsule

results; (c) remove instrument margins by relocating them to a

licensee-controlled document, ``Pressure Temperature Limit Report;''

and (d) administratively delete certain unneeded footnotes that exist

in the current figure.

2. Modify the actual surveillance capsule identification listed in

Table 4.4-5, ``Reactor Vessel Material Surveillance Program -

Withdrawal Schedule'' (for Unit 2 only) and update each units lead

factors and withdrawal time.

3. Revise the power-operated relief valve (PORV) setpoints in

Section 3.4.9.3.a to less than or equal to 400 pounds per square inch

gauge (psig) (as left calibrated), allowable value less than or equal

to 425 psig (as found).

4. Make editorial changes to improve consistency among various TS

sections to conform with the Westinghouse Improved Standard Technical

Specifications, and update applicable Code references.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below.

1. Will the changes involve a significant increase in the

probability or consequences of an accident previously evaluated?No.

No previously evaluated accident was considered to originate from

use of the heatup curves (change 1. above), the testing and use of

surveillance capsules (change 2. above), the setpoint of PORVs

(change 3. above), and editorial changes to the TS. Also, these

items did not have any role in previously analyzed accident

scenarios and thus no impact on accident consequences. Therefore,

these proposed changes will have

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no impact on the consequences or probabilities of any type of

previously evaluated accidents.

2. Will the changes create the possibility of a new or different

kind of accident from any accident previously evaluated?

No. No actual plant equipment or operating procedure will be

affected by the proposed changes. Hence, no new equipment failure modes

or accidents from those previously evaluated will be created.

3. Will the changes involve a significant reduction in a margin

of safety?

No. The margin of safety is associated with confidence in the

design and operation of the plant. The changes to the TS do not

involve any change to plant design or operation. Thus, the margin of

safety previously analyzed and evaluated is maintained.

On the basis of this analysis, 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

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

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

Carolina

NRC Project Director: Herbert N. Berkow

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: June 14, 1997

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) for the Crystal River Nuclear

Electric Generating Plant Unit 3 (CR-3). The proposed TS changes

reflect the operational limitations in mitigating certain Small break

loss-of-coolant-accident (SBLOCA) events. The licensee also proposed

changes to the associated licensing and design bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below. The proposed changes are

addressed in three major parts: (1) SBLOCA Mitigation, (2) Emergency

Diesel generator (EDG) upgrade and (3) EDG Load Rejection Test and

Steady State Loads.

SBLOCA Mitigation

The licensee's revised SBLOCA analyses show that for certain

sized breaks, a combination of emergency core cooling system (ECCS)

flow to the reactor vessel and emergency feedwater (EFW) flow to the

once through steam generators (OTSG) is needed to provide for

adequate core decay heat removal. Due to load capacity limits on the

A EDG, the length of time that the motor-driven emergency feed

pump-1 (EFP-1) would be available is limited. To ensure adequate EFW

system flow and core decay heat removal, several actions would have

to be initiated. They include A EDG load management, and EFW flow

through the turbine-driven emergency feedwater pump-2 (EFP-2) by

opening the cross tie valve, flow through both the high pressure

injection (HPI) pumps and EFP-1. The proposed TS changes reflect the

operational limitations and other associated required actions to

ensure adequate ECCS and EFW cooling capability remains. These

changes for system cross train dependencies and EDG load management

are required for the remainder of current Cycle 11 only.

1. The proposed Technical Specification changes, modifications,

and operator actions involving SBLOCA mitigation will not result in

a significant increase in the probability of an accident previously

evaluated. In addition, the portions of the change involving cross-

train dependencies and load management are being requested for the

remainder of Cycle 11 only, which limits the impact on any

previously established probabilities. The initiators of any design

basis accident is not affected by the proposed Technical

Specification changes, modifications, and operator actions involving

SBLOCA mitigation. Consequently, there is no significant impact on

any previously evaluated accident probabilities.

The proposed Technical Specification changes, modifications and

operator actions involving SBLOCA mitigation do not result in a

significant increase in the consequences of SBLOCA mitigation-

related accidents previously evaluated. In this regard, the proposed

Technical Specification changes, modifications and operator actions

will not adversely affect the integrated ability of the EDGs and the

EFW, SW [service water], RW [raw water], Control Complex Cooling,

ECCS, DC [Decay Heat Closed Cycle Cooling Water System], Decay Heat

Seawater, and Electrical Distribution Systems to perform their

intended safety functions. Therefore, the combined ability of these

components and systems and actions to mitigate the consequences of a

SBLOCA will continue to be maintained. In fact, the collective

impact of these Technical Specification changes, modifications and

operator actions represents a restoration of the ability to mitigate

the consequences of a SBLOCA, which are consistent with the

consequences assumed in licensing and design basis for CR-3. For

example, the installation of EFW cavitating venturis and the

improved operational range of the turbine driven feedwater pump

increase the ability of the EFW system to mitigate the consequences

of a SBLOCA. In addition, the Technical Specification changes,

modifications and operator actions do not significantly affect the

onsite or offsite doses which remain a small fraction of 10 CFR Part

100 limits.

2. The proposed Technical Specification changes, modifications

and operator actions do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The Technical Specification changes, modifications, and operator

actions do not involve a different initiator for any design basis

accident and do not create new design basis scenarios. SBLOCA

mitigation, utilizing a combination of automatic and manual actions,

is already part of the CR-3 licensing basis. Manual operator actions

necessary for the mitigation of SBLOCAs are currently addressed or

are being addressed in EOPs [emergency operating procedures]. Also,

these Technical Specification changes, modifications and operator

actions restore the ability to mitigate the impact of a SBLOCA,

which is consistent with the CR-3 licensing and design basis. Based

on the above, a new or different kind of accident does not result

from this submittal.

3. The proposed Technical Specification changes, modifications

and operator actions do not involve a significant reduction in the

margin of safety for SBLOCA mitigation. The Technical Specification

changes, modifications and operator actions for the EDGs and the

EFW, SW, RW, Control Complex Cooling Systems represent a restoration

of the overall margin of safety to a degree that it will be

consistent with the existing plant design and licensing bases for

SBLOCA mitigation.

EDG upgrade

This aspect of the proposed license amendment involves increases

in the service ratings of the EDGs. The required amount of fuel oil

in the EDG fuel day tank and fuel storage tank, and lube oil storage

is being increased to ensure that adequate volume is available to

support the new service ratings. The EDG refueling interval load

test parameters are being revised to reflect the increased service

ratings and to ensure that the minimum test load is equal to or

greater than the expected maximum steady state accident load.

Additionally, associated EDG Surveillance Requirements (SR) Bases

are being revised.

1. The proposed Technical Specification changes, modifications

and operator actions do not involve a significant increase in the

probability of an accident previously evaluated because neither the

EDGs nor the EDGs fuel oil and lube oil systems serve as the

initiator for any design basis accident and, therefore, do not

significantly impact any previously evaluated accident

probabilities.

The proposed Technical Specification changes, modifications and

operator actions do not involve a significant increase in the

consequences of an accident previously evaluated because the ability

of the EDGs and the EDG fuel oil and lube oil to perform their

intended safety function has not been adversely affected. The EDGs

and the EDG fuel oil and lube oil systems remain fully capable of

performing their safety function for all design basis accidents. The

increase in loading permitted under these changes will reflect the

manufacturers certified capabilities of the EDGs. Also, the

increase in the required fuel remains within the capabilities of the

fuel tanks. The same potential design basis failures that existed

prior to the EDG upgrades will continue to

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exist subsequent to the modifications. It follows that the

consequences of such failures will remain a small fraction of 10 CFR

Part 100 limits.

2. The proposed Technical Specification changes, modifications

and operator actions do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Also, the proposed Technical Specification changes, modifications

and operator actions do not involve any new accident initiators, or

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

In addition, the configuration and basic function of the EDGs and

EDG's fuel and lube oil systems are unaffected by the changes. In

fact, the EDG upgrades ensure that the previously evaluated

accidents are consistent with system and component capabilities and

the current design and licensing bases.

3. The proposed Technical Specification changes, modifications

and operator actions do not involve a significant reduction in the

margin of safety. The EDGs and EDGs fuel and lube oil systems will

continue to able to be perform their safety function for all design

basis accidents. There is an increase in the net margin of safety

for fuel and lube oil storage since required volumes have been

recalculated and increased, additional margin has been added to the

calculated results, and the required volumes are based on usable

tank volumes instead of tank capacity. These volumes continue to

bound the postulated worse-case accident scenario. The increase in

fuel storage required by the changes remains within the capacity of

the storage tanks. The Technical Specification changes,

modifications and operator actions further ensure that margins

provided in current design and licensing bases are satisfied.

EDG Load Rejection Test and Steady State Loads

The proposed changes for this part affects the TS Bases. The

basis of the EDG load rejection test is being revised to bound the

largest single load. A description of ``steady state'' is being

provided with examples of short duration loads and loads imposed by

the starting of motors. Also, addressed is the licensee's conclusion

that the refueling interval EDG load test is not invalidated by

loads imposed by the starting of motors.

1. The proposed Technical Specification changes, modifications

and operator actions do not involve a significant increase in the

probability of an accident previously evaluated because the EDG load

tests and load rejection test do not serve as the initiator for any

design basis accident and, therefore, do not significantly impact

any previously evaluated probabilities.

The proposed Technical Specification changes, modifications and

operator actions do not involve a significant increase in the

consequences of an accident previously evaluated because the changes

do not affect the ability of the EDGs to perform their intended

safety function. Rather, the Technical Specification changes,

modifications and operator actions provide further assurance that

the EDGs are capable of performing their safety function. Failure of

an EDG has the same consequences as it would if the changes were not

made. It follows that the 10 CFR Part 100 consequences of such

failures has not changed.

2. The proposed Technical Specification changes, modifications

and operator actions do not create the possibility of a new or

different kind of accident from any accident previously evaluated

because the changes do not affect the ability of the EDGs to perform

their intended safety function. The configuration and basic function

of the EDGs, including accurately describing the manufacturer

certified EDGs service ratings and steady state loads, do no create

a possibility for a new or different kind of accident. Although the

load rejection test is for an increased EDG largest single load, the

kind of accident addressed by both the load rejection test and the

refueling load test remain the same.

3. The proposed Technical Specification changes, modifications

and operator actions do not involve a significant reduction in the

margin of safety. The calculated loads imposed by the starting of

motors are short duration, have a low probability of occurrence, and

are expected to be within the manufacturer limits. In fact, the

margin confirmed by EDG refueling load testing and load rejection

testing will demonstrate a restoration of design and licensing

margin and confirm that the EDGs remain fully capable of performing

their safety function for all design basis accidents.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida

33733-4042

NRC Project Director: Frederick J. Hebdon

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

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

County, Connecticut

Date of amendment request: September 16, 1997

Description of amendment request: The proposed changes to the

Technical Specifications (TSs) would modify TS 3.7.1.1, ``Plant Systems

Turbine Cycle Safety Valves.'' During its effort to verify the current

design and licensing bases for Millstone, Unit 2, NNECO has determined

that the maximum allowable power level high trip setpoints with

inoperable steam line code safety valves specified in Table 3.7-1 of TS

3.7.1.1 are incorrect. The trip setpoints were not changed to be

consistent with a previously approved reduction in the maximum power

level high trip setpoint. In addition, NNECO is also in the process of

reanalyzing the inadvertent closure of the main steam isolation valve

(MSIV) and the loss of electrical load events. The results of the

reanalysis indicate that the MSIV event results in the highest peak

pressure in the secondary system and that the formula currently

contained in the TS Bases for TS 3.7.1.1 may not result in the correct

trip setpoints.

Specifically, NNECO proposes to: (1) delete TS Table 3.7.1 by not

allowing operation in Mode 1 or 2 with inoperable steam line code

safety valves, (2) modify the associated action statement in TS

3.7.1.1, and (3) update the TS Bases to reflect the proposed changes

and update the amendment history numbers to reflect previously approved

amendments.

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 change does not involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

This proposed change will remove the ability to operate in Modes

1 or 2 with inoperable main steam line code safety valves. Operation

in Mode 3 will be retained, provided no more than three main steam

line code safety valves per steam generator are inoperable.

The primary function of the main steam line code safety valves

is to prevent secondary system overpressurization. These valves will

also provide reactor core heat removal and design basis accident

mitigation. This proposed change does not affect the length of time

the plant can operate with inoperable main steam line code safety

valves before compensatory actions must be taken. (Four hours is

still allowed to restore the valve(s) to operable status.) This

proposed change does not affect the probability of occurrence of any

design basis accident and does not affect how the main steam line

code safety valves function to mitigate design basis accidents.

Therefore, this change does not significantly increase the

probability or consequences of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed change does not alter the way any structure,

system, or component functions. The proposed change will

conservatively change plant operation in Modes 1 and 2 by removing

the ability to operate at power with inoperable main steam

[[Page 52583]]

line code safety valves as currently specified in Technical

Specification 3.7.1.1. It does not introduce any new failure modes

and does not alter any assumption made in the safety analysis.

Therefore, the change will 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.

This proposed change to Technical Specification 3.7.1.1 will

remove the ability to operate in Modes 1 or 2 with inoperable main

steam line code safety valves. Operation in Mode 3 will be retained,

provided no more than three main steam line code safety valves per

steam generator are inoperable. The operability of the main steam

line code safety valves ensures that the secondary system pressure

will be limited to within 110% (1100 psig) of the design pressure of

1000 psig during the most severe anticipated system operational

transient. This change will not affect the operability requirements

for the main steam line code safety valves and will not affect the

length of time the plant can operate with inoperable main steam line

code safety valves before compensatory actions must be taken. This

will ensure the plant equipment required for design basis accident

mitigation will be available. Therefore, there is no significant

reduction in a margin of safety as defined in the Bases of Technical

Specification 3.7.1.1.

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,

Connecticut

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: August 29, 1997

Description of amendment request: Based on a review and subsequent

calculations of the cold overpressurization protection (COPS) enabling

temperature and the emergency core cooling system (ECCS)/charging

system Mode 3 requirements, NNECO proposes to reduce the COPS enabling

temperature. As a result, NNECO proposed the following Technical

Specifications (TS) changes: new heatup and cooldown pressure/

temperature limit curves and their associated requirements; new power

operated relief valve (PORV) setpoint curves and their associated

requirements; revisions to the reactor coolant loops and coolant

circulation, ECCS, boration systems, and COPS to incorporate the lower

enabling temperature and new restrictions for cold overpressure

protection system (COPPS), PORV undershoot, and residual heat removal

(RHR) relief valve bellows; addition of a footnote to allow a reactor

coolant pump (RCP) to substitute for an RHR pump during heatup from

Mode 5 to Mode 4, which is consistent with the improved standard

technical specification (STS); reword TS 3/4.4.9.3 and its Bases

section to be consistent with the improved STS; and revision of the

affected Bases sections to be consistent with the proposed changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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.

Probability of Occurrence of Previously Evaluated Accidents

Since the PORV setpoints and the COPS enabling temperature have

been calculated in accordance with 10CFR50, Appendix G and ASME

[American Society of Mechanical Engineers] Section XI, the change

will not alter the probability that an overpressurization event will

result in a loss of RV [reactor vessel] integrity. The new PORV

setpoint curves are lower than the current curves in certain

temperature ranges (below approximately 130F and above

approximately 220 deg.F), and therefore the operating window is

slightly decreased. However, the reduced operating window is still

sufficient for normal anticipated pressure fluctuations. Below

160 deg.F, operation of Reactor Coolant Pumps are prohibited if the

PORVs are armed for COPPS; therefore, PORV actuation will not occur

below 160F when the RCPs are running. In a water solid condition,

RCS [reactor coolant system] pressure is maintained via the letdown

low pressure control valve, which, when in automatic mode, maintains

the RCS pressure in a relatively narrow range. When the RCPs are not

running, the PORV COPPS system can be actuated. However, for this

condition, the allowable pressure range is 0 to 418 psia [pounds per

square inch atmospheric]. This pressure range is sufficient to

accommodate normal anticipated pressure fluctuations.

Above 220 deg.F, the minimum pressure range is from 300 psia to

595 psia; this range is sufficient to accommodate normal anticipated

pressure fluctuations. In this temperature range, a pressurizer

bubble is normally present, which will minimize any pressure

fluctuations, thereby limiting the possibility of a PORV actuation.

Based on this, it is concluded that the proposed change will not

impact the probability of occurrence that a PORV will be challenged.

When the RHR relief valves are used for COPS there is no

credible scenario which would result in excessive relief valve

undershoot. This is because these valves are spring loaded relief

valves which are designed to close whenever the RCS pressure

decreases below the nominal setpoint of 440 psig [pounds per square

inch gauge]. This provides assurance that there will be no damage to

the seal of a running RCP.

The proposed changes to the heatup/cooldown curves and the

reduction in the enabling temperature for COPS only affect

operational limits and can not be initiators of an event. The

restrictions on RC [reactor coolant], RHR and ECCS pump operation

can not result in an event initiator. Two separate operator actions

are required to start an ECCS or RC pump. These two necessary

actions as well as procedural controls are sufficient to prevent an

inadvertent ECCS or RC pump start. De-energizing the RCPs when

returning a loop to service can not initiate an event.

The proposed change will provide an operable charging pump to

ensure RCP seal flow and reactivity control will be available. When

the RCP is in operation, the charging pump provides the preferred

method for seal flow. The proposed change minimizes the time that

this preferred method is interrupted. A loss of charging pump seal

flow will not cause a malfunction of an RCP because the pump is

designed to use RCS flow as an alternate method at these conditions.

Not allowing two charging pumps to run simultaneously and requiring

at least one pump to be in pull-to-lock, assures a second pump will

not start on an inadvertent SI [safety injection] and exceed the

assumptions in the Appendix G analysis or initiate a Boron Dilution

or CVCS [chemical and volume control system] Malfunction event. If

an operator were to inadvertently start the second pump, a failure

of the charging throttle valve, FCV-121, and one relief valve

credited for COPS would be necessary to exceed the assumptions in

the Appendix G analysis. In addition, the actual time allowed for

swapping the charging pumps is short. The remainder of the hour

allows for documented verification of the disabling of the required

pump. The proposed change will not change any control systems for

these pumps or alter the system configuration that would affect the

probability of an uncontrolled increase in charging flow. The

procedure requirements to swap pumps and the likelihood of these

[[Page 52584]]

multiple failures occurring during the short duration allowed in

this footnote provide adequate assurance that an overpressurization

event will not occur. Maintaining at least one pump always operable

makes the system more reliable for reactivity control than the

current method which disables both pumps simultaneously.

The proposed change to maintain one charging pump operable in

Mode 4 [cannot] initiate an event because of the stable reactivity

condition of the reactor, the emergency power supply requirement for

the operable charging pump, and the fact that the plant is

procedurally required to be borated to the highest required boron

concentration for Modes 3, 4, or 5 prior to entering Mode 4. These

changes do not effectively change the availability of plant

equipment or the way that the plant is operated.

The proposed change to substitute an RCS loop for an RHR loop

during a planned heatup, can not initiate an event. The RCP will be

verified as operating properly prior to stopping the RHR pump and as

such will not initiate a loss of decay heat removal (by heating up

to steam the SGs [steam generators])/loss of flow. While the RCP is

in operation, it performs the RHR boron mixing function and the

decay heat removal function is not required for heatup. Using the

RCP to perform this function will not affect the probability that

the RCP could fail because it will be operated within its normal

operating design conditions. Aligning RHR in the ECCS lineup will

not affect the probability of a RHR pump to start. The pump will be

operable in this lineup. Currently in Mode 5, RHR is lost on a LOP

[loss of offsite power] and is manually restarted once the diesel is

running. With the proposed change, the RCP will be lost on a LOP and

the RHR pump will have to be manually started. Thus, the proposed

change does not affect the probability that the RHR pump could fail.

Since the current response to a LOP is to manually restart the RHR

pump, operator action is needed independent of this change. The

proposed change allows normally open valves to be closed in Mode 5

to align RHR for ECCS injection. This introduces additional manual

actions which could extend the time required to establish flow. In

addition, if one diesel generator were to fail, manual operation of

a valve in the ESF [engineered safety features] building would be

necessary. The mechanistic 'failure to open' of valves that is

introduced by the change as well as the need for manual operator

action to realign these valves increases the time to establish heat

removal. However, there is sufficient time to re-establish RHR

because this note applies only for a heatup in which the plant will

have been shutdown for at least several hours which causes decay

heat to be low (as compared to high decay heat immediately following

a plant trip). Thus, it is concluded that there is no impact on the

probability of failure of RHR to perform its required function.

The proposed change to the ECCS wording does not result in any

new failure modes that could initiate an event since manual

realignment from the control room is currently allowed. Nor can the

manual alignment of RHR valves initiate an event because this

alignment is only for accident mitigation.

Therefore, the proposed changes do not increase the probability

of occurrence of previously evaluated accidents.

Consequences of Previously Evaluated Accidents

The revised Pressure/Temperature curves were calculated in

accordance with 10CFR50, Appendix G, ASME Section XI, and Regulatory

Guide 1.99, Revision 2. This provides assurance that an inadvertent

overpressurization event will not result in a loss of RV integrity.

The restrictions on RCP operation and the requirement to de-energize

the RCPs in Modes 5 and 6 when returning a loop to service are

consistent with the assumptions made in this Appendix G analysis and

the RCPs are not required for accident mitigation for any previously

evaluated accidents and therefore do not affect the consequences.

The COPS relieving capability is greater than the maximum RCS

pressurization rate resulting from any allowed pump combinations,

and the PORV setpoints have been adjusted to take into account

instrumentation effects. This will provide assurance that COPS will

continue to perform its safety function. Since the COPS enabling

temperature has been demonstrated to be conservative at 275F,

allowing SI pump operability above 275F will have no impact on

vessel non-ductile failure.

The restriction between 275F and 350F on the SI and charging

pumps, has been appropriately moved to the reactor coolant loop

section to provide protection for the RHR system (RCS protective

boundary) and to the cold overpressure protection section to provide

protection for the RHR relief valves and the RCP seals. By

incorporating this requirement previously located in the ECCS TS,

RCS integrity is ensured.

With the RCS less than 160F, the consequences of the PORV

undershoot from the proposed PORV setpoints are that the RCS

pressures may drop below the minimum requirement for RCP seal

integrity. However, no seal damage will occur since a requirement

has been added prohibiting the operation of RCPs below 160 deg.F

with the PORVs not isolated while in the low setpoint mode. With

cold overpressure relief valves in service above the COPS enable

temperature (275 deg.F), restrictions are placed on the startup of

an RCP and the number of ECCS pumps capable of injecting into the

RCS to prevent unacceptable mass or energy addition transients. This

provides assurance that the RHR relief valve capacity will not be

exceeded and that PORV undershoot will not challenge the RCP

1 seal. The restriction on the maximum number of ECCS pumps

ensures that the integrity of the RHR relief valve bellows and the

RCP seals during mass injection transients (i.e., inadvertent SI).

The restrictions on RCS/SG secondary side temperature mismatch

ensure that an unanalyzed energy addition event does not occur when

an RCS loop is placed in operation.

The consequences of a small break LOCA [loss of coolant

accident] in COPS Mode 4 are not affected because the plant will

continue to maintain one charging pump operable in Mode 4. In

addition, additional options are provided in the bases of TS 3/

4.4.9.3 for disabling the required charging and SI pumps that will

allow faster restoration if required to mitigate a LOCA or loss of

RHR in Modes 4, 5 and 6.

An RHR pump will remain available in Mode 4 with manual

realignment from the control room as required to perform its ECCS

safety function. The changes have no impact on the capability of RHR

to function in the ECCS mode. RHR is credited during a safety grade

cold shutdown. The proposed change assures that the RHR system will

be available to perform its heat removal function during a safety

grade cold shutdown and thus, there is no change in the analysis

assumptions or consequences.

The changes also eliminate an inconsistency between the charging

system operability requirements for boration and the charging system

operability requirements for cold overpressure protection. The

requirement to maintain two charging pumps operable in Mode 4 will

be reduced to one charging pump. As stated in the proposed basis

section, a second method of boration is not required to be OPERABLE

in Mode 4 for single failure considerations based on the stable

reactivity condition of the reactor, the emergency power supply

requirement for the operable charging pump, and the fact that the

plant is procedurally required to be borated to the highest required

boron concentration for Modes 3, 4, or 5 prior to entering Mode 4.

This provides assurance that reactivity control will be maintained

and stable while only one charging pump is operable for cold

overpressure concerns. These changes do not effectively change the

availability of plant equipment or the way that the plant is

operated. The changes will not adversely impact the assumption for

the limiting dilution flow path and flow rate and therefore, the

consequences of a boron dilution event are not affected.

The proposed changes will maintain a charging pump operable for

reactivity control while ensuring that the flow limits in the

Appendix G analyses are not exceeded. Remaining within the bounds of

the Appendix G limits ensures reactor vessel integrity in Mode 4.

Since the change maintains the reactor vessel integrity, it does not

introduce any means of releasing radionuclides post-accident. The

consequences of a small break LOCA in Mode 4 are not affected

because the plant will continue to maintain one charging pump

operable in Mode 4. These changes are reflected in TS 3.1.2.1,

3.1.2.2, 3.1.2.3 and 3.1.2.4. Adequate protection is provided for

reactor vessel integrity while maintaining reactivity control

operability.

In Mode 5, RHR requirements are specified for decay heat removal

in the case of a loss of offsite power but none are specified for

ECCS accident mitigation. The first RHR train will be aligned for

injection prior to taking the second train out of service. This

provides assurance that this train will be available if needed in

Mode 5. Currently in Mode 5, following a LOP the RHR system can be

re-established by restarting the RHR pump once the diesel is

running. No valve manipulations

[[Page 52585]]

are necessary. With the proposed change, when the operating RCP

trips following a LOP, some of the RHR valves must be realigned from

the ECCS to heat removal mode. If one diesel generator were to fail,

manual operation of a valve in the ESF building would be necessary.

Since this footnote is only applicable during a heatup, decay heat

will be low. There is sufficient time to re-establish RHR even if

action outside the control room is necessary. Since there are four

operable RCS loops, a bubble drawn in the pressurizer and the RCS

pressurized, the plant will heat up to Mode 4 and natural

circulation will provide core cooling if the RHR system cannot be

re-established. Thus, decay heat removal is assured and there is no

affect on the consequences of a LOP.

Since the structural integrity of the RCS is maintained and

adequate core cooling and reactivity control will be available for

design basis events, the proposed changes will have no adverse

impact on the consequences of previously evaluated accidents.

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 temperature/pressure limits will continue to meet the

requirements of 10CFR50, Appendix G. Since the new limits continue

to provide assurance of reactor vessel integrity, the proposed

change does not create the possibility of an accident of a different

type than previously evaluated. Adequate RCS pressure-relieving

capabilities will continue to be maintained throughout the shutdown

modes. No new malfunctions will be introduced which could result in

a new accident postulated in Modes 3-5.

The restrictions on RCP operation do not create the potential

for unanalyzed heat injection transient as a result of an

inadvertent RCP start because two operator actions are required to

start a pump. The requirement to have all RCPs de-energized, prior

to unisolating a loop adds additional assurance that an energy

addition transient will not occur.

The proposed change to allow 2 charging pumps to be operable

does not create an accident of a different type because there will

be adequate controls to ensure that the second pump does not

inadvertently start and initiate an increase in RCS inventory or a

boron dilution. Procedural controls will minimize the amount of time

that both charging pumps are operable and at no time will two pumps

be out of pull-to-lock.

The proposed footnote to TS 3.4.1.4.1 to remove RHR heat removal

from operation allows normally open valves to be closed in Mode 5 to

align RHR for ECCS injection. This introduces 'failure to open' as a

potential mechanistic failure malfunction in the RHR system. This is

a malfunction of a different type since previously stroking of these

valves was not needed to establish RHR. The current response to a

LOP is to manually restart the RHR pump only, with no valve

manipulations required. The proposed change adds the manual action

of realigning the valves. Since operator action to re-establish RHR

following a LOP is required independent of the proposed changes,

crediting operator action does not create the potential for a

malfunction of a different type. Allowing both trains of RHR to be

out of service does not create a different accident because

additional requirements have been specified for RCS loop operability

and at least one RHR pump is operable for ECCS when the core cooling

requirement is being met by crediting RCS loop operability. Meeting

the Mode 4 TS conditions prior to heatup, ensures two diesels are

operable. As such, a single failure would only require one valve to

be manually realigned in the ESF building. Adequate time is

available to accomplish these actions since this note only applies

during heatup, when decay heat is very low. Further, with four RCS

loops operable and a bubble drawn in the pressurizer and the RCS

pressurized, the steam generators can be used for core cooling via

natural circulation once the plant heats up to Mode 4, in the event

the RHR cannot be re-established. Since core cooling will be assured

if a LOP occurred during heatup in Mode 5, the change in plant

response to this event does not constitute an accident of a

different type.

The proposed changes to TS 3.5.3.f to manually realign the ECCS

valves is no different from what is currently evaluated. During a

Mode 4 LOCA adequate procedural guidance is provided to ensure that

RHR will be realigned for injection. The proposed change allows RHR

to be aligned to perform its safety grade cold shutdown heat removal

function.

Therefore, the proposed revision 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 new proposed curves raises the lower bound on RCS

temperature, resulting in increased RCS ductility and therefore

increased structural margin against non-ductile failure. The new

curves take into account the dynamic pressure effects identified in

NRC Information Notice 93-58 and are calculated in accordance with

10CFR50 Appendix G, ASME Section XI and Regulatory Guide 1.99,

Revision 2. These changes to the P/T [pressure/temperature] limits

are reflected in TS 3.4.9.1. Additional restrictions have been

placed on RCP operation to ensure that assumptions used in

developing the curves remain bounding. These are also reflected in

TS 3.4.1.3, 3.4.1.4.1, 3.4.1.4.2 and 3.4.1.6. As such, the curves

will continue to provide the required assurance for reactor vessel

integrity.

The COPS enable temperature is proposed to be lowered from the

current 350 deg.F to 275 deg.F which provides a margin of 31F above

that required by NRC Branch Technical Position RSB 5-2. The

reduction of the COPS enabling temperature eliminates the need for

COPS to be operable in Mode 3. This will simplify the transition

between Mode 3 and Mode 4.

Additional changes have been made to the Overpressure Protection

TS to ensure that the assumptions made in the Appendix G

calculations remain bounding. These include additional restrictions

on charging pump and SI pump operability and the modification of the

PORV setpoints. The pump requirements have been transferred from the

ECCS specification and expanded to cover Modes 4, 5 and 6. In

addition, these same pump restrictions have been included in TS

3.4.1.3 whenever RHR is in service. This provides added assurance

that the RHR piping will not be overpressurized by an inadvertent

actuation of an SI or charging pump. Additional actions and

surveillances have been provided to assure that assumptions on

charging pump and SI pump operability will be met. The additional

options for assuring the inoperability of the SI and charging pumps

require two distinct operator actions to restore injection

capability from these pumps. Thus, these options are equivalent in

providing assurance that an inadvertent injection will not occur

while at the same time allowing faster restoration if needed to

mitigate a loss of RHR.

A requirement to have all RCPs de-energized, prior to

unisolating a loop is added to TS 3.4.1.6.c, to ensure that loop

flow will not be initiated which results in an energy addition

transient from the secondary side of the SG being unisolated. This

change will preclude RCS overpressurization when an idled loop is

returned to service and SG secondary side temperature is greater

than the RCS temperature.

The PORV setpoints were established to ensure that the P/T limit

curves are not exceeded as a result of a single operator action or

as a result of a single equipment malfunction, as required by the

current system design basis criteria (i.e., SRP [standard review

plan] Branch Technical Position RSB 5-2).

A clarification of the hydrostatic and leak test requirements

ensures a uniform reactor vessel temperature for the test. A 72 hour

time limit is placed on the performance of engineering evaluations

of out of specification condition. This provides added assurance for

RPV [reactor pressure vessel] integrity.

The changes also eliminate an inconsistency between the charging

system operability requirements for boration and the charging system

operability requirements for cold overpressure protection. These are

reflected in TS 3.1.2.1, 3.1.2.2, 3.1.2.3 and 3.1.2.4. The Bases