ML19308E184

Directs Appropriate AEC Personnel to Begin Review of Facilities.Div of Reactor Licensing Detailed Review Plan, Confirmed at AEC 670926 Meeting,Outlined.Review Schedule Encl

ML19308E184
Person / Time
Site:	Crystal River, 05000303
Issue date:	10/02/1967
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Grimes B, Ross D US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8003240721
Download: ML19308E184 (6)

Text

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.OCT 3 W Peter A. Morris, Director, DRL

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b DEL REVIEW AREAS FOR FIDRIDA POWER CORPORATION'S.CRISTAL RIVER UNITS 3 AND 4; DOCIET 108. 50-302 AND 303 i

A meeting was held on Tuesday, 26 September 1967, to confirm the DRL review plan for Crystal P.iver. Specific areas of vork for RP, RT, and 20 were identified. The meeting attendees were:

P. Morris C. Long L. Kintner R. Boyd D. Muller K. Knial

3. Lavine R. De Young D. Rcss Thic accrandua is to di cet ecT::cznad parties to proceed with the rvriew as cuti:.=ed herein.

Erought cat in the meeting was the fact that Crystal River is the third of a series (followinh Luke, Met-Ed) and that the review will be undicantly affected by review on the two previous applicaticnzs.

Ir. particular, answers to a sec of ouestions sent to ikc-Ed, to be retucced ca 0 Octcher, cust ba considered d: ring the cours: of the review. Satisfactary ansvers in the trea of prsseura vessel ther=al shcck and spectru=-of-break analyses, for example, vill notably decrease the review effort for Crystal River.

The detailed review plan is cutlined belev. The schedule for reviw follcus.

Note that the first technical casting is tentatively set for 31 October.

This requires preliminary questions, from RT and 30 to RP, no latar than 20 October in order that.a meaningful agenda may be prepared.

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RT will provide assistance to RP in the following areas:

1.1 Site and Envitortment.

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1.1.1' Coordinate efforts of external consultants (excluding

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Newsark).

1.1.2 Calculate Part-100 do=ca for design basis accident.

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1.1.3 'Raview deposition into and uptaka by marina life in Culf

..of Mexico for normal and emergency relasses.

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1.1'. 4 Review diffusion calculation including effect of wake of building and acJe of release (volume vs point).

1.1.5 Review pcpulation distribution, forecasts for plant lifc-time.

1.2 Pressure vessel 1.2.1 Review answer by Met-Ed regarding thessal shock to the pressure vassal. Coemeent on calculational methods, bases for acceptable material response, and other areas as appear necessary.

1.2.2 Review answer by Met-Ed concerning structural adequacy and general design of core-barrel check valve.

1.2.3 Provida ce==cet en the ability of the ccre barrel check valve (s) to serva as a core bypass p&th Icr a cold-1q, break (cce 1.3.3 eelow).

1.2.4 In secticu 2.3 belov, it is noted that RP will raview tha adaqua.r of the pressure vassal relative to the Te.ntative Supplementary Criteria (dated 23 Aer.st). HowevGr, 2""

will provido assistanca if and as t--. ested, folleving; the P.? review.

1.3

nginaeud Safety 7aaturac 1.3.1 Provids co=nent on preposed chc=ical spray system far i

Icdine scavansin;;.

In this regard, perform e c-:.e:parative review of fis dcoign as compared with Wstin3 ouse.

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compatibility of chemical with concerned caterials, respcuse of chemical to temperature and radiatien, tandency of chem-ical to platsout during long-term storage, and effectiveness of chemical as an iodine scavenger.

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1.3.2 Provide cecuent on the Het-Ed answer concerning heat transfer coefficients in the core during recoverv, folleving the I.CCA.

j 1.3.3 C =sent en the applicability c" thz fc11oving' probicm:

During a cold-leg-braa's blowdown, tha core barrel check valve (s) open and permit some or all of the reversed flow to bypass the core. Thus for the first few ( ~ 10) seconds the core ficw may be relatively stagnant. Withcut blowdown heat transfer the clad temperatures may beccee significantly himber durine subseeuant charms of the toc.L Dk"*oA!%

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g 3. W NOTE: This problem was postulated to 'the applicant at-the general meetin5 on 26 september. It is reasonable to assume that, if this is a significant probles, it will be addressed formally to the appli-cant at the appropriate ti:ne. Th: cxtent of F.T assistance is therefore divided as follows:

l (a) Between now and 20 October provide guidance as to the severity of the problem. Subsequently, (b) comment on the adequacy of any detailed reply to a formally-addressed question.

Analysis creas includa predictad bot-spot temperatures, metal-water reactions, number of failed elements, and decrassed flow area following clad ballooning.

1.4 Accident f.n:1rnin 1.4.1 Analyse and/or cccac2ent on the following accident postulation:

The aheft en one pri: nary pump shears while the raccter is

100". pc tar and flev. Tha antiretatica devica on the drivo chaft deas not function. "everse fic" w*tes through the pu=p.

(Mo:c: Thero arc. c.c davices propccad to take auto-catic p m r level reductica naps. *Ceus this accident caounts to cu erbitrar/ ficu redaccion with no az.tcndant correction. See 1.5.1 below for further review areas.)

1.4.2 Review the adquacy of the '.,'estinp, house turb*~ "op valves as stea:2 lina isolation valves.

(Note: We have been infor=ed, informally, that scrae turbine stop valve redesign is being contemplated.)

Comment on response of primary system to various steam-line breaks. Comment on adequacy of safety valves' design for actuation following turbine trip; that is, for step lead reductions in en:cas of 4C7. the sofory valves must lif t.

1.5 Instrw-mtstion and cea.:rol 1.5.1 Eavisv tha instrumentation and centrol syste:ss in the follow-ing araas:

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oms-shaft.

protection against Ices-of-flow, including sheared-m D91R DJ e) o

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~(2) response 'of suermd-neutron-power devices (input' to power / flow) to various shorts, opens (3) adequacy of aanaing devices for initiation of action of engineered safety features. - Give coexsents on possible diversity of sensors, such as low pres-suriser level, high water level'in primary cavity, etc.

(4) adequacy'of bypass design on ECCS actuating devices (for normal plant shutdown).

II. Rsactcr Projacts will review the folleving are n :

2.1 Adequacy of pressure vaasel-design as compared to Tentative Supple-centary critoria.

2.2 Ader;uccy of ECOS desigs as coccared ta ;;uppicacutary Criteri.2.

2.3 Mcquacy of auxiliary eicctric powr desin es compared to Supple-weatary Criteria.

2.4 Adequcy o f plan t docip ss e-i,W

.2 70 cenu et T e en criteri:.

2.5 Shcred fnatures of Usica 3 cd 4 (and pesaibly fo-wil-toed Unit:

1 md 2 as related to manning) lacluding nuclew facility manage-sent. chamical makeup, vasta ecilecticn and disposal, spen: fuel storagc.

2.6 tumteary review of op.2rator actions follovir.3 a LOCA; ad recomen-dationa for autcoation where indicated.

2.7 Application of ACRS connents on cake, including:

t f '(1) inspectability of primary system (2) quality assurance (3) positive moderator coefficiant (4) fuct rod rsspon:o to arpected transienta (5) xenon oscillatica (6) centsi=stne liner inspectability.

2.8 Miscc11er. acus.

(Sea ne=o, C. C. I,cng to F. A. Morris on this subject dated 15 Septenber 1967.)

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. III._ Reactor Operatiema will reviev the fo 'owing : areas t -

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n-3.1 Technical caspetency of the applicant, Florida Power Corporation, as owner, operator, and construction manager.'

3.2. Technical coc:petancy of construction contractor.t (identity not yet specified).--

i 3.3 Emergency plans for Units 3 and 4.

l IV. The following areas will not be reviewed in depth. These designations

'1 are specified on the assumption'that design is essentially that of Duke and/or Mat-Ed 1

4.1 Primary Coolant System 4.2 control and Instruaentatics Systeu 4.3, Containment Design 4.4 Coro Desi p A revicu schadula is attached.

Dis trihetton:

Suppl.

Assistar.t Directers EP Ur. inch Chiefa RT Branch Caiaf:

RP3 d3 Recding D. F. Ross

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50-302':.303 c..

CRYSTAL RIVER UNITS 3 AND 4 ELORIDA~ POWER CORPORATION A.

Application Filed 10 August 1967 B.

Meetings-1.

General 26 September 1967 2.

Technical

• 31 October 1967 C.

Questions 14 November 1967 D.

Answers 21 December 1967 E.

Consultant's_ Report 14 December 1967 F.

Meeting 4 January 1968 G.

ACRS Site Visit January 1968 E.

ACRS Report 19 February 1968 I.

ACRS Subeccmittee 6 March 1963 J.

ACRS Meeting 7, 8 March 1968

1. l I

K.

Safety Evaluation 5 April 1968 l

Schedule based upon:

1.

Updating of-application to incorporate appropriate safety recom=enda--

I tions from Duke, Met-Ed reviews.

2.

Conducting review as another of the same generation.as Duke, Met-Ed.

.* Preliminary questions from RT, RO into RP on or before 20 October.

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