Northeast Nuclear Energy Company; Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

NUCLEAR REGULATORY COMMISSION

[Docket No. 50-423]

Northeast Nuclear Energy Company; Notice of Consideration of

Issuance of Amendment to Facility Operating License, Proposed No

Significant Hazards Consideration Determination, and Opportunity for a

Hearing

The U.S. Nuclear Regulatory Commission (the Commission) is

considering issuance of an amendment to Facility Operating License No.

NPF-49 issued to Northeast Nuclear Energy Company (the licensee) for

operation of Millstone Nuclear Power Station, Unit 3, located in New

London County, Connecticut.

The proposed change to Technical Specification 3/4.4.3,

Pressurizer, would replace the pressurizer maximum water inventory

requirement with a pressurizer maximum indicated level requirement. The

proposed amendment would also make editorial changes and modify the

associated Bases section.

Before issuance of the proposed license amendment, the Commission

will have made findings required by the Atomic Energy Act of 1954, as

amended (the Act) and the Commission's regulations.

The Commission has made a proposed determination that the amendment

request involves 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. 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:

Northeast Nuclear Energy Company (NNECO) has reviewed the

proposed revision in accordance with 10 CFR 50.92 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 10 CFR 50.92(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 Technical Specification provides added restrictions

on pressurizer level to ensure that the pressurizer will not

overfill or empty in a transient and that RCS [reactor coolant

system] pressure control will

[[Page 20220]]

be maintained. The proposed Technical Specification requires

pressurizer level to be maintained at the programmed level. The

programmed level is a curve that varies linearly from 28% at no load

Tave to 61.5% at full power Tave. This is more

restrictive than the current upper limit of 92% of volume and

provides added assurance that pressurizer overfill will not occur

for those events where prevention of overfill is a criterion and

that the pressurizer would not empty due to a transient. In

addition, it assures that there is enough steam space available to

prevent RCS overpressurization in a transient. This requirement also

applies to manual operation to ensure that pressurizer level is

maintained in a band around the programmed level of +/-6% of full

scale. A two hour restriction on operation with pressurizer level

not within programmed level +/-6% of full scale has been added. This

will provide added assurance that operator error in pressurizer

level control will not result in a transient. Based on the above,

the changes do not negatively impact the probability of occurrence

of the previously evaluated accidents.

For Modes 1 and 2, the Chapter 15 FSAR [Final Safety Analysis

Report] accident analysis assumes that pressurizer level is being

maintained by the automatic control system at the programmed level.

For most of the accident analysis, pressurizer level is assumed to

be at 61.5% for power conditions and 28% for hot zero power. For

events where pressurizer level overfill is a concern, initial

pressurizer level is assumed to be 6% over the nominal value of

61.5% at full power. This bounds the automatic control system

uncertainty as documented in WCAP 14353. Thus, the proposed

Technical Specification LCO [Limiting Condition for Operation] for

Modes 1 and 2 is consistent with the Chapter 15 FSAR accident

analysis. When pressurizer level is being maintained by manual

operator action, a 6% operating band is specified. This band is

consistent with the 6% error assumed for the pressurizer overfill

events, but it does not take into account instrument uncertainty.

Because of the infrequent use of manual operation combined with the

multiple main board indications and the randomness associated with

instrumentation uncertainty, it is unnecessary to apply instrument

uncertainty effects on top of the operating band. As such, the 6%

band is bounded by the current Chapter 15 FSAR analysis. Thus, it is

concluded that the proposed Technical Specification is consistent

with analysis assumptions.

With regard to Mode 3 operation, an evaluation has been

performed for those events analyzed in Chapter 15 for Mode 3. The

only accident analysis provided in Chapter 15 of the FSAR for Mode 3

is the boron dilution event. Pressurizer level has no impact on the

results. As stated in the evaluation, the other events either would

not occur, or the plant response would be extremely slow or not

meaningful without power generation.

For Inadvertent Operation of ECCS [emergency core cooling

system] that Increases Reactor Coolant Inventory, the MP3 [Millstone

Unit 3] FSAR Section 15.5.1 clearly identifies this transient as an

event evaluated at Power Operation. This is consistent with SRP

[Standard Review Plan] Section 15.5.1-15.5.2 where the initial power

condition is specified as the licensed core thermal power with

allowance for measurement uncertainty. Thus, the current licensing

basis does not require analysis of this event for the shutdown

modes, including Modes 3 and 4.

Thus, the current specification which assures that a steam

bubble exists in Mode 3 is sufficient [ ] to ensure consistency with

the accident analysis assumptions.

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 Technical Specification changes provide tighter restrictions

on pressurizer level to ensure that pressurizer level will be

controlled as intended. The Bases change better reflects what

assures the validity of the accident analyses assumptions and the

bases for the maximum level. A two hour restriction on operation

with pressurizer level not within +/-6% (full scale) has been added.

This provides added assurance that pressurizer level will be

maintained consistent with the accident analysis initial condition

assumption. The changes provide added assurance that RCS pressure

control will be maintained and reduces the likelihood of pressurizer

emptying or overfill. These changes modify neither accident

mitigation nor system response post-accident.

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 Technical Specification changes provided are consistent with

the initial condition assumed in the Chapter 15 accident analysis by

placing tighter restrictions on pressurizer level. The Chapter 15

FSAR accident analysis assumes that pressurizer level is being

maintained by the automatic control system at the programmed level.

For most of the accident analysis, pressurizer level is assumed to

be at 61.5% for power conditions and 28% for hot zero power. For

events where pressurizer overfill is a concern, initial pressurizer

level is assumed to be 6% above the nominal value of 61.5% at full

power. This bounds the automatic control system uncertainty as

documented in WCAP 14353. Thus, the proposed Technical Specification

LCO for Modes 1 and 2 is consistent with the Chapter 15 FSAR

accident analysis. When pressurizer level is being maintained by

manual operator action, a 6% operating band is specified. This band

is consistent with the 6% error assumed for the pressurizer overfill

events, but it does not take into account instrument uncertainty.

Because of the infrequent use of manual operation combined with the

multiple main board indications and the randomness associated with

instrumentation uncertainty, it is unnecessary to apply instrument

uncertainty effects on top of the operating band. As such, the 6%

band is bounded by the current Chapter 15 FSAR analysis. For Mode 3,

the current specification which assures that a steam bubble exists

in Mode 3 is sufficient to assure consistency with the accident

analysis assumptions. The Bases are modified to reflect the proposed

changes and define the consistency with the Chapter 15 accident

analysis. Therefore, the change does not reduce the margin of

safety.

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.

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.

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 Administrative 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 6D59, 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

[[Page 20221]]

may be examined at the NRC Public Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC.

The filing of requests for hearing and petitions for leave to

intervene is discussed below.

By May 26, 1998, 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 located at the Learning Resources Center, Three Rivers

Community-Technical College, 574 New London Turnpike, Norwich,

Connecticut, and at the Waterford Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, Connecticut. 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

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 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: 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 Lillian M. Cuoco, Esq., Senior

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

Hartford, Connecticut, 06141-0270, attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for hearing will not

be entertained absent a determination by the Commission, the presiding

officer or the presiding Atomic Safety and Licensing Board that the

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

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 dated April 7, 1998, 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 located at the Learning Resources Center, Three Rivers

Community-Technical College, 574 New London Turnpike, Norwich,

Connecticut, and at the Waterford Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, Connecticut.

Dated at Rockville, Maryland, this 17th day of April 1998.

For the Nuclear Regulatory Commission.

Daniel G. McDonald Jr.,

Senior Project Manager, Special Projects Office--Licensing, Office of

Nuclear Reactor Regulation.

[FR Doc. 98-10843 Filed 4-22-98; 8:45 am]

BILLING CODE 7590-01-P