ML20134A599

Notice of Consideration of Issuance of Amend to License DPR-65 & Proposed NSHC Determination & Opportunity for Hearing.Amend Authorizes Licensee to Increase Spent Fuel Pool Storage Capacity from 667 to 1,112 Storage Locations

ML20134A599
Person / Time
Site:	Millstone
Issue date:	10/29/1985
From:	Butcher E Office of Nuclear Reactor Regulation
To:
Shared Package
ML20134A589	List: ML20134A587 ML20134A599
References
TASK-2.E.4.2, TASK-TM TAC-44864, NUDOCS 8511070421
Download: ML20134A599 (14)

Text

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7590-01 UNITED STATES NUCLEAR REGULATORY COPNISSION NORTHEAST NUCLEAR ENERGY COMPANY, ET AL.

DOCKET NO. 50-336 NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE AND PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION AND OPPORTUNITY FOR HEARING

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The U. S. Nuclear Regulatory Comission (the Comission) is considering issuance of an amendment to Facility Operating License No. DPR-65, issued to Northeast Nuclear Energy Company (the licensee), for operation of the Mi}.lstone

,, Nuclear Power Station. Unit 2. located in New London County, Connecticut'.'

The amendment would authorize the licensee to increase the spent fuel pool storage capacity from 667 to 1112 storage locations. The proposed expan-sion is to be achieved by reracking the spent fuel pool with a combination of poison racks and non-poison racks in a two-region arrangement.

_, Region I consists of two 8 x 9 modules and three 8 x 10 modules and would store high-enrichment, core off-load assemblies. The region consists of poisoned spent fuel racks with a nominal center-to center cell spacing of 9.8 inches.

Fuel assemblies would be stored in every location. The five modules of Region I total 384 storage locations and are designed to accomodate 1.7 reactor cores of high enrichment nuclear spent fuel. -

The spent fuel rack design for Region I is based upon the comonly accepted physics principle of a " neutron flux trap" with the use of neutron absorber materials. The racks are designed to store Millstene 1.4 x 14 fuel with an initial enrichment of 4.5 weight percent U-235. The poison material to be used is Boraflex.

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1 Region II consists of 14 modules of non-poisoned spent fuel racks with  !

nominal center-to-center cell spacing of 9.0 inches. The modules consist <

of 962 cells with useable capacity of 728 storage locations.

l Region II is reserved for fuel that has sustained at least 85% of its design burn-up. The spent fuel rack design is based on criticality acceptance criteria specified in Revision 2 of Regulatory Guide 1.13 which allows credit l for reactivMy depletion in spent fuel. (Previously, the physics criteria for fuel stored in the spent fuel pool.were defined by the maximum unirradiated

) initialenrichmentofthefuel). Fuel assemblies are stored in a three-out-of-

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four logic pattern. The fourth location of the storage configuration remains empty to provide the flux trap to maintain the required reactivity cont'rel.

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2 Blocking devices will be used to prevent inadvertent placing of a fuel assembly in the fourth location.

i The spent fuel racks in both regions are fabricated from 304 stainless j

, steel which is 0.135 inches thick. Each cell is formed by welding along the j intersecting seams. This enables each spent fuel rack module to become a free-standing module that meets the seismic design requirements without j mechanical dependence on neighboring modules or fuel pool walls for support.

• The rack modules are classified ANS Safety Class III and Seismic Category I.

I Both regions of the spent fuel pool have been designed to store fuel assem-blies in a safe, coolable, subcritical configuration with K,ff less.,than or j equal to 0.95.

The racks have been designed and will be provided by Combustion Engineering,

Inc.(CE). CE racks of this type have been most recently licensed by the NRC l i for use at Florida Power and Light Company's St. Lucie Plant and at Arizona I l '

I Public Services Company's Palo Verde nuclear plants. This amendment was i

requested in the licensee's application for emendment dated July 24, 1985.

The additional assemblies that can be stored will have a lower heat generation rate and radioactivity content than the assemblies currently

! allowed to be stomd. However, the increase in the total num'er of assemblies that can be stored will increase the total fuel pool heat load

, and radioactivity content but only by a small amount. The replacement spent

- fuel storage rack modules are freestanding without depending on neighborings modules or the fuel pool walls for support. Racks'of similar design have -

been licensed at other nuclear facilities. The use of two diverse regions is not unique and two region spent full pools have been previously appr.oved s

by the Comission. . .

The technical evaluation of whether or not an increased spent fuel

, pool storage capacity involves significant hazards consideration is centered on three standards:

{ A. First Standard

-- Involve a significant increase in the probability or consequences of- an accident previously evaluated.

- -Th'e licensee's safety analysis of the proposed reracking has been accom-plished using current NRC Staff accepted Codes and Standards. The results of the safety analysis demonstrate that the proposal meets the specified accept-ance criteria set forth in these standards. In addition, the licensee has

' reviewed NRC Staff SE for prior spent fuel pool rerackings involving spent fuel pool rack replacements to ensure that there are no identified concerns not fully addressed. The licensee has identified no such concerns.

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l The licensee has identified the following potential accident scenarios:

(1) spent fuel cask drop; (2) loss of spent fuel pool forced cooling; (3) seismic event; (4) spent fuel assembly drop; (5) criticality accident; and I

(6)LoadHandlingAccident. The probability of the occurrence of any of the first four listed accidents is not affected by the racks themselves; thus, reracking cannot increase the probability of these accidents.

!, All potential events which could involve accidental criticality have been

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l examined in the licensee's safety analysis. It was concluded that the bounding -

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accident was dropping an unirradiated fuel assembly into a blocked fourth location in Region II. The probability of dropping a fuel assembly during e .-

fuel movement operations is not affected by the fuel storage racks.

, The proposed Millstone Unit 2 spent fuel pool reracking will not involve

an increase in probability of any previously evaluated load handling accident i as accepted standards and procedures will be utilized as described in the licensee's safety analysis.

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The consequences of the spent fuel cask drop accident have been evaluated f as described in Sections 5.4 and 9.8 of the Millstone. Unit 2 Final Safety halisisReport(FSAR). By controlling the decay time for fuel stored within a specified distance from the cask set-down area to not less than 120 days prior to cask movement together with an administrative control specifying a minimum

! required boron concentration in the water of the spent fuel pool, t'he consequences i

l of this accident type will remain well within 10 CFR Part 100 guidelines.

i There is, however, an increase in the value of the 2-hour whole body dose at the site exclusion boundary for a postulated cask drop accidqnt. The new racks increase the storage density of spent fuel within the distance L of the

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i cask set-down area. This results in a calculated increase of the 2-hour ,

whole body dose from 140 millirem to 240 millirem, an increase of 100 milli-  ;

t rum. In review of this submittal, the licensee has recognized this increase and has designated it an unreviewed safety question. The calculated dose  ;

5 is well within the guidelines specified by 10 CFR Part 100 and, as such, the consequences of this type of accident will not be significantly increased ,

l from previously evaluated events. '

- . s The consequences of the loss of spent fuel pool forced cooling accident i

! have been evaluated and are described in the licensee's safety analys'is. There is ample time to effect repairs of the cooling system or to establish makeup j

flow to the spent fuel pool. Theconsequencesofthistypeaccidentwi1hnot

) be significantly increased from previously evaluated accidents by this proposed i reracking.

l i The consequences of a seismic event have been evaluated against the i

l - appropriate NRC standards. The results of the seismic and structural analysis i

show that the proposed racks meet all of the NRC structural acceptance criteria and are consistent with results found acceptable by the NRC Staff in previous hisonrerackSEs. Thus, the consequences of seismic event will not significantly

increase from previously evaluated seismic events.

! The consequences of a spent fuel assembly drop accident are described in  :

. i l Section 14.19 of the Millstone Unit 2 FSAR. A complete list of assumptions is i providad in FSAR Table 14.19-1. Results of the analysis are well below the limits of 10 CFR Part 100 and are presented in Section 14.19.3. The consequences j l

of this type accident will not be significantly increased from previously evalv-ated accidents by this proposed reracking.

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i The consequences of a criticality accident have been evaluated for all potential events which could involve accidental criticality. The bounding criticality accident was found to be the dropping of a fresh fuel assembly into a blocked fourth location in Region II. Administrative controls in the form of a Technical Specification of minimum boron concentration for the water of the spent fuel pool will preclude the bounding criticality accident; therefore, the consequences of this type accident will not be significantly in-

- creased from previous accident evaluations by this, proposed reracking..

i The consequences of a load handling accident have been evaluated. The .

work to be done in the spent fuel pool will be perfonned in accordance with

- accepted construction practices, standards, and procedures. The conseq'uences of this type act.ident will not be significantly increased from previous accident evaluations by this proposed reracking. Therefore, it is shown that the pro-posed Millstone Unit 2 spent fuel rack replacement will not involve a signifi-l cant increase in the probability or consequences of an accident previously evaluated.

B. Second Standard

"- ' Create the possibility of a new or different kind of accident from any accident previously evaluated.

l The licensee has evaluated the proposed rack replacement in accordance with the "NRC Position for Review and Acceptance of Spent Fuel Storpge and Handling Applications." appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plan sections, and appropriate industry Codes and Standar6.

In addition, the licensee has reviewed the NRC SE for the previous Millstone Unit 2 spent fuel rack replacement application and for other prior spent fuel l

pool rerackings.

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The change to a two-region spent fuel pool creates the requirement to perfom additional evaluations to ensure the criticality requirement is maintained. These include the evaluation of the limiting condition (dropping a fresh fuel assembly into a blocked fourth location in Region II). This i

evaluation shows that, when the boron concentration requirement is met per the proposed Technical Specifications, the criticality criterion is satisfied.

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Although this change does create the requirement to address additional aspects

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of a previously analyzed accident, it does not create the possibility c'f a pre- ,

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) viously unanalyzed accident.

C. Third Standard ,

Involve a significant reduction in a margin of safety.

f The issue of " margin of safety," when applied to a spent fuel I

rack replacement, includes the following considerations:

l a. Nuclear criticality considerations.

b. Themel hydraulic considerations.

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c. Mechanical, material, and structural considerations.

i j The margin of safety that has been established for nuclear criticality is that the neutron multiplication factor (K,ff) in the spent fuel pool is to be less than or equal to 0.95, including all uncertainties, under all conditions.

For the proposed modification, the criticality analysis is described in the licensee's safety analysis. The methods utilized in the analysis c'enfom with f  ;

! ANSI N210-1976, " Design Objectives for LWR Spent Fuel Storage Facilities at

( Nuclear Power Stations"; ANSI N16.9-1975, " Validation of Calculational Methods 1 for Nuclear Criticality Safety"; the NRC guidance, "NRC Position for Review

! and Acceptance of Spent Fuel Storage and Handling Applications" (April 1978),

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as modified (January 1976); and Regulatory Guide 1.13. " Spent Fuel Facility '

Design Basis," proposed Revision 2. The computer programs, data libraries, and benchmarking data used in the evaluation have been used in previous

! spent fuel rack replacement applications by other NRC licensees and have been reviewed and approved by the NRC. The results of the licensee's analy-I sis indicate that K,ff is less than or equal to 0.95 under all postulated con-

- ditions, including uncertainties, at a 95/95 probability / confidence level.

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l Thus, meeting the acceptance criteria for criticality, the proposed rer'ackin -

does not involve a significant reduction in the margin of safety for nuclear 5 criticality.

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For themel hydraulics, the relevant considerations for evaluating 'If

, there is a significant reduction in margin of safety are: (1)maximumfuel j t

. temperature, and (2) the increase in temperature of the water in the pool.

! The licensee's themal hydraulic evaluation shows that fuel cladding tempera-  :

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! tures under abnormal conditions are sufficiently low to preclude strvetural l ,,

failure and that boiling does not occur in the water channels between the

! fuel assemblies nor within the storage cells. However., the proposed rack

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replacement will result in an increase in the maximum heat load in the l

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l Millstone Unit 2 spent fuel pool. The licensee's safety analysis shows j that the maximum temperature will not exceed the current margin of safety  ;

} .(150"F). For the maximum nomal heat load case (full-core discharg's at i

f 150 hr after shutdown, which fills the spent fuel pool to its capacity),

j the pool temperature will not exceed 150'F. Thus, there is no significant reduction in the margin of safety from a themel hydraulic standpoint or f from a spent fuel pool cooling standpoint.

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I l a I The mechanical, material, and structural considerations of the proposed  !

rack replacement are also analyzed in the licensee's safety analysis. The

racks are designed in accordance with the applicable NRC Regulatory Guides, Standard Review Plan sections, and position papers, and appropriate industry j Codes and Standards, as well as to Seismic Category I requirements. The

-l materials utilized are compatible with the spent fuel pool and the spent fuel assemblies. The conclusion of the analysis is that the margin of

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safety is not significantly reduced by the proposed reracking. -

- In sumation, it has been shown that Northeast Nuclear Energy Company's~

a proposed spent fuel storage facility modifications and proposed technical i .- l j specifications do got: o

1. Involve a significant increase in the probability or con- I l

j sequences of an accident previously evaluated; or

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2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

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

i Because the licensee's submittal and the above discussion by the licensee 1ppear 'to demonstrate that the standards specified in 10 CFR 50.92 are met, and l because reracking technology has been well developed and demonstrated the i

Comission proposes to determine that operation of,the facility in accor-

dance with the proposed amendment does not involve a significant hazards i consideration.

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The Comission is seeking public coments on this proposed dt "e'emination.

Any coments received within 30 days after the date of publication of this notice ,

i will be considered in making any final detennination. The Comission Will not

normally make a final determination unless it mceives a request for a hearing.

1 Coments should be addressed to the Rules and Procedures Branch, Division of Rules and Records, Office of Administration, U.S. Nuclear Regulatory

, , Comission. Washington, D.C. 20555.

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l By Dec. 4,1985 . the licensee may file a request for a hearing with '

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, respect to issuance of the amendment to the subject facility operating license L

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and any person whose interest may be affected by this proceeding and who wishes i .. -

to participate as a party in the proceeding must file a written petition for leave to intervene. Request for a hearing and petitions for leave to intervene  !

, shall be filed in accordance with the Comission's " Rules of Practice for

Domestic Licensing Proceedings" in 10 CFR part 2. If a request for a hearing i

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

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or an Atomic Safety and Licensing Board, designated by the Commission or by l the Chairman of the Atomic Safety and Licensing Board , panel, will rule on the J 'Tequest and/or petition and the Secretary or the designated Atomic Safety and i

m Licensing Board will issue a notice of hearing or an appropriate order.

As required by 10 CFR 52.714. a petition for , leave to intervane shall j 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 i

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

! shouldalsoidentifythespecificaspect(s)ofthesubjectmatterofthepro-

! ', ceeding as to which petitioner wishes to intervene. Any person who has filed, a petition for leave to intervene or who has been4dmitted as a party.may amend thepetitionwithoutrequestingleaveoftheBoarduptofifteen(15)daysprior ,

to the first prehearing conference scheduled in the proceeding, but suc,h an

. amended petition must satisfy the specificity requirements described above.

Not later than fifteen (15) days prior to the first prehearing conference scheduled in the proceeding, a petitioner is required to 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, and the bases for each contention set

_. forthwithreasonablespecificity,pursuantto10CFR52.714(b). Contentions

shall be limited to matters within the scope of the amendment under considera-2 tion. A petitioner who fails to file such a supplement which satisfies these

! c requirements with respect to at least one contention will not be pemitted to participate as a party. .-

The Comission hereby provides notice that this proceeding is on an

, application for a license amendment falling within the scope of Section 134 I of the Nuclear Waste Policy Act of 1982 (NWPA), 42 U.S.C. 610154. Under i Section 134 of the NWPA, the Connission, at the request of any petitioner or party to the proceeding is required to employ hybrid hearing procedures

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with respect to "any matter which the Comission detemines to be in contro-versy among the parties." Section 134 procedures provide for oral argument on those issues "detemined to be in controversy", preceded by discovery under the Rules of Practice, and the designation, following argument, of only those factual issues that involve a genuine and substantial dispute, together with 1

any remaining questions of law to be resolved at an adjudicatory hearing.

i Actual adjudicatory hearings are to be held only on those issues found to meet i

the criteria of Section 134 and set for hearing af,ter oral argument on the ' ,

j j proposed issues. However, if no petitioner or party mquests the use of the' hybrid hearing procedures, then the usual 10 CFR Part 2 procedures apply.

(At this time, the Comission does not have effective regulations'iNplemen-l ting Section 134 of the NWPA although it has published rules which became effec-l tive dovember 14, 1985. See Hybrid Hearing Procedures for Expansion of Spent Fuel Storage Capacity at Civilian Nuclear Power Reactors, 50 FR 41662 (October 15,

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1985).

Subject to the above requirements and any limitations in the crder granting l

j leave to intervene, those pemitted to intervene become parties to the proceeding wd 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 Connission will make a final detemination on the issue of no significant hazards consideration. The final detemination will serve to decide when the hearing is held.

If the final detemination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment l

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and make it effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.

If the final detemination is that the amendment involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

v Nomally, the Comission will not issue the amendment until the expira-l i

, tion of the 30-day notice period. However, should circumstances change during  !

s the notice period such that failure to'act in a timely way would result, for - '

example, in derating or shutdown of the facility, the Comission mai issue the license amendment before the expiration of the 30-day notice period. provjded .

that its final detemination is that the amendment involves no significalit hazards consideration. The final detemination will consider all public and State coments received. Should the Comission take this action, it will

! pubitsh a notice of issuance and provide for opportunity for a hearing after issuance. The Comission expects that the need to take this action will occur very infrequently.

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

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j with the Secretary of the Comission, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, Attention: Docketing and Service Branch, or may be

delivered to the Comission's Public Document Room 1717 H Street, N.W.,

Washington, D. C., by the above date. Wherepetitionsarefiledddringthe last ten (10) days of the notice period, it is requested that the petitioner promptly so inform the Comission by a toll-free telephone call to Western Union at (800) 325-6000 (in Missouri (800) 342-6700). The Westgrn Union

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operator should be given Datagram Identification Number 3737 and the

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I following message addressed to Edward J. Butcher: petitioner's name and telephone number; date petition was mailed; plant name; and publication I

date and page number of this FEDERAL REGISTER notice. A copy of the

! petition should also be sent to the Executive Legal Dimetor U.S. Nuclear Regulatory Cossnission Washington, D.C. 20555, and to Gerald Garfield, i

Esq., Day, Berry and Howard, One Constitution Plaza Hartford, Connecticut l- 06103, attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended pe$itions, I

supplemental petitions and/or requests for hearing will not be entertained

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absent a determination by the Consnission, the presiding officer or the pre- .,

siding Atomic Safety and Licensing Board, that the petition and/or reque'st i

should be granted based upon a balancing of the factors specified in 10 CFR 2.714(a)(1))i)(v)and2.714(d).

{ For further details with respect to this action, see the application for

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amendment that is available for public inspection at the Connission's Public

_. Document Room, 1717 H Street, N.W., Washington, D.C., and at the Waterford Public Library, 49 Rope Ferry Road, Waterford, Connecticut 06103.

Dated at Bethesda, Maryland, this 29 day of October,1985.

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! FOR THE NUCLEAR REGULATORY C009tISSION i

• h Edward J. Butcher, Acting Chief '-

! Operating Reactors Branch # 3 i

Division of Licensing l

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