ML19345H193

Revised SECY-81-245,recommending Commission Approval of Fr Publication of Encl Interim Amends to 10CFR50 Re Hydrogen Control & Degraded Core Considerations.Fr Notice & Summary of Public Comments Encl

ML19345H193
Person / Time
Issue date:	04/17/1981
From:	Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
Shared Package
ML19345H189	List: ML19345H188 ML19345H193 ML19345H197 ML19345H203
References
FRN-45FR65466, REF-10CFR9.7, RULE-PR-50, TASK-RIA, TASK-SE SECY-81-245, SECY-81-245-R01, NUDOCS 8105010195
Download: ML19345H193 (95)

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s-g April 17,1981 SECY-81-245 RULEMAKING ISSUE (Affirmation) t For:

The Commissioners From:

William J. Dircks, Executive Director for Operations

Subject:

INTERIM AMENDMENTS TO 10 CFR PART 50 RELATED TO HYDROGEN CONTROL AND CERTAIN DEGRADED CORE CONSIDERATIONS l

l

Purpose:

To obtain Commission approval for publication of the final amend-ments in the Federal Register.

Category:

This paper covers a major policy question.

Issue:

(a) Whether licensees of operating nuclear power plants should be required to provide inerted containment atmospheres for Mark I and II BWRs; (b) Whether licensees of Mark III BWRs and PWR Ice Condenser facilities should be required to provide hydrogen control i

systems that can handle large amounts of hydrogen; and (c) Whether the regulations should be revised to incorporate changes specifically identified in the TMI Action Plan to help mitigate the consequences of a degraded core accidert.

Discussion:

During the Policy Session on September 4, 1980, the Commission approved for publication in the Federal Register a notice of pro-posed rulemaking (contained in SECY-80-399) concerning interim requirements related to hydrogen control and certain degraded core considerations.

The notice of proposed rulemaking inviting public comments (Enclosure "A") was published in the Federal Register (45 FR 65466) on October 2, 1980 with tb comment period scheduled to expire on November 3, 1980.

Contact:

M. R. Fleishman 443-5921 l

SECY NOTE:

This paper is scheduled for a briefing at an open Connission meeting on Thursday, April 23, 1981.

l 8105010195 3

1 The Comissioners 2

In response to the notice of proposed rulemaking, coments were submitted by 35 persons having the following grouping by affiliation:

Nuclear Steam System Supplier 4

Utility 18 Architect / Engineer 3

Law Firm 1

Individual 4

Nonprofit Society or Association 3

Government Agency (Local, State or U.S.)

2 A detailed sumary of the comments is provided in Enclosure "B" including a listing of the comenters, a categorization of the coments, a paraphrase of each of 160 coments, a grouping of the coments into 65 unique comments and finally a table indi-cating the relationship between the 160 individual coments and the 65 unique coments.

The comenters were about equally divided between those in favor of and those opposed to publishing the interim amendments.

The following represent a distillation and paraphrasing of the more significant coments:

1.

The various rulemakings currently being pursued by NRC snould be integrated, i.e., safety goal, degraded core con-siderations, minimum engineered safety features, siting, and emergency plannilg.

Resolution:

The staff agrees.

A Degraded Cooling Steering Group was established with, among other things, the responsibility for coordinating all r,iemaking actions related to the sub-ject of degraded cooling.

This group has completed its work and proposed a plan to ensure future integration of these ar. tivi ties.

2.

Many of the dates specified for impiementatior, oy the rule cannot be met for a variety of reason, such as procurement lead time, need for the design studies, availability of acceptable equipment, etc.

Resolution:

The staff agrees.

The dates have been revised to be con-sistent with NUREG-0737, " Clarification of TMI Action Plan Requirements" which includes industry feedb:.ck from regional meetings and responses to earlier NRC orders.

The new dates are believed to be a reasonable compromise (sse major change number 5 below).

1 The Commissioners 3

3.

The interim rule should be delayed until NRC sets policy regarding hydrogen control, degraded core cooling, etc.,

such as will be done during the long term rulemaking.

Resolution:

The staff disagrees.

The items covered by the rule are believed to be of such safety significance that they should be codified immediately.

The rule will also serve as the basis for upcoming regulatory actions and will help to expedite the licensing process.

4.

In view of the great detail involved, the changes should be implemented using other approaches, such as by providing guidelines set out in regulatory guides, rather than by regulation.

Resolution:

The staff disagrees.

In view of the seriousness of the safety issues involved and the need to implement changes quickly, it was felt to be more appropriate and enforceable to require these changes by formal regulation.

Additional guidance on implementation of the rule is provided in NUREG-0737.

5.

The bent: fits of inerting are questionable, and options other than inerting should be permitted by the rule.

Resolution:

The staff disagrees.

The issue has been extensively reviewed and discussed with the Commission and the staff's position, described in the SECY-80-107 series, remains basically un-changed.

That is, that the Mark I and II BWRs should be provided with an inerted atmosphere during normal opera-tions.

However, the revised rule does allow other options for Mark III BWRs, ice condenser PWRs, and remaining PWRs, depending upon the results of design st,udies.

6.

The hydrogen control design studies should not be required because it is inappropriate in a regulation and because unambiguous event descriptions and acceptance criteria are not available.

Resolution:

The staff agrees with these comments in part. As a result, the rule has been modified to clarify the types of analyses required. They can be grouped into four l

classes depending on the containment design. The results w.

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s The Commissioners 4

will support both the long-term rulemaking on degraded cores and the hydrogen control system selection for Mark III BWRs and ice condenser PWRs.

7.

The interim rule is not flexible enough to allow for plant-to plant va iability.

Resolution:

The staff disagrees.

Although the rule is somewhat detailed, it is still general enough to account for plant-to plant variability. The rule will be complemented by NUREG-0737 and Regulatory Guide 1.97 which will provide the additional guidance that is needed to account for the plant-to plant variability.

8.

The rule covers, to a large degree, previous, less formal agreements, made between NRC and licensees as well as earlier letters to licensees, and should be made consistent with them.

The rule should have provisions to prevent nullifica-tion of previous compliance agreements.

Resolution:

The staff basically agrees.

The rule has been revised to be consistent with NUREG-0737 which clarifies the TMI Action Plan requirements.

NUREG-0737 includes modifications from earlier letters and NUREG reports in order to account for compliance agreements previously reached.

9.

Alternates to the use of external recombiners should be permitted.

l l

Resolution:

The staff agrees and in fact interpreted the previous version of the rule as permitting this.

However, in the interests of clarity, the rule has been modified to indicate that inter-nal recombiners are an acceptable alternative (see major change number 4 below).

10.

Reliable reactor vessel water level indicators are not commercially available and approved by the NRC.

Resolution:

The staff disagrees.

The rule has been modified, however, to indicate that for PWRs, the water level indicators augment the incore thermocouples; and for BWRs, the incore thermo-couples augment the water level indicators.

s s

The Commissioners 5

11.

The interim rule should state that it will be the basis for licensing and may not be challenged in individual licensing proceedings.

Resolution:

The staff basically agrees.

Words have been added to the Supplementary Information section of the Federal Register notice to indicate that "the Commission intends this interim rule to provide the needed basis for regulatory actions that I

cover licensing and continued operation of nuclear power plants."

It is also indicated that the items covered by the rule are intended to be removed "from litigation in individual proceedings."

These and all other suggestions from the commenters were reviewed and considered by the staff in preparing the final rule.

The comments did not reveal any major technical questions, considera-tions, or arguments not previously considered by the staff during development of the rule.

However, the final rule included in the Federal Register notice (Enclosure "C") incorporates changes that reflect the comments received.

The regulation has been printed in comparative text for ease in identifying the changes.

The major changes to that originally proposed are as follows:

1.

An entirely new paragraph has been added which requires Mark III BWR and Ice Condenser PWR containments to be provided with a hydrogen control system that can handle an amount of hydrogen equivalent to that from up.to about a 75% fuel cladding-water reaction.

2.

Another new paragraph was added to require all BWRs and PWRs in which hydrogen could be generated, and which do not rely upon on inerted atmosphere, to ensure that systems needed.

l for safe shutdown and containment integrity can perform their functions during and following hydrogen burning.

3.

For the hydrogen control analyses that are being required, the rule has been revised to limit the scope of the analyses so as to only justify the hydrogen control system selection and the demonstration of safe shutdown and continued integrity.

l 4.

Internal recombiners have been specifically identified as an acceptable alternative to external recombiner capability for long term hydrogen control in the containment.

5.

The various implementation dates have been reconsidered and many have been revised in order to be consistent with the i

3 The Comissioners 6

agreements reached in conjunction with the issuance of NUREG-0737. Additionally, some technical changes have also been made to achieve greater consistency with NUREG-0737.

OELD has raised concerns about the new requirements contained in major changes 1 and 2.

These concerns are stated in Enclosure "E".

If the design analysis requirement is extended to all BWRs (major change 3) OELD would have the same concern as in Enclosure "E".

Consistent with discussions with the Comission in connection with the development of NUREG-0737 schedules, the staff is concerned that some of the schedules in this rule may pose an unnecessary burden on some utilities which could have an adverse safety impact.

The staff is presently developing information on the need for relaxation of some of the dates and will be prepared to discuss possible approaches at the time of the Commission meeting on this paper. The staff will specifically consider previous Comission coments including the need to consider impact on power reliability.

Recomendation:

That the Comission:

1.

Approve the publication of the final amendments, as set forth in Enclosure "C", which would require the inerting of Mark I and II BWR containments, and hydrogen control systems for Mark III BWRs and ice condenser PWRs and the implementation of certain design and other changes to help mitigate the consequences of degraded core accidents.

The Comission should take note of the fact that 0 ELD has recomended (Enclosure "E'?) that the two new requirements l

contained in major changes 1 and 2 be issued as a separate l

proposed rule. The Office of Nuclear Regulatory Research agrees with OELD.

However, NRR prefers that these two items be part of this final rule, and the NRR preference is reflected in Enclosure "C".

2.

Note:

l (a) That the notice of final rulemaking in Enclosure "C" I

will be published in the Federal Register to be effective 30 days after publication.

l (b) That pursuant to 5 Sl.5(d) of Part 51 of the Comission's regulations neither an environmental impact statement nor a negative declaration need be prepared in connection with the amendment since the amendment is nonsubstantive and insignificant from the standpoint of environmental impact; (c) The reporting requirements in connection with the design analyses required by this rule are being submitted for 0MB review and approval under the Paper Work Reduction Act.

Since the rule is not expected to be made effective l

until the end of June, OMB approval is being requested i

no later than that time.

)

The Comissioners 7

(d) The Regulatory Flexibility Act does not apply to this-rule since the Act's requirements do not apply to a final rule where the proposed rule was published for coment befcre January 1,1981, as is the case here.

(e) That the Subcoanittee on Nuclear Regulation of the Senate Comittee on Environment and Public Works, the Subcommittee on Energy and the Environment of the House Comittee on Interior and Insular Affairs, the Sub-comittee on Energy Conservation and Power of the House Comittee on Energy and Commerce, and the Subcomittee on Environment, Energy and Natural Resources of the House Comittee on Government Operations will be informed; (f) That a public announcement will be issued.

Sunshine Acts:

Recommend affirmation at an open meeting.

6 L~

S Wil iam J. Dircks M ecutive Director for Operations

Enclosures:

"A" - Notice of Proposed Rulemaking "B" - Sumary of Public Coments on Proposed Amendments "C" - Notice of Final Rulemaking "D" - Value/ Impact Statement "E" - Memorandum from H. Shapar to R. Minogue dated April 15, 1981 Comissioners' coments or consent should be provided directly to the Office of the Secretary by c.o.b. Monday, May 4, 1981.

Comission Staff Office coments, if any, should be submitted to the Comissioners NLT April 27, 1981, with an information copy to the Office of the Secretary.

If the paper is of such a nature that it requires additional time for analytical review and coment, the Comissioners and the Secretariat should be apprised of when coments may be expected.

This paper is tentatively scheduled for affirmation at a open meeting during the week of May 11, 1981.

Please refer to the appropriate Weekly Comission Schedule, when published, for a specific date and time.

DISTRIBUTION Comissioners Comission Staff Offices Exec Dir for Operations ACRS Secretariat

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!!egulatory Cenmiden. Wm hinen, it:volve severely. med y degr:nled COMM13010l1 U.C.20 r..triv J.c'St 4./,N1 re Nr corn.Then A va been suppt.Er.ttu Anyt".Nev.tcreThe detrrmined la be a' au..h safety 10 Cril Part 50 recent accident ut Ta:ee Milc Vand, sign..ricance that th-y shoni.! bh cc:fifwd by regulation in order to provide Unit 2 (TMI-29. resulted in a severely, Domestic Licensing of Production and damaged or degraded reactor core with v.surance that the pu '!ic health and Utill:ntloa Facil:iic ; inter.im a conce'nitunt release of radioactive safety will be adequately protected.~ihe Mc:;t:!rcrnents Related to Hydrogen material to the primary coolant system specific dates to be required for Ctntro! cnd Cortain Ocgraded Core implementation by tiu proposed and a brp>M of fuct &ddi.~$cticn ammdments have bec;; revised ficm tla ConsMoratons metal. water (.dtcunium.cxygen) r AGENCY: Nuclear Regulatory in the core with hydro:;cn generation dates in t!'c previcudy mentioned Commission.

wellin excess of tha amounts required le:tca, ilowever, the i:n;: lamentation to be considered for design purposes by dates are being reconsidered during this Actiorc Proposed rule.

10 CFR 150.-11. Standards for ph:.se of the rulemak!n:; and specibc SUMPAARY:The accident at Three Mile combustible n is control system in li ht comments on this sub,4ct are welcomed.

d Island. Unit 2. resulted in a sevciely water cooled power reactors. The Tha following discussion provider. the d:ma;cd or degadad reactor core with accident revealed design and back;;round information, justificatica the concomitant release of rcdioactive operational hmitations that existed and cle:ification of the a:nendments, material to the primary coolant system relative to mitigating the consegrences tcl:: live 13 degraded coro occidcnts, tl.at End generation of hydrogen from fuel of :he accident and determining the are proposed to be alepted by the citdding. water reaction well in excess status of the fac!!Ity duri:g end Con mission in response to the TM;-2 of the amnunts required to be assumed following the accident.To ccrrect thia

,cc! dant. Additional specific gu! dance for design purposes by the current situa!!ou the Comm!nion has decidad ta for co:aplying with the proposed Commission reytlations. Furthermore, revise its re.;uhtions so as to amendments may be found in the tha accident revealed design and incorporate improvements derived faom previously mentioned letters to operationallimitations that existed studies of the TMI-2 accident.The licenrecs of operatity nuclear power relative to raitigating the consequences initial findings relative to the TMI-2 plants. The proposed amcndments of the accident and determining the accident havo been published in themselves are presented citer the status of the facility during and NUREG-0578. "TMI-2 Lessons Lccined discussion.The discussica !s organized following the nc:ident.The Nuclear Task Force Status Report and Short.'

in the same order as the ;'roposed Regulatory Ce nmission (NhC)is Term Recommendations," dated July amendments for ease in convenient therefore initicting a Inng-term 1979, and NUREG-05as. TMI-2 Lessens cross reference.

rulemaking to consider to what extent, if Learned Task Force Fina Report /* dated IIydrogen Mcnagement (i 50.44(c}(3) (i) any, nuclear power plants shculd be October 1979.*

and (ii))

de.:igned to deal effectively with The NRC's Office of Nuclear Reactor Section 50.44. Standards for degraded-core and cere.mcit accidents.

Regulation has also sent letters to all combastible gas control i,ystem in licht In the interim, the Commission is licensces of operating nucleo power water cnoled power reacters, requires a considering artending its agulations to plants, operating license c' plicents.

licensee or license cpplicant to show improve h> d:c,en manage rmnt in light.

licensees of plants unde.,:onstme: ion that, during the tLnc it: mediately water rccctor facilities and to provide and pending construction permit fellowing a postulated! css-of-coo! ant specific design and other requirements applicants, informing the:n of the accident (LOCA) but before effective to mitigate the consequences of followup actions that should be taken in operation of the combustible gas control sccidents resultmg in a degraded reactor light of the lessons learned from TMI-2.

system, either:(1) an uncontrolled Specifically for all operating nuclear hydrogen. oxygen recombination will not core.

CATES: Comment period expires power plants, letters were seat on take place in the containment, or (2) the November 3.19Co. Comments received September 13.1G79 concernin8 p! ant can withstand the consequences of eftrr November 3.1900 will be

" Followup Actions Resulting from the uncontrolled hydrogen-oxygen c:nsidered if it is practical to do so, but NRC Staff Reviews Regarding the Three recembination without loss of safety. If Mlle Island Unit 2 Accident." in which a neither of these conditions can be tssurance of corsideration cannot be set of recommendations was presented shown, the containment must be given except as to cci..rnents filed on or that was to be implemented. Regional provided with an incited atmosphere or baf:re November 3,1930.

aa oxygen deflCient condition in order to ADDRcSS: Written comments or September 24.10 9 to explain in more provide protection against hydrogen suggestions for consideration in d tail each of the recommudatius. On burning and explosions during this time.

connection with the proposed October 30.1979. letters were again sent Section 50.44 gives credit to Emindments should be submitted to the au licensus of opeating nulcar performance of the emergency core Secretary of the Commission. U.S power plants to provide addit,ional cooling system (ECCS) by specifying Nuctoar Regulatory Commission, '

clarification of the NRC staff that the amount of hydrogen assumed to Washington, D.C. 20555. Attention:

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be c ntributed by the metal. water Docketing and Service Branch. Copies of Lesses Learned Short Term reaction shall be eit!"r five times the c:mments received may be examined in 9

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t tal amount of hydrogen calculated in the Commission's Public Document The Commien has identified a demonstrating compliance with the c

Room at 1717 l{ Street NW number of recommendations that are ECCS acceptance criteria (i 50.40, Washington D.C.

specifically related to accidents that Acceptance Criteria for Emergency Cor.:

FOR FurtTHER INFORMATION CONTACT:

Cooling Systems for Light Water Morton R. Fleishman. Office of

'corire of these reports may be obtained from Nuclear Power Reactors) or an amount Stindards Development. U.S. Nuclear cro sa:es program. umor ot Techmc.I rcgated to a specific depth of fuel clad leformtion amt Docmat Centrol. US. &rt nr Rmlatory Commson. Washmgton. D C. ;:c533.

reacted (approximatcly onc percent of Enclosure "A"

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the fuel clad). w5!c5crer emoun:ic is smil. n dete*m!ne-! b ; m5ibiUatic *[rcridet seu /. " nn!ccf e 9 greete: As a residt,if a Nontre c-analyca. (:) there are ro r:- Eennt b:1!w:n ; : e t ! h/ -:d:r!y'N lictnse nrplicant can rhnw that the reunter-Mic.psfety di-i:. ::tiver:(b) drc,mpec:ca * "' S e more thcn calu.lar d raetal.w tcr re:ction !> we!!

the cost of ir.erting is emrih and (c) there edeuuatc!y c.w.tet fer by the w! thin the ECC3 accepan:c criteria has been *.ubstential satfnetcry attrmption us-d for the percert of feci (e.g ltss ih n c.S p.2:.r.nl of de fac!

cxperier.c

..i.h inert:ng.M. :k I clad that rmc2 I: r T.r?:rcs of clad reacts w!Jch wneld be well within containments. (2)!ce condenser providing a recenaMe bound to the tha LC nercent sper.!hd in the 1:CCO containn enu for ptc.st:7ed watcr c*ud:et, en erm ; !h.:t of 73 perant criteria). It is a str..i;;htte.rwrd task ta

.rcactar(i"Am n! ants not bc inerted at theeld be e e : :t C.e p:rcet te,bd demoura:e the.t C.c:i. is ac m.ed to this time. W;.iY it.ca te.; it e car. denser cLu 1 ! Z :S M Tic!'.73h s G;n ^3 in. rt t% contain:nent ofits plant. crcn if contain:nat:; would result in a creater Indic0!cd previ'# " this can cho the plant I.as a sma:1 co.:tcinment decrease in the calculsts.i res!daci ri:k, acc:. tat Io the 1.3 aga th:.! wv.*

Le volume weh as in a ht:: k I cr Afark !!

compared te UWR containn cuts there generated from other p,ssib!c r erces boiling water rea: tor IEWP.).

are sign!!icant r::asons f ar not incrting /suca as from ::;c GaN,:ss stce! Mant

!Imt op:ratin LWR plcn's with such as, (t) inertie3 would re:uit in an eac!:,en or rad.o:yt.,- cecornpc;.::en.

' hl. irk I containt.O; nts bwe been incrted increased prmo:m haard. sin::e thess Commistica wJJ require that io

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contaimnents must be entered due to gt.!danca provided in in ca:Iv

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frequently dering routine cperatices for V

" Control of Combust.ble.ueC:cS maintencnce pmpeces: (b) inertir.g cculd a:ovents c, hydmacn. and that thy Concenitations in Contair. ment lead to a d: crease in safe:y pefermance methods to be consid fred for nym ogen following a Loss-of-Coc:cnt Accident."

of the ice consancer becaure of centrcl ine!ude but not be !!mhed,to: (1) in which the designs were required to increa,ed ddfindly in maiatcn..nce: and Inertmg cf cont.cenent:(0) usa et accommolate five peicent metai-w6ter (c)inerting is not within pr.;ven hyd.elen recombu:er:: (3) use cf per;c reaction dering the LOCA b!awdown.

techno!or.y fur this type cf ontciament systerns:(4)u,e of a halen supp:cruzat This guide was in die:1 p*ior to and would h: ve to be tho ou.$ly system:(4 use of a fdtred. vent eyem; procu! cation of 5 :3.44. The ana!yred cad tected (wi*h I:kely chan;;es (0) use cf a hydre.zn combustien ccmb:n'ctica of contair cent volume and in destsn) before imposina such a

8) stem:(7) use of w3ter fog spray: and zirconium invente:y of the EWR hir.rk I requiremer.t. (0) 0:hcr PWR (8) combmaticne on ta:re. It is conta!nn. cats mede !! recessary to !rcrt contcinmen's not be inc:ted at tids time, wanticipated th:f. phcre app gricte.

these conta!nments to crcure that the While inerting would likt:y ;3ro !uce a owers grnps wil b: far:acc fer Wular hyriregen cencentration m containment small dec' case in residual risk, cny fatw, s to perf0:::inr e studms and nate dephection.

thus el.mld be no'rd that this cropored followinc. a LOCA w ould not exceed the decreate weu!d be substr.rtiaPa !ers itsho..

lower !bemabiliiy liuit of 4 volume than for AIark I cr II BWP. cents:n:r.cnts percent m air. Secticr. 50.44 and because d tiu substantia:! !a c;cr amendment will recqre that 11e ?.krk I 3

associated changes c Regulatory Guide containment salumes and the fact that can!amments for t..: ; armont YtrLee 1.7 at:cMug credit fer ECCS these contair.ments can prelably and Hatch 2 5WR p! nts be treated iny 3

perfer: nance patentia:ly reduced the withstand hi.;her pressures from manner si-!!ct to vner operating II.,n amocnt cf mctal wuer reaction that hydrog:n burns. rota of inue plant; mth 5fcpJ Manments. m2 was retu:r:d to be censidered in the considerations would permit first domestic D yvR h! ark n is schedu'ed containment desh;n. This permitted substantial:y greater vebmes of for fuelload.ng mlate-MO. and th hrst Hatch :. a ?.iark I ima plant located in hydrogen to be gancrated.

domestic LWR hturk H !s sched:17d for

! admg ebout Ntember IM.

Baxley. Ccct;;ia, and ope:ated by the The Commission is therefore Additional mformation and discussicn Ccorgia Power Co., to cparate without [considerin.; amending i SU4 to require en the inerting cf 1:gr.t;wster. reactor that:(1) all operating A! r!;CWRs he nmen s may be.cund in SECY-inertin;. !n addition. it would permit cn l

hlark 11 BWR p! ants now in the inerted:(2) hiark I chd II BWRs now W107. Pmposed Inicnm Hydtcgen operating license (OL) resiew precess to under cont'ructier.be inerted upon ntml Rcqm,t,cenu for Small l

operate without inerting. Dy a rtiling of operation: cnd (3) licensees now ntainments., and SEoY-CO-107 A and I

the Appeal Board (ALAB-029.

operating PWR plants cr ifark HI BWR

f.. Additional Inferraatica Re: Proposed I

September 18.1974. 3 AEC 425 (1974)).

plants, or license applicants that picn to aterim Mmp,c

}4'1'*If""8',a Centro.

inerting was not required fcr the operate thesa p! nts, study the sr.rioas and NUREG-0573 and Verment Yankee pSr.t. !ccated in methods of centro:!ing the besvior of w,

we refenned Vernon. Vermont, cnd operated by the large amcunts of hydtc en bef:fre the

• • h'e'Commistion rc::lizes that i

Vermont Yankee Nuc! ear Power containment is threatened, or of Corpcratien.

mitigatmg tne consequences '

amending i 50.44 to requi e inerting ' f o

The Commission beh, eves that accidents mvolving the generation of Afark I and II BWas does not settle the pending ecmpletion of a long-term large amounts of hydrogen. anese issue of how to trest degraded core rulemaking proceed;ng related to studies should take the adverse safety accidents in general and hydrogen consideration of degraded or rnelted aspects of the design features as well as raanagement m.particulcr.There are j

cores in safety regulation, which will their costs m. to account, recognizing that numernus reaulaticas which may be r

include a reevaluation of all positions on tradeoffs may be necessary. Ihey affected by the consideratien of the inerting of reactor contamments. it should coser a range that incl:: des both degraded core accidents.Some of these would be prudent. m the interim. to the realistic (bast estimate) ultimate arc Criteria 10 and 50 of A,cpendix A to establish certain inctring requirements strugth of t e montainment as weh as 10 CFR Part 50 which are related to for operating plants and those that will the conservative design basis. The containment design, and i 100.11 of 10 be operated in the near term as follows:

maximum duration for hydrogen 1

(1) 51ask I and II BWR containments be generation was specified to,be eigh.ded t

. Copies of those docu: rents are avat!4Mc from hours becauw.t was felt this p ovi e,. cm. mon p;,m om,, mnt ncm3 c 2n ii inerte4. While the decrease.m residual i

risk due to inciting these containments an acequate limit based on cred.i.!c te..et Nw w.*rven. o c.

l Enclosure "A" I

m..-

n-

-m..-~.a.,

...-< / Thurtdn, Ociahr.r 2.19m /. Preemd 7th ti hral Rej6*r / Vol. 45. No.1r f3..M

.w r.

n.n a n.~.m..

..-.w..~..

C FR Purt 10].6 ai.h 's re aicd t.: tha Tl..t n a spuriou; cr c: A.ntent prudent :s in *. : '1 mar.. : 'c. d. e v 3o oneefSa,,sciicebct:d contrn!!L ; do at b.#

tyM r

sM!rm of at:ch r re-cars. Aho Ne epn6 remnindct Lf ( elai. part!cda:i. in ca:.!aina:1t indtfet.civ.s w;rbi the etmv. hr.. I.Seme i mm.;.j a.t rcg.ird to the hydro;ca ;;cncratica result la the vent:ngi,f tb u:ntainment techar!c :y is wcli mia!.iis cu wd assumptions. is not being contdered for to the environraent.

tclatively iaa. pensive. th O u tis.lon revision at this th"e. Since s 50.4 ffd) is The pa: pose of th.% pwpused is considerty aairndia ; Q, mdations not being cha: s d. the hydreren et atrol amemiment is to gevu2 as;.rance th: t (new i..:.:IN;E:v)) to reqc.a that Ji system does not have to accommodate facilities with hoci:urs for external plants that currentiv rc!v. vent:nq:s any more hydrer: t than that n.rociated recombirers cr pcheciie.! pur;c the primnip

ns 'fe,r c'embat.th ps with either five time.s the mets!-wr.ter systems : hat are suscg:ib'e to sintt!c cor.*ro! be p,..;ded with the ccpability reaction calcuhataa to occur during a fuiiures, and that wousd remit in to instal: external hydrogen accombiners pos' dated lou-cf-codant accihnt er a operation of the unit bvnd its desi;n foMowina the start of an.s ci ' nt. This one percent metal. water reaction, capacity, have design mWdict tions will requite the use of ded:catar.

whichever is gr ater.The Cc:rmission is made to correct this situation. Systems cor.tainmet r,cn:' rations as mil as performing a systematic review of all of designed in meet thes*: preposed apcropriate shicidine, electried pen er its res3ulations frva lhe standpoint of requircments would not previde an'd operationhl prodedura?. ! S 2 ccasistency re!ative to the tatmcat cf throurh-linc Icakage pnths between th:

re;.:mliner crpibility wodd cnly be degraded core accidents. Fur:hcrmore, containment c! nosphere and the required to :..t!<Ty the cen}atiMe :!r.s the Comrnissinn hns initiated a longer enviro 4 ment and we.dd diminate the centrol retiairuments of i10.4 and thus term rulemaking cifort retative to possibility of violating the cantainmer.t wodd not le rcq : ired to have any mora

. consideration of degraded or melted integrity through a siecle active failure capability th. n that needed to ccatrol cmes in safety re?.:btion that wodd during hycrogen recombiaer or pur;;e hvd:c;;en rcruh2 rom at r est a 3 f

Identify, amons oth.:t thinp. whether

, syste n operatien.

t.

pcreent r;tvl water reac! Ion. 'Ihis and to what extent cesigners need to if IIy rogen Recornbiner Capaiality wedd re ride a long. term by&mn t

consider degraded c. ore ::cidents.

(i 5044ic)(3)(iv))

control capab!!!!y but not a ccpr.taliity to Dedicated Ilyhogen Control As discussed previously, existing quickly reduce the concent ction of Penstrations {s 50.44(c)(1'(iii))

regulations permit plants fcr which the large amount: of hydrogen such as was l

Parrtgraphs D.11(d). (e), (f), and (g) of notice of CP hearing was published prior pecerated duria;t the TMI-:'. a:cident.

10 CITt Part D require the incorporctica to November 5.1970 to use only purge The need for this capability will bc of hydrogen recombiners or pest-syste:as for combt'stible ps control addrested thuing tha long-tc~n accident nurge sys' ems for the c::. trol of subject to cert:in site tsdiation dose ruicmakmg on ccnsideratica ci cnmbustibie gas ennecntrati;cs tnside criteria. Depending on the containment degraded or me!ted cores in r2fety containment, depanding en the date of design for these piants, tne design bs. is regulation.

s the notice of hearir; en the auplication for the purge system wodd lend to a liigh Point Veias is the Resqtor Coo! ant for the constructica permit (CF) for the release cf the contain acat cimcsphere sN'" (5 r })

plant. All plant: for which the CP forlong term hydrogen contrc!in a hearing noti:e was published after period of time varying from a few weeks During the TM1-2 sccident, n November 5. M7J. must use cembusubla to sevent months.

substantial vcNne of hydrxen was gas centrol systems such as hvdrc;ca Recombiners designed to meet current generated in the primary syrtem. Helium recembincrs for the post-accid:nt Coma.!ssion requirements are not fill-gas and fission product nob!c gases

~

were also rdear.ed. In other ca: ors control of combustible. ires in the capab!e of preventing the hydrogen containment buildin;. phnts fer whic h combustion and the resultant under different circumstance:. nitro 2cn.

the notice of hearing was pt.blished containment pressure spike that which drives the passive injection water prior to November 5. M70, may use post-cccurred in tl.e TMI-2 containment system in PWRs. may also be a potta!!al i

accident purging cf the containment for building during the course of the source of gas in the primary cystem.

combustible gas control depending on accident. However, had there not been There is a concern that the accumulation whether or not they moet certain site the capsbility to use a recombiner at of pockets of noncondensib!e gases n radiation dose criteria.

TMI-2. it is possible that under a the primary sysic n of a reactor may

. The TMI-2 plant had provisions for d:!ferent accident scenario venting of interfere with the caturalcircdation l

post. accident insta!!ation and operation the highly radioactive containment pattern that is regarded as an important of an external hydrogen recombiner for atmosphere may have been necessary in safety feature ir. some accident combustible gas control However, the the weeks fellowing the start of the sequences. In ober sequences. poc,xets j

of noncondens bl i tases may interfers design of the extctnal recombiner accident.

hookup at TMI-use.the 36-inch As indicated in the discussion under with pump ope:atic.a. It has therefere containment penetra'tinns for the normal Hydrogen Management, the Commission been conchacd that under certain containment pur;c system by tapping 4-has initiated a long-term rulemaking circumnances it would be desirable to inch lines off the pur;n lines outside the effort relative to consideration of have r. ovisions for venting containment building between the degraded or melted corcs in safety noncon lensible gases from high points i

building and the outer containment regulation.Part of this rulemaking will in the primary system.The vents would isolation valves.Operationof the involve a thorough reevaluation of have to be operable from a remete

(

hydrogen recombiner required the hydrogen generation and control location, such as the control room.

opening of the inboard 36-inch including the design bases for current Furthermore, the introduction of i

containment isolation valve in both a hydrogen recombiners. However, bared hydrogen from the primary system into containment purge system inlet and

.m the TMI-2 experience. and the desire the containment building can create a outlet line. With this design. once the to reduce the likelihood of releasing flammabic atmosphere un! css hydro;en recombiner is put into significant amounts of radioactive precautions cre taken apinst this operation, containment integtity is material to the environment. the eventuality. Obviously, the decision to l

vdnerabic to a sin;k active failure.

Commission believes that it wadd be p! ace such a sent system into operation i

Enclosure "A" P00R inh d'1

s

.'l M;t : / Wl.4'2 No W / j '. s h'..

2. ' o' r ", ifT" / I'; *U;ed

• u C5-li.dc

.~

wou!d t i b: made with tb rrectest ab o che son ;;:ctacm. W c e,

ra %

.' cr at.at.

L.an a carr tr.1 r m eiat.

Allan may te ho buic Ju....

te a Lad.n.;. n 5 i.

7 f..ccodate ir The pacycce of this preposed ireprovementa in tim faciliti.es,.

curwv >;;ve, it is p 1 U!^ irr su61cir m arrenJmcal is: (1) to pros ide n.ar. tor componcrm or sy:.tems.

noblu r.w to be adcad.d : nd not coolant syWm and ter. car sesstl head Areas that ray be umi duri; : and re! rend n0:n the chaw!i that the high poir.t se:.ts remotely operable from fe !uning an acciaent should he reacWcs o cioiosiina ocanal nation vay the cential room in reacters where these

p. aided with nNrepriate shiridi~; end be urdi:ly consernt;.a (hb b). Becausc are not alwiy in plne: nnd (2) to uthor radiation p:c::ctico m>

• . w so the ::.dicia. fine cencertra' en was provida ossemnce that the u a of these that perronnel re' T~mi q nece : ry greatly c arer.timax: e n:! ' plet vents will not l'urther ae:ravata the safety functions wdl net receive recte personnel were required to perterm their challer se to containmeat ce the course th m a 3 rc= w:.ck body due ur :s opeia:icua ;;5e v. :: c ' rer riz atery of the emergancy.This prorosed equivalent to any part of the budv.

protectiva equipmen: when this use was requirenrnt is also interpectSI to menn When determiniq an'icipated not ne? vary. Artect one::'!Nions that other sp!cma whi:.h mer be personnel da:ez ier the purpose of were prob:.bly beluv. '< vela requainn required to mciutain adeque.ta core meeting this. fat.3ity dt.e:n uitan, protecin c rctions. Oac acceptab!c cooling rnch as the isolation cor. denser une must be tid.cn to determine the method to minimi::e th.: c:. '/e n is in in D W R: anJ f nr decay heat nmoval necessary cccpancy tina in 2 ewcific me: sun the radicied ne by < amma system, have th= capability to be area. For exe ipb. areas in wi.:ch t :c:c energy spcc: rum ana!pis.1.quipment for remotely sented if the accumulation of will be continum occupricy are there measurementa is commercially noncondar.sibt'e gases won!d cause their required to have much lower ddgn avai!ah!e.

loss of function. It is reccen.ced that it is dc:a rates than :*re:c where mih!

r.ficeive menitorin; of radiciodia:

Moracticabla to vent cach cf the many occup ncy is c>.pceted. Therefore, the level; in ILc imildin :..::.dct a :cident duusende of tubes in a U-tube steam facility design should be based vrou condit;:ms. cnu!d inch:de the use of generator. Procedures shculd be expected cccupancy, as well as fin portabia 4.nolyzera cud imtruments.

developed which ensure that sufficient radioactivity present and the s'.h.! ding Altemettseiy the cqiility should be r

liquid or staam can enter the U-tube available.

provided to r, move the inter sar.g.lc to a region so that decsy heat can he The pu pose cf the proposed dcai;;ra low b.djm. tad. low conta nination effectively renoved from the reactor requirement (1 SWaib)(1)) is to crea, fer anclysis, to Wevent coolant sysum. Proposed f; SaA4a(s),

factiltate operations during end measutoment inaccuracics due to f diowirg an mcir':nt in areas aGec:cd bac!:;rm nd radist cn.

" Training te Mic;ata Depd d Core Accidents."inc!udes plans and training

,by systems thst :nay contain abummally The ptapose of tne proposed for the use of the high point sents in

. high levels of r di6tivity and to requiremcat is 50A4

sid!2))is to cmcrgerev situaticns.

enture that saf 'y equipment in improve the accuracy cf r.casurement of Protectinn af Sciety Eculpmcat c.nd proximity to the recalting radiation airborne radiciodina con =n: rations Arcas Which :4av De Und D.ni: e cnd ficids is not und :ly degraded. I' s'.ould witliin nur.! car power piants.

Fol!owir t ud Aicideot (!i 50.44at )D))

be noted that tha radiccctive rm.tcrial Samp'; m Dering and F:J!owing en h

During nad fo!!o.vlag an accident that release specifiad in S 50A4a(b)(1. Mil

.gccir:t (s 33aga(c))

results in a severlv degraded cot e, a represcnts the Com:mssion*s positmn at Tir.ulv it. formation fam reactor la go emotmt of radioactive material tnis time; homm. this position wd! b*

coolant hnd ccntainment air samples may be relased fran. the fuel as a result reevaluated during the, tong-term can be imrortant to rer.ctor encrators of cladding failure and be carried micmaning ca consider & tion of for their cher3 ment of syste:h.

throughout the facility by the waterin degraded or melted cores in safety

/ conditiem. can influence their which it is dispersed, and by the regulation.

subscouent ac:icns to maintain the circulating air. Systems that were not Io. Plant Iodino Instrumentation facility in a safe conditien. During and designed to contain large amounts of (i 50A4a(b)(2))

following an accident, significant radioactive material, such as the 10 CFR Part :'O provides criteria for amounts of fission products rnay be auxiliary building at Thil-2. ruay control of exposures cfindividuals to present in the reactor coolant and become highly contaminated. The radiation in restricted areas, including containment air, creating hi;h radiation resulting radiation fields may make it airborne radiciodine. Since iodine levels dtroughcut the famility.These high difficult to effectisely perform accident concentates in the thyroid gland, radiation !*vc!s may dday the obtaining of information from samples. The high recovery operations or may impair the / airborne concentrations must be knownin order to evnMte the po functionir; of systems important to l

safety.Thue systems, although not the thyroid. If the air';orne radiciodine radiation, and high levels of airborno l

specifically identified to perform post-concentratien is overestimated, plant contamination may render in-plant accident functions, may nevertheless be personnel may be needlessly required to radiological spectrum analysis of significant value after an accident. In perform operations while wearing equipment inoperable during and after addition. areas which may be used respiratory protective equipment. Such an accident.

during and following an accident. such action can sharply limit communication At Th!I-2, all of the above preb! cms as control rooms. radwaste panels.

capability and may diminish personnel were encountered.There was no emergency power supplies, and performance during en accident.

c;aability to obtain and analyze in a instrument areas may f all within the The concentration oi radiciodine in tia ely manner the reactor coolant and radiation fic!ds of these systems, air is determined by use of an iodine containment air samples under the Procedures for the use of these areas adsorbing cartridge through which air existing accident conditions. The during and following an ac;ident may has been pumped. The cartridge is acquisition of reactor coolant and be all that is necessary. In other removed from the air pump and allowed containment air samples was delayed instances. permanent or temporary to ventilate to permit any adscrbed for several days while personnel shieldmg may be w!uab:e. iutate nobla pscs to t ae tu tnc atmosraere.

radiation protecticn pu cauti,m were instrument and ccatrol capabil;ty may The cartrid;c is then analyzed fcr taken. Once the samples wete obtained, i,

s.,.--. ~.'~.,.

/ Ve!..M. Nn. l'M / *.Tb".Mt h",. Oc te.5-p Pmi / Pr: r ~; Sb b :. Reekte r r,3 typ 1*r r

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rhem w.=r- @ ?: nt dela in th:ce:hi t:s s 'n 6 a i :. r* c' !b

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me.t.

-rform;ng tLe nX tealsr t.am r aM M u t s e.nu '

'n culd Add:tiern

. m m' er

.m:fyns of the w. des.Tht nii-2 have h5d red its e,:.. t h si Sctn in t! e U

' mt dra.mt pw.J.,!y

.pectrum an:.'ph ecuipment w 3 nead. d. The lessoa !c +>n:

in tha raw cor.tribet::i t, e.. 'wr.t :.r. re : n to insgerable le ce of h:gh b. ci. ;tered was :mt n me positiv ccrtml and i!cedim! ":.9 v.M; deady ruhEsh a radiatien an:i : s sw pb h:"i h ha knowkCm ef tb le ' u. w of ther :

nt-2.1 for m: Jr e anta!:. t s. : et packaged and !Nwn to an of site systu: a n needed ta r,ravide the levet inic:4t.en in !$c contri room. viith labomory far an6 sis.

operctd.3 :daff wi;h tha :na.Umn:

Ins:rumtr.t ranm v.hich W de la summary, r.A.ation levcb at n!!-2 usab:e equipmant and to resuscs or raad:ntm :.cet..m :Wedmg ineis.

dc!n cd acquistian i,finfon.rc.mn to con:rul the ic:cr.sa c!...:,r.ct.se The 1;rd u..: af.afus.a,...c,a..Mch conf:rn th:2t si~df: cant core dama:ta metenab to the c: rirmert.

was st.bs. m.at:. c"nside. tu mi had occurred. Nnot ac ;uimn c.nd The purpcsr.s I d.5 p:. ;.:xd crit:cali.a;,o.taa'e ia de: tr *..:.g analysis of reactor coolant samp!cs amen:1 cent are to r. cy!ra thr.t every centdnment catCden: ct n*! 't was a within a few hx s ener the initia!

reasonabh: c~ ort b3 r a.!a ;., cuminate measuremett. d: hdrewn scram would have ind cated that or reduce 1 e leaksfe hem !!:ne.

concentrat;on in !!.c chnta!=1.cn!

signiEcant core d me;e had occurrcJ.

systems, to require th:.1 p ri,ce: tests atmoschetc. Hv:imen gas wn Wid such !,f-n.uon. ear!ict remedial are perf.,rmcd to ensurc 93t ie leake'ta produced ar a kw!t of the andan cf a:ticas could Fava !.. a taken.

fro;n these systems is M -ind to the

>.irecnium :u.r.l frem the fud t:ad Hra Simibrly. cnalysia of cn car!y maximum extent practicch:c. tad ta and primary coc: nt water in Qe reccior et,ataimaent air::mi e would have provida the plant staff wim cu rent core. The gas was vented fro.n the d

indicated the pres:nce of hydre; ten. and knowledge cf the system !:aa;e rates.

reactor cocSnt system to the of radioisciopes i OcMing sisuificant Accident : fonitoring lastru:r.antation contammeza r.tauapnere. !!:0 fiee core dam::;e. :nd the potential for a (3 m.e (:))

hyic;en in co.tainment rerahed in a hyd tq;en exp!csba in de conta:n nent The ccc: dent at n!!-2 demonstrated rapid bu:n I:nd pre:sure spike avant in The samphn;; fac!!. ties are required to that contcin=cnt conditions can crise the containment. Samples of meet the destp criteria of i MAic) cs that cre nore severe than Qose that containment et. qhere were tJen well as those cf s nuaibM). All were postulated to cccur dunng design followin3 the uccident at TM-2. but the sources at racuton shcdd be basis accidents.

process imoived n:k to wedrn cnd did considered in cu:ua ing de derigns.

Appro:anately ten hours citer the not yield real.t=c ir.fo metion. Tre such as the sarr;de Uncs. the 3:yte,!es start of the accident at T.MI-2. a OS psig e;ents cle r!r :. haw that it is evrtici themselves. cad wha c.c:aaet:ve ::n:s pressure sske cccurretiin me that th operdct have cent:.mo 6 near the samp:=g stat:cas and analysis containcu nt bui' ding. It is 1-licted 6at information abcut the cont:rcent fac n,nes.

the pressuie soike was due to tue rapid atmosphere hydic;cn concenraiicn for I.eakage latgity Outs:de Containment burning cihydrogen gas in the indi:atien of th :#.ed and use of reactor (1 soma [dj) containment atmosphere. it is 1 nown pressure vessai wn.;ng or centa;nment Sev:ral of thc cag!n cred :afety that the presscre spiie reprt:ented a combust;bic gas cer. trol systems.

features and auxiii.:ry systems. beated scriaus ccaditian widda centainmcnt Indicatien of ccntainment prenure, outside reactcc cenunment, wdl or ; cay and that the pressure in.ticatien its elf c0ntain=cnt hter :evel. &nd hydrugen have to function d::rina a serious could have been. but was not den ccnctn: ration ir. the conteinment trandent or arcicer.! dth Ir.rge accepted as. cnticalinformation to the at'ncsphe will p:ov!de c-iti 31 radioactive invenkrict in the process plant operators.The events at Bil-2 inbrmation to t!.e operator en fluids.The leakepe frem these sys: ems, clearly reai!!rm the necd fer containment cenituns during and when operated. must be minirni:cd or containment pressure indicsticn in the fcilowing an sccident and shoch: he eliminated to pres ent the release of centrol room. The instrumintation range qu !!!!ed to the criteria for si;;nificant an, ants of radioactive should extend from a lower limit of five instrumentatian impcrtant to sefety.

materials to the environment. Examples psi below normal operational pressure These paran'eters should be of enginected safety features include to an upper limit of three times the centinuously provided and re:crded in residual heat reemval. containment design pressure for concrete the control r:cm cf all nuclear reactor

. spray recirculatien, and high pressure containments and four times the design power plants.

injection recirculaien. Examp:es of pressure fer steel containments.

The radiation les,elinside auxiliary systems include sampling.

The sequence of events dunn; the containunent is closely related to the makcup and letdawn. and waste gas.

accident at n!!-2 indicata another item potential release of radioactive These systems cre checied during pre, of information which could have been, materials in the plant effluents. At Ut!-

operational testing and startup testmg but was not immediately accepted as.

2. the radiction menitor in containment

• but are not usually in:!uded in any critical information in the diagnosis of had a range li: nit of 10* R/hr. which was periodic leak testing program. It is the accident namely the free liquid adequate to meet the conditions of the inventor in the containment bmlding.

accident. In reviewing the monitoring important that the plant operating staff f

know the leakage rates of these systems During the accident. re:ctor primary capabilities of other plants. however, it and minimize this leakage to the coolant water vented throu;h the drain was found that there are few operating maximum extent practicab!c.

tank re!!ef valve and dramed to the plants %vith instrumentation capable of Some of these systems were used containment sump. Prior to containment measuring levels in excess of 10i R/hr.

dunng the nil-2 accident with resultant isolation, water in the containment During the initial post. accident period at releases of radioactive materials to the sump was discharged to the auxilicry nt!-2. questions arose as to the validity auxiliary building ventiliatian systems.

building sump tank. Eccause of the instrument readout and to the These releases are believed to have containment sump pump cperation had operational characteristics of the resulted from leaking re!ief valves, routinely occurred several times a day instrument under the accident waste cas compressor seals. valves.

before the recidant. the traa-%r ~ocms e :vironment. The Cemm!srian censiders manifold lines, and open rupture discs.

was not reccgni:cd as a potential scurce that the hi;h.icvel mon:torm;

t Tuk-! Rc:.b:-r / W. E '!... tT. / ':Wday. Odabr :'.13 / hrn u d W.tes

.- - 1 i

1

-..._..-.--m.

Inmmnentas,a E t!z c.ntaine.ca;.a i.n!W!aue by the n=Gcr 2n., dest

'IheCor de, x1. :!Wr

• tt t

~1Mi-2 w :.A,nc '9..:ec:::: il.2 system.

trahndogn,

- s xi A D t.: dst!ag I:.diatica h veN hovrevct. it A si.d.c cceda'cn o. ed in ne 2nenita k.: m.w. nve is Nhuh'es dso considers thet d.c nm;c limit t.f : is section of tb plant v.-a crPer an 1 rcdiaMn a in.:cid: a :-' ei instrumentation thould be incv.. d ta designed to detect and m4.2: re the release.. at u Mted : nan s

reater than to' X/;.r and con >g cf at prescr.ce of particuk

e raucactive radicioCn cmcc.intione ca : u crder

[ ;least two chanm.ls, each separated material m plant gascous eIncents. In of10:pCi/c. ftins t':erafer: conc! ded ph'ysica!!y from th: o'hcr. and tht be this ca:2. the presence of nc;d: nases in ths.t umphr. t c.l p':nt pcm 2%nts, instrumentation *yste n should be the gas stre:m passine t; r.wgh the ti,th laboramry ana!pis Of sa:nySs qualified to the criteria for

- monitor's particulate t.itu was sufficient suusequeat to reiMe. is ii.e or..y instruricatation important to safcty, to cause the part:cukte vetir,n of the c urent!y vdd ic6nique fer t.v.k-in<r Guidance on the ranges and monitor to read off sca! and ac.cident !cvc! rei:sses of radscina specif!:atiens cf the accident rnr nMorie,',

er-oneously indicate that large and part:,cu!.*s.The proposed instruments discussrd ab:ve can be quantities of particulMc: were being amendme.:t eccirecuire that this found in proposed Revisien 2 cf releas :d fram the p!sr.t vent.

technipe be tqd udi! on-!!ne Re2ulatory Guide LS7 " Instrument : tion The g..-ablem is generic. A recent mo :itartyg...n.:ty for ace:hnt !cre!

for IJghtMs::r.Cc: led Nue! car Power survey cf existing radion.tive gaseous tauc% dine c.ptopad. a.xisting relsnes is de

.a.cnt =cnita.;ng,

Plants to Assess Plant and Envire.s efiluent monitorinst capGlities of Condition

• Ocring and Followini! an operating plants shcus inat less than 20 equipment shm:!d ha used to c.onitor Accident '" which is presently out of percent of cpeatsg pi,nts have

[admiodine re! case-i withm moniters that would hau sned on instrumentailcn dp!~n hm;tatiens, whh public comment.

scale u-der the c::nditicas ci the nII-2 samp!ing nr.d an:.ysis betr g usca to Radioactive gasaous effluent monitors accident. A so,it can b.e i.howa that the cited men;turite,capabilitics to designed to cpesete uncer condit.cas of accident !creb.

normc! cperation and ant:cipated potentid releases frcs pnttuiated, accidents rnay be severa, crders c.

Dc!cction of Inadeacate Core Cooling operational occurrcnces do not have magnitude higher thsn was encountered

($ 0 ;.la(fi) safficient range to function under at n!!-2. tander sucn circumstances.

condition o'lo.e stater level 1,'the During the TF.S:t accident, t c release conditions associated Wth certain types of sccidents. Gene-al n ne of 6e emuent modps new in, sWee at any cpuatmg punt woulu reactor vesseli.nd inadequate cere Des';n Criterion 61 of Appendix A to 10

1. '**I" '* b' coolin mas m acm:ir A for a I:1g CFR P rt :G requir:s that ef.'!uent A n@emc t fe emue,t mem. tors to per!od of time.*36 prob:erc.was the discharge paths be monitored for s

radioact.Gty that may be released from

" ave an epwatmg unge s,cient to result of a com'ainaison or factors pemh on sccle reading: undct acen!cr.t inc!tdMg an imerficient range of portukted accidents. Th::;ascoua conditient is needcd to prosIdc existing insirvmer:tr' Ion inacequate cificent rz.onitoring system for n!' was meamnpul n! ease infcmat,on for emergency prue dures,inadequat,s t

evaluated during t:.e licen:ing rerkw fishe emergency acticnshamples of operator trai n:w. unfavorable and wa: fcund to be adequate for anticipcted release pcints that showd be inst u nentlecni n(t:stteml c:kula:cd relececa Trem prcvict:dy monitored are: the auxihary out.d=g.

Infomation), cnd perhaps insuf!!cient postulated acddents: he.vever, the n!!- radwaste budding, waMe ps dreay instrumentatbn.Theinstrumentat;on

• 2 cxperience E ves rise to new tanks. main condenser rar c;,ector, uWR c! ready cvaAhh on some cperatine i

postukted design basis accidents and main condenser vacuum pump exhaust.

reacters that couhlindicate inadequate their associated releases; PWR steam safety valves and core cooling includes core exit At ut!-0. the ncbie gas section cf the atmosphere steam dump va!ves, and resistence Ecc, era!d ieg ana het lea thermocou*ks ca gaseous radioactive ed.luent monitor BWR turbine building. Oiher examples ture detectets w uld be areas thlat co nmynicate

[RTDs).in-corEneutron detectcrs ex-ea a e o nt ton up p /

"'N **

'I"'** #"'" ** Y c re neutron <.ctccters. and reac'or cc (Xe-133). During the initial phases of c ntain pnmary coolant a r.cntam. ment coolant pump curnnt meters. Genera!!y.

the cecident, radioactive nobh: gas gases, such as letdown and cr crgency such systems ve:re iacluded in the effluent monitor readings wera off scale, core c hng systems, and external reactor design to perform functions with actual release concentratiens other than monitcring of core ccoling or hydrogen recombiners;tted by plant calculated to be en the order of;0.,

Dased on data submi Indication cf ves:el water level. Tne as yCi/cc to 1 pCi/cc.

operators. the installed capab+!ity exists designed and fidd-tnodified A section of the TMI-2 plant vent for mcnitoring noble gas releases up to a instru=cntation et n!I-2 did not gascous radioactive effluent monitor concentratiott of approximately IX10*

provide sufficient information to designed to detect anu measure Ci/cc which is a factor cf 105 higher indicate reduced reccter vessel coolant radiciodine releases, while remam..mg on than the maximum range of the level. core voiding and detericrated scale, gave an e roneous indication of instrumentation in use at T*.II-2.

core thermal co:tditions to the reacter high r:diciodine content in releases It has been es!culated that the operators at the time of the cecident, from the vent during the initial phases of maximum concentration of undiluted although,in retrospect, this could be the accident.The Indication was cauwd noble gases which could be present in inferred frorn the instrument records. An by adsorption of short.hved noble gases effluent paseous releases is on the order edditional diagnostic tool that was not in the charcoal cartridge, with the of105pCi/cc.The Commission available was a continuous indication of presence of the noble gases being read considers the upper detection 'i: nit of 105 the margin to saturated conditions in the and erroneously Interpreted as pCi/cc for noble gases to be technically primary coolant. A meter to provide this achievable. Guidance on the ranges and information is prcposed to be required

,Cches of this drafi r*gulatory aukle fray tw specifications for the required for all PWRs-obtamed fre't: the Diuses c.f Tectnu! inter-ition instrumen;s con be fuenl:n p:cp > sed The purpos e of this proposed and Dnmer. Carurul U S. rMar Er:C.tsry cerw.m. wa L:pon. o c ms.

Revision 2 of Regulatory Guipa 1.97.

amendrnent is tu provide the reacter P00RBR M

s j - h'er / Vnl t3 A*n. ' 9, / '*%rr F - Ctn%r 79/ Fra- - - r' " '.m f

.",se-em

[g-.3 s we.-a.-

I

.- c.

wa., e i

.2

-wow

....-..,.c

.-w.

.+

cre,,

e

~s.

e !,, q.... t g,. g.. p,.n.,. 7,.. q

,3 ep e.?. wf 6 i-e+ v--

ty !n t a.nend.., 0;; t i t c (fy, ca.t : 1...

p oc@.r:*s. v. t.

. r::::.s ry h contemp.dc).

th,t;.5:-

tcudik rter.;:U: ?: -i

^:acnt ac:..sr.s iA} :r? d dir a tr 5:. t scn; v r.p t --

to corr ct or avuid :- M ?:r; cf PAM L3-0 F.Sl'O LP " " l C R e.

.m 13 tb,. g.

.,13 g; y, in PflOE F!! /-W UT!LL S W 5 ; a:::' n r f A ~ - A : A.ee in ar.d iandequate cere cc..;. :: :

Tr.did:r; to MitivM lL;;,rs.ded Cme "U 'Ji 36 L3

[ J :du hel;;J. d

1. Section 50.4 I cf 10 CFI'. part 50 is are <.ized to r,at:s:y the flow As cihnt;(150.4;vg;)

amended by rer;dng para :ru'-5 (c) to tur.kmenta oDna c..9:n.nl The T.,.,.,

, cccv...a :u..:a; poi. tco. out read as fauw.v::

rec,~ient>iners or ee r ru, terns. or b

s re 06. c.-

r c.

c,;gta...

,a,,

the 2. red to ir:r:ove 4 :: c O., y ts m.-

rccc ;;t.ze. dine'now. r u?s c., mnd l GnA4 Sts.%n ter cemt.ustit.!: gss t y,y.y7 e ge y ;v,,,3g7e73 nr ;,a7 mitote the conre&nces ta c.cca.cnt.;

. contret, ste in 1:et wata: c: y.s: power g,.;geg3 33 g g ;,_,g tg = 3,cr af,,, g3 ru ctors.

g g gg g g g

resulting in a degr:.ded raacter core.

That it. the cper:.t. ; per.sonnel s!.u:!d App,..g;x A, n n.a.,j ggge, 33 (c)(1),er each bolh,ng er pressurized ccaurned sfa 'a Wit.r$ bcih fcr have ine capability to res;:ond to events r

that are morc severr :han #hoso

!!ght-water nuc! car pewer rea.: ot fuelad cogainmp3 g, pagan p,,mem said for trondents and ac ura's con !dered as with oxice pq!ets witr. a cyt.vfacat opuatbn w cc e:1r';:a! aco:r.b; tw. or the design basis faz &: p!mt.

zircaloy cbda:r:. it shall in senwn that nur e s; stem: nmi cre 3::cd t2 :, dsry The purpose of this propo;ed cuiing thn t:r e period follo.w:/.

h.c' tide requircmeata cf the external amendment is to require aditional postulated L0uA but pt:or to 4cetwe 7;;;. hines e pu bd. facilitics that ovstems.

training. for all opm atin1 personnel.

operatton of the ccmbustib'c ps control ggy);;v }cnut.ry 1 specifinlly in rc:stM:1 to ;cc! dents sy: tem c:Ser: !!).t.: un:ct tnylPd 7c;y m,.57 pur e ush. cts as tne p:i.e.e involving a degt:d:d core. This traiains hydro,pn, oxy,en recombeatma '.reuld

.,w,ifrew d r ?.deth nds wdl prerare tha cperet;:.g ;;ersonnel to nat taxe pace m the containment er (n) fo:!oicing a LOCA saril be proud use all the ava!!abic instramentation the plant could withstand :$ e with the capability to insta!! external t

end equip nent to pecper!/ respond to consequences of unc:ntrc:bd hydrogen

• reco.biners io:!aieir? the start of r.n cxys;cn re:cmbir.ation withert mss of eccident that meat t'5 combustis!c aus such accidents.

safety function.

ContrclrequircraenIs Of this ecction."The in addition tG the pmposed interim rulemd!nts, the Com.aesic.n has (2)If neitbar of these conoj..uens can containmer.t pe".Stratiens aat crc veed initiated a long terra ruler din <;in p'e su wn, te ontainqnt m.,be cut m:et the c.!:e:'a in paren.wt.

r 'N'd "IIh an inrrteu ctmovere er d'V- % 'y N da what extent its rep ia'ticus should be an cxygen dcuc. tent condit;ca in ordar to g g'n".. Wh whicn it is ceasidain ? v.ha:aer end to

. applicabia to eternal sortie pr vge protec.a

amst hyd c7en.

recombiners, amended to derd r.hth e y with :

burning and enplosions dering this time t

br.>ad ran,;e of a:&n's invoh. in<: a. dpmod.

2. A new i 50.Ga is added to 10 Crit reacter vj.tou cc :l

r. leva cagrace (3) However:

Pari 50 to read as fo!!ows:

j or n.diee...s part c. ac Icag. term t3u Jm.T,19. en m. cab!c b. ut r.ot kter

-Dese v::remants to n:'Mte

() As soc, as practi rulemckfr.3. en a vance m.iice of n*ca g m. 3

+

pro ~7ored ra'emcking in s. Lecn isseed pmMp ead thecmeouenecsof accuentsresA4n 8 5 bo;h. n3 light water reactor fac:lity fer a dead.a core ~

soliciting comn rats on certain s~reeffic t

que:tions. Any views end cesunents 3 Ch the appIt'cau n f r a : nstruction (a) gj 3 pojof g,pe. By January 1.

j s

submitted in tusson<< to tiis notice of t a docke'. d tween March Ia.

19L4 each boding sad pressu-ize.1,Ucat.

proposed interim rulemab;t wdl also he IF'-

water nuclear power reactor snan ou be considered, alon' v.ith comments (i D=$# #="! '=' 'h !I h provided with reactor co,alant system I

c submitted in resPOn'sa to the advance performed (A) for each bo...

r wh:n 4t-ir a li anc: reactor vessel hea, uh pumt notice of proposed reiemaking'in the wa mea cn the vents ' remotely operated from the overall long-term rulemakin;t related to pphcation far n construction permit control room to provide impro.ved.

consideratien of d~ raded c'r melted was docketed after July 1 197' and. (B) operational cepabit:ty for mainta.m:ng lation. Pendmg g

cores in rafety rm"long tcrm rolemakina

- a h pmsaurize. Eghwater reactar adequate core cool.ng following an a

completion of the fa tuty, to evakate measum that can accident. Since thre vents form a part e

it is the Commis:ica's jud-ment that implementation of th.: req [urements be taken to mingp,,the conclumees of of the reactor coolant pressure l*#E" *"

'"I*

f nyor gen g<merated boundary. the design cf the vents and 1 creby bein3 E cFosed wi!!. in the within a hours a.ter the start of an esscciated centrols, instn:ments and e

interim. p. ovide sufficient assuranco d

i=E ' ^ b pow cr sources must conform to the f

(h d S I

that the health and safat} of the public

• t '" up tg about 75 preent of the requirements of Appcndix A and will be adequatel Erotected relative to fuel claddma with water). These desicn Ap;cndix B. In particuIar. these vents Y

degraded core accidents. Other

""i"

'.'.prcp sed des.u;n (or shcIl be designed in such a way that no documents which support this position 8

8' C UNN" sin;:!c fciture could result in either (1) a hyd g n?m containment chaH be loss cf the capabdity cf the vents to are SECY-80-107.107A.107B. NUPIC-0578 and NUREC-0585 which were c mpleted and submitted to the...

perform their safety functions or (2) referenced presiou21.

C6mmission by (S months from cuective i^ d -

Y

" b *DM'

• M' (Separate views of Commissioners date of rule) or the date of r ocketin of 8

Cilinsky and Bradford are attached-)

the app 1ication for the operating hccnse.

andi1ro s iYhich.'. icy bc Used Durie"y Accordm, gly, notice is bereby given whichever is later-and Fodi owm.qun e-ident. By lamnar that. pursuant to the Atom.ic Energy Act (iii) By June 33.1081. faciuties that

' 1.1931 each bonmg and pressuri:cd n

of 1954, as amendcd, tg,e Ener2y rely upon external recombincts or purge light. water nuclear power reactor shall Reorganization Act of 1974, as amended. systems to satisty the requ:rements of h

and section 553 of !Me 5.nf th7 United

( 59 U -ha'! H -roti.A:i wit"

,b < rt that tic t4 s.a (*-ube stt eu gerarnors glates goge agoph.un ut t c gudowing cont.dn:nent p,:nutran" ens for the do na eqwte senung.

P00R D E M

4 F~NI lbr;ist.ar / Vol. 45. No. 93 / Thus... Ato* cr :! a 4 ( Promc

. L. s t 'm;

....~. - n...~ ~.,

.....~ - - -.a

~,. -

.-.c

-~

be p.uvHW erth I e.<4pte acces ter~: 6 wr+uiin. cwni ti. %) ef P:):te e.:

nO'^ 'as to aro: "hkn triy be used durina sci this secti..i cerw.tratiw s :. Me e wente..e c

followinq ari cccidert crd protectica ei (2) Te cap:Mty 3 r? - a s m irla 0;*cir:1.

.4:..- J safety in iipment s th.et ca accident, must inc.ude te cap bine tv dein: so.

(4) Oauntifyi:s..h : cencen*;a2 : c!

whdi rc'st.its in the rdem ol!arn pre rtly. c;* t. 'hw inc + a rrb:odinin enJ e ' ietwo amounts of rad;oactne rmiterial. will nel radiatica enesure to any M..daal in particulates in tr.a niic.cne cifh:cnta et li rJt pmonnel ocenpancy or de.rh exce:s of a tem to the w@ tx:, or its c9. anticipved r Ante peint by sdcty u uipment. by ca rrdiation ia.:Q equivc! eat tu cny pr.:t of th n C.t.but 1.1: n.

s.

that 1.:sy exist durin enJ M!owin;t t'::

(3)Th"'0D l'A:tv to W,t'vely (O All!! 3 irr! ~ *ys An3 rm#!c '. t r

~

necidrnt. to the extS! th.! rer;uired analyze a sarnpfe rmst ta I& 8 en the cystr'rs used far cecie::t mon t. ?ne safetv functions cancut be use of either ia-line mcniterir y an ei :2 b: d::i:~. s ar. ! g:r.19.d iv.i:h acconsp?ished.

onsite radialogical and chemical extended rare.csj to per'orm their (i)The facihtf dcshn must be haced annlysis facilitv. end mut ty evije, as fu. action fe!!ntin,v :ce' dent on a release of radioactive material frort needed. quantification of ta: fe!!: wing-charagteri::c,m ine. ndaa:*iva the fuel to the prirun coolant system (i) These r.idpt:otopes crcas<ary to matenal re! ease tama des:nbuc., m, that b oct less thnn ;ia: ; ef the core determir.a tSe u7 pee or cc:: d,n:a.;c:

pcreg cph (b)(1F) f HJa sect:en.

eqt.ilibri.an :ob!c pn in.eniory. 50~4 of (ii) egJr per in 'he ccz.Wmcnt (fI Celectio.1 of edq rate Cera -

atempme:

Coo.,ng the core quilib::ur. Ec!can inv.atory, (1) Ecc : bod, ing an,, pressur:,:c lisht.

5 and 1., o, t.:: rem 2im. cit c'cre fissien (iii) Total dissolved ga.ses anr2 dissolved hydrogen gas in the reactor water nuc! car p:we: reactor i!cecsc e pro. ducts. For. equipment a.nd area.s. ba chall davolop n-d 'rm!?.mant pr.m. drer coolant:

aftacted.uy t.:a rac.x cuctan:. it saed

- 4 m t.i,e reactor con,.at.t; and a n.a. f ra..uung t.o 02 : w oy the at,trators -

(iv) Boron.da in the reacton.oc.i rddto:Ctz.c IT 3 ten. } 15 int:.strii.,u, tion.ofctate.y !T.;X0d g,j

,.g j 7j ggjg g

c.nmed. that t.i,e c3aove ut i

(v) Cu..an

. ant.

to reco;nize trac.istence of inau: ;uate l

a.

with the ecclant watcr.1 or equipment Containment.

reccior cere rai.m il ble and creas affected oy the centainment gg) g c3 g, oiling and preuerlud light-incirumentaticn.' ava a atmosphere,it sh y n:sumed that act water nuclear pcwer reactor IIccnsee

( ) Each pressurized II:ht. water b

less than l':0% el tne cc:e cquilibriuta shallimplement leak reduc. ion nuclear power tea ter shall be provided ncole gas inventory and *.3 o! the ccre measures go t!.at leakage finr1 tystems wiih a primc y coiant :aturaticr: meter equilibrium haia;en in.er tory are outside containment Dystrne d.ui (subcodin a':!st) 6t at:vid:s ', t*m atmosphem and ar r.du,the cont.im:. ant uni. cts:dy u:,spersad in would cr rauld ccnt:in hici.:y control Iccm a codnt o'es. r::c h u. en.

. tic.nal:TL of th" radioactiv: ficid: denn; ard Tal!owing a lin Ind:catien.cf the tr::.sary ccent core n edi;num.. Mea:n ircentory and serious it nWnt cr acc: dent)is saturati:n condhien.

pred:: cts are undern:h em, rsion I's of t. e remaicir.) ecre :

eliminated or minimited to the (3) liv Ja tucry 1. Ic82. cach boility nbated on maximum cxtent practic:ble to prevent and pressurized let. water nucicer surfaces exposed to tne e.ortainment the release of significant amcunts cf power reacter shsfi ' e provided. i*.h c

atmo p..r 2.

radioactive mat 2 rial dudn.t and instrucentation sud. cs a rea:!ce uss el (ii) The fach.ity da:J;n must be sut!*

following nn accident. CocrJdes ation water level tr. dica er wrach suppt:e4 la that en f :d'vid,ual opr !ct will not shall be given to r ducdons of potential the conMol room i reccided, receive more tnar. a 5 nn whole body release pkths that could result irc:n untmbiguous. Umctind: cation, of dose, or its equivalent te any part of the design or c03ratcr deficiencias.

inadequate care cw.ib The i:.dartica bodysthue perform:np necessar'/

(2) Each La Ucg and pre::rized li;ht.

must coverthe cc.n;.!cte ren; frera safety funct:ca duimund followmg an water nuc':ar pewer resctor licensee normnt cperation to complete ccra accident.

shall est:blish and implement a program uncovering and rive advance warrin: cf (2) In.P!bnt lodine Instrumentation. BY of preventive maintenance tieliminate the soproach o. taedNua'e cc:e ccOng.

January 1.1 PSI.8cach boiling and-or minimi:e. to the maximuni utent (4)'Allinstrum:nts used to dete:.t tna pressurized light. water m: clear power practicable leaka;c from srsteras existence ofin de:;usta cera c:c:i y reactor : hail be provided with outside coatainment.This skilaclade shall be deshned end qu:I.II:d to instrurcentation. equipraent and periodic inic; rated leak tests at perfcrm their fur.ction follmving an nssocicted t aining and precedures for intervals not to excced each refuelin2 acci&nt characia;:cd by the deternining, under accid:nt conditions.

cyc!c as well as the reduction of radioactive materiel t! esse terms the airborna radiciodine concentration potential release paths by appropriate described in paragraph (b)(1)(1) cf t!.is

'in areas within the facility where plant operator training.

section.

personnc! may be present dttring and (e) Accident 3:enitoring (g) Trcining tMiltig:te Degraded following an accider.t.

Instrumentction. Each bodir:: and Core Accidents. By April 1.1931. each (c) Sampling During cnd Following en pressuri cd li;ht. water nucicer power boiling cud pressuri: d light-wat,er Accident. By January 1.1s52, each reactor shall have the capabiinv during nuc! car power reactor licensee snall boiling and pressurized light.wate" and following an accident fan '

include in its training program for all nuclear power reactor shall be provided (1) Providing and recordin;in the operating personnel training to with the capability for personnel to control room a continuous indication of:

recogrdze, centrol and mitigate the obtain and quantitatively analyze a (i) Containment pressure by January 1.

conscquencies of accidents in which the reactor ccolant er containment 1981:

core is severely damared.The training 2

atmosphere sample during and following (ii)liydro;ca concentration in the shallinclude the use of all available an accident.

containtnent atmosphere by October 1 structures, systems and components that (1) The facility design must be based 1981:

can control or mitigate degraded core on the radioactive material release (iii) Containment water level by accidents.

january 1.19M 2:

!?an 101161b Pe5. L M M. 6 F '. ^"

'Or 2a d.a e after the crat..e 6te d the rWe, (iv! Cant.dtment radiJ:Mi IeVCI by M;; $cC. :U. 2S i.mer.M. I% L :.

> C.i what.uu$tter.

October t.1.iat:

Sta t.1:c ( I: 115.C. 0133. ::ctibs. L a l J P00R D HINAL

- ~

s G*i474 i*cden:! Nh ter /.Verl. 4.'i. No.193 / Timm%". Oc' err 2. *.'Wt / "r " m.d "

3

-s.,....,._n,

, m

~

7 D to.! s.! W.nhWo.i. U.C 61 :Mh d.ry of ofIydt g :3,wHv co.~'ar;hi.' M tN!

rcW.Tm.' N.!': P ~~ ~'

Cc ? *t&

Seph niber ?.ra ymer *!cd et T: a.S'e i !snJ.

... C.. ' % !n.4 [.C Jr. c F.>r the.% ekar Ec;%ttry f.'i. :rM c.:on.

Cyn.nissione ? s:ferd ayea v.iu DM m 11. U.. fa mu.m

• hes.o comments c>.c:pt for the Cuna. niu. Wh'

' " ::c::t ac sumuel J. U.a!;.

t Sccm:ryc/t!:e Commissi.m.

statement smt- : 8:a;t the tc!c; hnc.; (:'c1) trXm i,

pub:! cation of tha r h !2r commnt '!"

st:m c:m.;:Y ;nor: wien:

Comm,te,cer Gilkinsky's Separate effe::t proposes to retain the outdat1d Views i>vrebn Nb!ir lin, of the re~ulation:."

III8'3;I' I I' 5'h '3'U '

P Proper < d 1:u'.r.-Interira Ituiu'rcments p n o.e. m.ms rae.1 wa-m. e is.na The Nudee.r Rec de:v. thin.rrm Related to "nfm"an C#rti c-4 wze,coeg3333 i.a (NT41 is respons;ble for lice: mire and Certain De<;rar!,'d Core Comiderations rguhiti 1-' nn!c.;r pov.4 e r!re;. O im As the sucanary of tids Federal a nudgii pwer plant can be built at r.

Register Natice states at the outset. "the partituiar s;tr. a conw ayioy rcrma accident at Tntcc hhle Island. Unit 2.

10 Ci:H Part 50 must bc obumiima :r MC. As a r:sulted in... ocnerativa of hydrogen map 3rpe cf Ge apphniu fa a from fuel ch:ddiN;;-water reactien wcal qmesh Wnt of Production t.3d construct:on permit. le gpl: cant f;1r,, a in excess cf thw :onounts required to be

.. Mon acu@: Conridert.U:n or Safety lumi; o,s Repart. Dit rept rt asrumed for d:Wn purposes by the U

N P'".'"l?.jry's d.=sien cdtcra and O

current Ccmicir sica reguhtions." Yet, in auAatiort prehmir design inferr.mtion for the approvine this Notice, the Commission ACENCY:U.S. Nuc! car Reg':latory pmpmu. m.c: car p wer pl.wi nnd in effect hopesca to retain the outdated Commission.

provides ir ferw tion e : tha nrepom site. The ranort also sse s:s variou3 reauIations. Thu pr opose ! ru!n does ACT!cN:Advanct. notice of proposad abnor.aal dn itiona and r=idart little more m the crea of hydrocen ru!cmakia -

cun'rol than to prevent eliinination of a situaticcs and descr:.bes sr.:ety featurcq e

t be provicea to prevrat accidents or,if protective meast.re now in place-use of Et:Mrmav:The U.S. Nuc! car Regulatory a nitrogen aMuspbcre in the Commission is ccnsiderirq amendmj its t.yy sh;uld occur,10.7 t7.ta tpe:r f. containment of bot'im; water rccctors to regulations to dctermine to what extant chects a pab!!c hea in ana care:y.

s prevent hydrnon bming-where such comn:ercial nuclear power plants shoald I" ""'I "'P " " P " M '

climination might be permitted by be desi; ned to cepa with reactor ammts of radioactip r-iniaI are current wP atum In the ce' cI accidents beyond timic considued in SUUUUIC" d"N I"f 7 <" ' }.',,N l

pressurized watt r reactors with f::o the current "desiga buis accident" reactor fact Mousn u;n raa m.,m condenser cont inments cnd the latest apprcach. In particular, this rulemakir.g materic! sareally ru.ur..n;,a the suel 8

design boil:n:t w :ter reactors the would consider the r.egl for nuc!ca

. e'b prcpored tu e thes not advance the powcr plant dew;v to ue esaluatm.

i d inae accident condit ons. For a ppreciable

~

ammta p rce,loacum mM to be la el of pretcc"en at all. It is this aspect over a range of dataded core coo!!ng t

of the preposad rule that I disspee with, events with resu.lting core damage and

e necd k:(1esign improvements lo re! cased trem tea feel. it ms!

The existir.g tu!c (10 GR M 4 t)is based on a view that only a small C Pe with these events.

experience damnga trara one or more of y g g

amount of hydroy.c can be generated This advance notice et proposed during an accident. Prior to tha Three rulemakmg is bem; tssued to invite hydrau:ic-mechcnical nerident at Mile Island accideat, reaction of s advice and recommendations on several normal fuel temperatures con burst fuel much as one percent cf the reactor fuers questions concerning design and cladding resu: ting in release of radioactive material normally retained zirconiura cladding with water was operational imprevem. nts for denh.n3 regarded as extremely unlikely, even in w th destraded core coonng. Therefore, in the gap between the fuct p'ellets and a major loss cf coolant accident. Now the preliminary vie 3vs expressed m this the fu21 clad. A more scMous type of we know t!nt larger quantitias of n tice may change m light of comments accident involving higher fuel hydrogert might be generated in lesser received. In any ca:.e. there will be an temperatures might in addition to accidents-as much as 50 percent of the opportunity later for additional pubhc rupturing f.:el cladding. cause oxidation of the claddine. This. in turn, would zirconium claddi:;;in the TMI-2 core is c mment m cmection with any estimated to bas e reacted with water proposed rula that may be developed by cause hydregin to be generated and thereby releasing :everal hundred the Commission.

released which would compound the severity of the accident. A still more ki!ograms of hydmsen to the DATES:The comment period expires serious accident might invoh e very high containment. In proposing not to require December 31.1980.

fuel temperatures and oxidation ci a additional hydrogen control measures in ADDRESSES: Interested persons are large fraction of the core's zirconium. In ice condenser con'ainments the staff in ited to submit written comments and this c3se, not only would lar;;e amounts and Commission have argued that the suggestions to the Secretary of the of hydrogen be released to the changes brought about after the Three Commission. U.S. Nuc!aar Regulatory containment building. but other thermal h!ile Island accident have now really Commissicn. Washington, D.C. 7C353, reactions could result in the relcase of made the probability of generating large Attention: Docketing and Service radioactive material normally held quantities of hydrogen insignificant.

Branch. Copies of comments received by captive in the fuel pe!! cts. ri : ally, an That assumes a rather more complete the Commission may be examined in the accident so severe that core melting understanding of the operational Commission's Pubhc Document Room at occurs could release large amounts of characteristics of large power reactors 1717 H Street NW., Washington, D.C.

radioactive material to the environment than i believe experience shows. The Comments may also be delivered to if reactor containment inte;;rity were Commission should have proposed that Room tut.1717 !! Street NW also to be lost.

all powers reactors demonstrate'an Washincton. D.C., between a.15 a.m. cnd Based on these conciderations, a ability to cope eficctively with amounts 5.00 p.m.

broad ranc,e of nuclear power plant abnormal conditions and accidents with P00R D Elim

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I a.dequact of ;! : r t.rpact et r.utcar r;gra t.::. that rcya cu rmmtn e

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power piant des::n.

deshn, nwltiple pbvricalintrie:s.

This re.pe:t r-cemrnen&>! S it ^-s

• As discurnd in T 1!c 10. C' aater t, gr!ity cscu cr:0 f;r t'a vn.

etudle: M 'a n vi:'r?J r.f s'! i. ME 8

Code of Fd.: d naulatioim ;i *mila),

manufacture and ortat:en and lou-of. m t.ntn.:ddc - "i en..se-In the Saft v h lviis itcpart t!:e continur.d survul: a:: cud testin:po fa!! crc nerid: P.s. vi.h ;:aur app!! cant jiw,"ii:d to detache Trevet such accim at;.

attent!.m to hrman hii" a. ' :. repart margins cf rWtv fer bei er d and Fu:thennore. b invi min mer.tc.-

stated tY.: "Arn these :: ! n cnv chnet.nal e p::n.:na :nd te d.trrrafna plant dar'ps usin ' Be "dni:n b. de energe dul: M: modi. nti::. in the i

of structures, sytte.ns, acci:ient" approach, the NRC ducs cet design of S r.!s that w:0 h? n revent

?

"the adeque'nts provided for prevention revirw all structur:s. syskms. 4,nd accident : r.nd mitigate the,it and compcne ef accidar. s cn;f 1:.: m:J r.t!v t c.f t!:e cmrpac:L: bat r*.ther 7: view. in co :seg:- * ; Ter exat F.?

consecuercer cf c:cidens." To essirt var zin t Inc!s of detuil, eniy thesr*

consideration should lu *en h the. applict.r.t l'100:*.*:lyi':q w;'h th:s coa. !dctrd " safety "rada" l'y tho equipm(nt that wocid f a. 4!Lt: the reoula tion. thc.NF.C ha= publ;bd app!:cr.nt submittbg a S:.fety Ana'ysh contrud safe vent:n; r3 h+an gss Plgulaterv CWde 1.70. Standard Format Report. Items conr:dned by the frorn the rci.ctor con!im :yst.: n." and and Confer.t cf Edi.v Ana'vs Reports upph::mt to be oatrid: the s: ope el

" consider., tion should Wnn to deri n b: sis accident analy. es at:

overall p. tight crck 'e cIthe Id-for Nuclear it: vcr P!rr.ts.' which 3

describe: the %formr. tion to be provided g:nercHy not ccasidered to be "sa.*r.tv down/traFe "; system n'h 19 ption in thn Safcay w:c!p:s Report. Lt gt:d " and are net rt: viewed by the N:iC of rr. turning : m s to the et;nts.:ncat n::2try to rce whether thy wf;; perfurm a building."

1.aitirol.r. v enon t hiRc:h.a to an intended or meet vazioca dependa!J:;ty Similuly, the ja~tary 10:5 r. mot:,

1 Guide 1.70 pioWJ:s gmdan

e. poli:aat ccace::.ing

• d: sign bat:4 criteria. 'lhis mcthed of d.isti:Ic.ibn I.

Three L' ' : I&nd. A T..

m o f:a assumplicas cr:g::b!c to d.c NRC fer bared en the notka that thinas credited Co neic r. r r.d M..v P. ' ?" '

purpota= cf c:!:rmin7 cdu.ua:*f of tha in the analysis of a desi:n basis eu.nt cr stater. "* *

• we h. v: cm kr beyed ulant des!;;n to m?et 10' Cat ;M:t WJ spedfhd in the regulaticas are the reint at w'ach the cxi cr '. cided criteria." Ec :iatery Guide 1.ru ewlains impcrte :t to safet" :nd thcs arc "rufaty desh:n tatis =.crilent rev!aw. ":rnch is softMer'. 'l Fa t recces n F. '"e I i

that these dw *a t%s erv:nntio.a can, tec.ia" whi'c all e!.e is "non-cafe:y enough in pir*.'t mary im; c: tant for the most put. be found i: rmhtory g:ude." Non-safety pada items do not guides it.at dnl with radie!cgical receive conimuin; rcra!atory design wedneue? or to oddess R the releases and s::msts ute ci Replatory surcrvision cr eurvei;ktte to see that relevant d.e gn inrues. Sc.no irpriant Guides 1.3 cnd 1.4. Assumptions Used they cre properly nnintained or th.t accidents are cuiside or r.re not l

for Evaluation of the Potential their design is not chann:'d in :ome way adec.uately arscesed u ithn tu ' des:gn Radiological Cons::;tenecs cia Los&of-that : night Ir.terac! nera:ivdy with c4her enve!cpc': 4 v. stems 4.rc not Yifaty 4

Coolant Accident.8 Regulatery Guid1 systems. Iastrad. thcsc items simply rc:atca, : anc mtc; ration of hurnan 1.70 further states that "This analysis receive what attention rnay be dict ted - factors into tha desisa is cross!y should be referred to as the ' design by rout:ne indestrial codes and by inadequate."

basis analp It.'" Cperating events desires to enhance ;!an cra!Iabil,ity.

Commisskn's htentiocs correspontsn; to design basis Historically, a further asmnpuen in

-Accordircly, it is t.ne Conan.saian.s.

assu:rptions r.re terrred " design basis design review and licen !rn was that if acddents." and satisfactor'/ ana!pis reacter plant systems can. hand!c la:tc.

intent to deterreine what chanus,if any, conclusions concerring them a!1ow a scale design basis accidents, they caa m react c plant d:si;;ns ard safety judgment that the facility can be also handle a spectruni of smaller analyses am needed to tue mio operated without undue risk to the accidents that are regarde.d a2 being account reactor accidents b: yen. thpse health and safety of the pub?ic.

"within the design enulare.-

considcaed in the current des:;n b:.se It should be noted that these events The accident at Tlwe Mile Island accidc nt oppmach. accidents imder are analyzed pdmarily for the purpose resulted in care da na:;c rnare ses cro considerat:on mclude a range cf ass.ot-COM cculing, core d:. mage..md carc-of establishing the adequ acy of than that considered in cutrent desisa melt:in; evcat'. bath msico and outside engitieered safety features, such features bcir.g those struct tres, systems, and

'There are oder desi:n rm;uirements which historical desgn enve! orcs.

components, designed into a pant to would pieswpo.e events s Arc s.ergr.: ccre Furthermore tne Commissicn will

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• I i* "'i" "'"* ' ' d 5 mitigate the cons.equences of po.stulated occu red. I ur e=a:rp;e. red.oacme source terms of

.g,mer, tW m no retene of fWen design bas,s accic,ents, and which 7,a..:ia,i igo,ma.2,n o.,umni ::a. wn p.,3,3,m., e,,,,cn,,,,,,,,,,, g,,, g,,

i supplement other plant fcatures c :ce t.on of D.str.n:e rert..r. fer r#.ser.nd Te.t prewn.<J in a cnt rett too or n.).sm4.

designed to meet performance -

m artar 5 :*5.*v !'ich tvrty a r aiar resciar

>-The nea f-r cs.n=e: The txevy of Tut.-

acc&nt, ernse teu W4n adman,M avm! t>le fact et U S. Governr eh.t Pnm.n;CMce.

sEC.Cifications for normal opcrati0ns 3nd vario.as en;:.neered :amy fe, tun s and rettain ether tv.e.hingon. D C.

3 02, ant 1Ctpated abnormal conditions.

plant spt.=ms and cunrents.

C@es m.y tm obtained frorn th? CTO S !es 8 \\. 4 : L f e.i

< si L '.N: 41. *.. c.

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r.J i

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r Cosim.a.ivn, h.%.gr,n. DC :35ss Virpnha ::131.

Commisuun. b danpop. D.C. u;5s.....

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i coce darf en"./: bt tr.tr d :41 ci red :r :) n a ta.li;ily, ivi pr.N.3ie m to tr~r M. n;.; ;uan'i6..;l bwh no..tal operat'r;;;pte:n3 and de sign t~.*ov.:nient.:t.. ta.t. v.c y nut.

rii.irn ratv. ul. e nad 0.v.. u.

a c@:et..cd saf-ty f e.:tures. Tharciare, or, if 59. u./:t de $!ca i.TN o.

en:s ran v.m.!.! 3 cu t!N m rett a stoc tm.=

th;s advtr:co notic 2 of pru[nused bs taude tend ! hhut o;... a;a.. cost?

L;n,.!e a:td at % rat s.Tian.; c;..d..+..v/

nica.alina is bei:rt puWhed to povide How wonH your recoim.ta.!ailons rnd c st? Do th e u nith! tcdimm a it.e r:::; bted !nde :; cnd the pt.'.U::.n aff:ct rM.a.faty cena tvNs!

r!:'.e :pected C : m.h a r.t:-. :::

epper' unity to pr. @ v!v!:a and

. & Th W.tle Eni rec 4 nt was cTT.nt pc'enti ' 'm ;.us in F:t !..at scenrrmenditions in the Comminion em termine.ted after the cere wan Jamnert me materie.iiw fri.-inced mis such.rs wlut ehvald he ine conte,t of a

. severMy b'.it beTera cubs.im!.i m!Gc;t 9v.ct nt o r wr e regulation ;equirin3 imDrovementa to occurred, a condition bcV0nd ti.e cuttent g.on: catratio:a of hs c: ogen in the enpa with de<;radm! < nre coolinpnd designAa ts-accidont <:venta consl& red r+ n in.;npp.' e..?

whh nccidents r:ot coverca adequately in the safety anslysis. !N.:!J t'.e iG.C

7. Sheald th: NJC require by trcditional des!p t anedopes.Th.

require that events of th:2 type b:s inwraration 13 0 t.cn' air.rce nt d rien, rule nid.:c:t prec eir. will acdres, the cons:dcred in fatur's

<fr.'y c.t.! pes? If systms for co:.t nj :.,: combus: ion 6.

objectivos of such r. regulntion, the not, tyhy not, or, if so. v.; ::t criteris byFcen? Da 'w *avor methods r/

des %n and cporctiopul haproverrents would ycu wpose to Jtge det!gn control thst M"e ; co'r.burion cr do being conridercd. th.nftect en othcr acccp?chil;ty7 you favor cont'rld:e4 b.rnin;;7 ff you safety considerst!on i. and the costs of

3. Althou;h tne consequences of core.

f,,s or suppression cf corabustion'. whet such des!:'n improvments ccmpared to melt sent't:nts nave 1 cen con @.ered to g 3.,. rg,33.,.geg ;. a teammnd s.l.

t cvected beneats.

some evtr.nt m asassing :n clear power

,, y,.2 they ve>~.o e I'metion of the Rece;,rdzing the need fur promp:

plant ia; ty. sucn as in rcTaitements tur k

deti n:apabWi. of cu rent i

actioci to corract crecific deliciencWs siting, ememency response phn:, ar.d cor.tahunents't If 4,a fen or contro!!erl 3

ideat.fisd durin:t the Three Mile Island certam ewineered safety t:ature:.

botning. do yr.u rccommend opeo accident and subscrivent investi;;ations, expbe:t cont.tdcration of t,.:'tcr cont :f.; ment ty ped !! so. what undt.rsti.adir.; the safL:r ad e ta,ss shifercutes wonid y ou recomme.nd! II:r.v and disadvantages of th sm.ral do 3 uur recommendam.as af!cct oth.r featurc4 mcnlien :d in freticus

. r.afc!y ccnsiderations?

questions?

10. hon:d the N!!C rer,uire desi:9

10. In wciWag the co:M of des!gn cnd J

thw;_c to account for increased operatimmi improvemen't to c,tn with radioactive material that may be degredcd ccre cochng naainst the

rentpc,rted durin; w accident hy.

benefits of thcir use, what <;aanti;a!!ve systeins normahy senctioning with much method 2 or other guidance would yc,a tome levels of radicactivity such as the ru~:est to faci 9 tate prap.vation of a

~

steam and resijual heat removal useful value-in.p set esacssm4 nt? Would s.ystems and the cor.talmnent dra!n:;e you consider useful or a; pro; riate 8F5?cmi compar: tens between rur: mar p r.ver

11. Shocll the NRC require more plant risks crd other risks to ri:ich extensive orera:cr trcining. strict littral peopic cre exposed?

comp!!ance mth emu and improved

17. Whct ra:pects of doraded cooling detailed operatin; precedures. Increased or melted core accidents i.re sufficiently re!!abdity cf emer::cocv cooling or dertsy unknown <*r unecrtain at to h pede heat removst capw.aty, and exp.mded design end ana:ysis of mitWting contml room mirdm:pn manning as systems, thus requiring additictaal alternatives or suupic:nunts to degraded research or experimentation' cool.n: demn improve.ments?

13.T1.c NRC has under way a

12. Should the NRC require an separate rulemak!ng proceeding

+

alternate add-on, relf-canlained decay concerning reacter siting and an heat removal system to prevent emergency planning rule has recently do;;raf ation of the core or to cool a been approsed. If you are farailiar with de;;teced core, m centrast to the thcre sernate activitics, hcw muld you pytvic.aJy discersed schemes which are modify pretent and proposed aim.d tcward mit ;ot!.e; the requirements for emerge::cy pl.mning cons:quenc.es of 6;;n,ded core cechn:r?

and ree tor siting if accid:nts Levond liow ucu!d such a daeay heat ren:ov il the present design basir were fibe

(

,3 T considcred in nuAar jowerplant safcty analysesi O. Should the NRC require syst:ms such as the makeup and puri!Ication Cated at Wuhirgon. Di' :his :wh day c.f systems ta be located in o leak-tight September 1053.

building? Would :ues a requirement add For the Nac!est Regula:ory Commissicn.

to or detract from everall plant safetyf Samuel J. Ch;Ik.

14. V! hat design. q2ality and seisreic secrererf. cf m commitrion.

critaria would you recornmend for any tra ce em r. mum r.;

additional systems to prevent the cumo er,3 7m.c.,g potential breeching of containment such as systems for controlled fdtcred l

l Venting, hydrogen combustion control.

and core retention mentioned in previous questions? Da you favor evahmting designs of such systems on a realistic t esis, as cpposed to the conservative method used to evaluate engineered safety features? Do you favor establishing design criteria for such sysicms that ere equally stringent.

less stringent, or more stringent than l

those applied to engineered safety I

features? please explain your response in terms of criteria you would I

reccmmend, including consideration of l

redundancy, diversity, testability, inspectability, and structural design I<

i limits (including seismic requirements).

15. Can probabilistic analysis be used both as an aid in determining and comparing the adequacy and usefulness of the several featurcs mentioned in previous questions and as an aid in determining the design enteria and Enclosure "A" i

ENCLOSURE "B" l

t l

l I

t

~y

s f

COMMENT LETTERS FOR INTERIM RULE Letter No. of No.

Date Organization Commenter Comments 1.

11/03/80 Edison Electric Institute J. J. Kearney 1

2.

10/30/80 L. P. Leach 6

3.

11/03/80 Philadelphia Electric Co.

J. S. Kemper 5

4.

11/03/80 Conner & Moore M. J. Wetterhahn 2

5.

10/31/80 M. I. Lewis 10 6.

10/31/80 Bechtel Power Corp.

A. L. Cahn 9

7.

10/29/80 BWR Owner's Group D. B. Waters 2

8.

10/28/80 L. J. Ybarrondo 1

9.

10/30/80 J. F. Kowalski 1

10.

10/31/80 Fluor Power Services D. M. Leppke 3

11.

11/03/80 Offshore Powc. Systems P. B. Haga 5

12.

11/03/80 Commonwealth Edison J. S. Abel 3

13.

10/29/80 Ethos Research Corp.

A. Bates 5

15.

11/04/S0 Atomic Industrial Forum D. C. Gibbs 11 16.

11/05/80 Yankee Atomic Electric Co.

D. W. Edwards 6

17.

11/06/80 General Electric G. G. Sherwood 9

18.

11/04/80 Stone & Webster R. B. Bradbury 7

19.

11/03/80 Consumers Power Co.

T. J. Sullivan 4

20.

11/03/80 Georgia Power W. A. Widner 1

21.

11/04/80 Westinghouse Electric Corp.

T. M. Anderson 13 22.

11/06/80 C-E Power Systems A. E. Scherer 4

23.

11/07/80 Florida Power R. M. Bright 2

24.

10/30/80 Omaha Public Power District W. C. Jones 2

25.

11/07/80 Duke Power Co.

W. O. Parker 1

26.

11/10/80 Virginia Electric & Power Co. J. H. Ferguson 7

31.

11/13/80 Washington Public Power Supply System G. D. Bouchey 7

32.

11/04/80 Alabama Power F. L. ".layton 1

32a.

11/18/80 Philadelphia Electric Power J. S. Kemper 1

33.

11/17/80 Public Service Indiana S. W. Shields 3

34.

10/30/80 County of Suffolk D. J. Gilmartin 1

35.

11/17/80 Northeast Utilities W. G. Counsil 12 l

36.

12/08/80 Long Island Lighting Co.

J. P. Novarro 2

37.

12/17/80 Arkansas Power & Light Co.

D. C. Trimble 8

l 38.

02/03/81 Dept. of Energy A. J. Pressesky 3

40.

02/09/81 Consumers Power Co.

D. P. Hoffman 2

160 i

l 1

Enclosure "B-1"

)

F 1

e TALLY OF COMMENT LETTERS 35 applicable comments have been received with the sources distributed as follows:

nuclear steam system suppliers 4

utilities 18 architect / engineer firms 3

law firms 1

non profit societies or assoc.

3 individuals 4

government agencies

_2 35 Note:

1 - Comment 14 was for Long-Term Rule 2 - Comment 27 applied to another Federal Register Notice 3 - Comment 28 was for Long-Term Rule 4 - Comment 29 was for Long-Term Rule 5 - Comment 30 identical to Comment 17 6 - Comment 32 same number applied to 2 different comments 7 - Comment 12 applies to Interim Rule and to Long-Term Rule 8 - Comment 39 included in 38 2

Enclosure "B-1"

i l

CATEGORIZATION OF COMMENTS i

Category Area Letter /Coment G

General 1/1, 2/1, 5/1, 5/2, 5/3, 5/4 6/1, 6/2, 7/1, 7/2, 8/1, 10/1, 11/1, 15/1, 15/2, 15/3, 15/4, 15/1, 17/1, 18/1, 18/2, 20/1, 21/1, 21/2, 21/5, 21/6, 22/1, i

26/1, 26/2, 31/1, 32/1, 33/2, I

35/1, 37/1 l

M Recommended Additions 2/6, 5/5, 13/5, 31/3 1

Section 50.44(c)(1) 13/1 2

Section M.44(c)(2)

None 3

50.44(c) 3)(i) 2/2, 5/6, 6/3, 12/1, 13/2, 15/6, 16/2, 17/2, 18/3, 34/1, l

35/2, 36/1 l

4 50.44(c)(3)(ii) 4/1, 4/2, 5/7, 6/4, 9/1, 11/2, l

11/3, 11/4, 12/2, 13/3, 15/5, t

15/7, 16/3, 17/3, 18/4, 19/1, 21/3, 22/4, 26/3, 31/2, 33/1, 37/2, 38/2 5

50.44(c)(3)(iii) 5/9, 15/8, 17/4, 18/5, 21/4, 24/2, 35/3 6

50.44(c)(3)(iv) 3/1, 3/2, 5/8, 15/8, 17/4, 18/5, 21/4, 24/2, 35/4, 38/3 7

50.44a(a) 2/3, 5/10, 6/S, 6/6, 17/5, 19/2, 23/1, 32a/1, 36/2, 37/3, 38/1, 40/1 3/3, 6/7, 11/5, 12/3, 13/4, 8

50.44a(b)(1) 21/7, 22/3, 24/1, 31/4, 35/5, 37/4 9

50.44a(b)(2)

None 10 50.44a(c) 6/8, 10/2, 15/9, 16/4, 17/6, 18/6, 19'7, 21/8, 31/5, 35/6, l

35/7, 35/8, 35/9, 37/5 11 50.44a(d) 21/9, 31/6, 31/7 i

l 1

Enclosure "B-2"

d i

i 12 50.44a(e) 3/4, 3/5, 6/9, 10/3, 16/5, i

17/7, 19/3, 21/10, 21/11, 26/4, 26/5, 26/6, 35/10, 35/11, 37/6, 37/7 13 50.44a(f) 2/4, 2/5, 15/10, 15/11, 16/6, 17/8, 17/9, 19/4, 21/12, 22/2, 23/2, 25/1, 26/7, 33/3, 35/12, 37/8, 40/2 14 50.44a(g) 15/10, 17/8, 21/13 i

2 Enclosure "B-2"

I l

\\

g s

LIST OF COMENTS 1.

Edison Electric Institute - Non-Profit Association COMENT 1 (Category G):

There is a need for the interim rule in order to l

provide a basis for continuing NRC licensing activities.

2.

Lawrence P. Leach - Individual COMMENT 1 (Category G):

The interim rule should be withdrawn until NRC establishes policy regarding hydrogen control, degraded core cooling, plant accident instrumentation, etc.

COMMENT 2 (Category 3):

Most likely that no containment should be inerted if total safety of public and workers is considered.

COMMENT 3 (Category 7):

The existence and use of the high point vents may result in a net decrease in overall plant safety considering the additional complexity of the system as well as unanswerable questions of vent discharge, vent testing and reliability, operability, etc.

This item should be deferred and addressed by NRC research.

COMMENT 4 (Category 13):

The subcooling meter is of marginal value:

1) during a small break LOCA the reactor system is in a saturated state

2) the plant operators don't use it since it is not used during normal operations.

COMMENT 5 (Category 13):

Water level indication in the primary system would be valuable during small break LOCAs.

The January 1, 1982 implementation date is optimistic.

l COMMENT 6 (Category M):

It would be desirable to have a direct t

indication of whether or not natural circulation flow has been estab-lished.

One method that should be considered is the difference between primary and secondary temperature.

3.

Philadelphia Electric Company - Utility COMMENT 1 (Category 6):

External recombiner capability should not be requiru until a final decision has been made on the, gas generation rates that should be accommodated.

COMMENT 2 (Category 6):

Suppliers have indicated that the January 1, l

1982 date for completion of the recombiner capability backfit cannot be met.

Delay time is estimated as two years minimum.

COMMENT 3 (Category 8):

The proposed rule should address how or if airborne radioactivity outside of containment should be handled.

COMMENT 4 (Category 12):

A continuous indication of containment H2 should not be required for inerted BWRs since the combustible gas control system is already in operation.

Batch analyses of atmospheric samples should be sufficient.

1 Enclosure "B-3"

COMMENT 5 (Category 12):

The proposed implementation date for the installation of the noble gas effluent monitors and radiodine and particulate monitors do not provide sufficient time for the necessary acitivities.

4.

Conner and Moore - Law Firm COMMENT 1 (Category 4):

Design analyses should not be required until the long-term rulemaking is concluded.

COMMENT 2 (Category 4):

The design analyses have not been sufficiently specified to provide uniform and comparable studies from different licensees so that they can be reasonably compared and utilized.

5.

Marvin L. Lewis - Individual COMMENT 1 (Category G):

The proposed changes have as much potential to increase the risk of DCAs as to reduce the risk.

COMMENT 2 (Category G):

Best way to eliminate DCAs is " Shut 'em Down, Foreve r. "

COMMENT 3 (Category G):

An implementation schedule should be required to prevent stalling.

COMMENT 4 (Category G):

Pre-TMI NRC staff is still in place.

Another approach to nuclear safety is needed to replace current NRC process.

COMMENT 5 (Category M):

Workers and operators must have means to determine their personal safety, or else they may panic and leave their posts.

COMMENT 6 (Category 2, 3):

Actual in place test data to measure the effectiveness of inerting systems needs to be generated and/or published.

COMMENT 7 (Category 4):

Design Analysis is insufficient.

In place tests are necessary.

Chance of H developing over the long term must be 2

considered.

Consideration of blocking of cooling lines and erroneous instrument readings due to small amounts of H must be considered.

2 COMMENT 8 (Category 6):

Rule should require that hydrogen recombining actually be installed rather than require just the capability to install them.

COMMENT 9 (Category 5):

The number of containment penetrations should be minimized.

This requ" ement can be counterproductive.

In some instances, "non-conflicting use" may be preferable to " dedicated."

COMMENT 10 (Category 7):

There are circumstances where the migration of gases would be stopped short of the high point (e.g. blockage of debris).

Also, the high point vents must be tested in place.

2 Enclosure "B-3"

i i'

6.

Bechtel Power Corporation - AE COMMENT 1 (Cateacry G):

Provisions can be implemented without the need for a rule change.

The R,G. approach should be used in view of the detail involved.

COMMENT 2 (Category G):

Rule names items previously agreed to less formally.

Rule should permit approval of items previously accepted.

COMMENT 3 (Category 3):

The benefits of inertir>g are questionable.

Other options than inerting should be permissible.

COMMENT 4 (Category 2); The design analysis required by the proposed Section 50.44(c)(3)(ii) appears unnecessary.

The staff analyses given in SECY documents are sufficient to justify not inerting Mark III BWR and all PWR containments.

The NRC rules should not be employed to direct studies and open ended modifications in the absence of specific criteria.

COMMENT 5 (Category 7):

Reactor vessel and pressurizer venting is sufficient on all but B & W plants.

COMMENT 6 (Catecory 7):

Single failure criteria for the high point vents is an unjustified escalation of previous NRC requirements.

COMMENT 7 (Category 8).

Clarify source term with regard to amount of coolant included and noble gas treatment.

1% particulate plateout source is a new requirement.

COMMENT 8 / Category 10):

Boror, analyses should only be required for PWRs.

COMMENT 9 (Category 12):

Requirement for continuous noble gas effluent monitor indication is inconsistent with clarification 4(a)(4) of II.F.1 in 9/5/80 Eisenhut letter.

7.

BWR owners Group - Utility Association COMMENT 1 (Category G):

Evaluation of hydrogen control capability beyond 50.46 requirements should be part of long term rulemaking.

~

COMMENT 2 (Category G):

Integrate requirements of interim rule with those of 9/5/80 Eisenhut letter.

8.

L. J. Ybarroado - Individual COMMENT 1 (Category G):

Both interim and long term rulemaking documents in the Federal Register should be withdrawn.

The NRC needs to establish and submit for comment its policies on the content of these before issuing specific technical specifications.

9.

John F. Kowalski - Individual COMMENT 1 (Category 4):

The design basis for assuming the high percent-age (75%) of fuel cladding reacting with water is not understood.

A 3

Enclosure "B-3"

J

?

more realistic approach would be to develop plant-specific relationships between fuel clodding-water reactions versus challenges to containment integrity and then to establish threshold areas.

10.

Fluor Power Services - AE COMMENT 1 (Category G):

Recommend that final form of this rule omit statutory deadlines or insert some provision that gives ficxibility to address missed deadlines.

COMMENT 2 (Category 10):

Some of the sampling information required would have little or no practical significance.

The NRC should review EPRI's recent report on post-accident sampling and revise the rule accordingly.

COMMENT 3 (Category 12):

Implementation deadlines for rules on accident monitoring instrumentation should not be set before Revision 2 of Regulatory Guide 1.97 is issued.

11.

Offshore Power Systems - NSS COMMENT 1 (Category G):

It is essential for the Commission to state explicitly their intention to utilize the interim rule or a basis for licensing until the final rule on degraded core matters is developed.

COMMENT 2 / Category 4):

Acceptance criteria should be established before applicants are required to analyze the need for hydrogen control and to evaluate mitigation schemes.

Applicant should be required to analyze only the scheme proposed for Commission review (not to evaluate all of the schemes listed in the rule's discussion).

COMMENT 3 (Category 4):

The discussion section indicates that sensitivity studies are to be conducted over a range of H2 concentra-tions.

It should be made clear that these studies need not consider concentrations greater than the 75 percent limit.

CLMMENT 4 (Category 4):

The discussion section states that containment capa'ility evaluations be based both on conservative and realistic o

assumptions.

Only the realistic evaluation should he required as the basis for decision making.

COMMENT 5 (Category 8):

The discussion section states that systems not specifically identified to perform a post-accident function but whose operatio9 may be of "significant value" must be assured to operate.

This statement should be clarified.

12.

Commonwealth Edison - Utility COMMENT 1 (Category 3):

The inerting requirement for BWRs should not be proposed without a risk reduction basis.

No public domain inalyses exists to shcw that hydrogen combustion is possible or will exceed ultimate containment capacity.

Operator risk is increased by inerting Mark II contaiments.

4 Enclosure "B-3"

s e

COMMENT 2 (Category 4):

Studies from Sandia (including NUREG/CR-1561) indicate that substantial R & D must be done before reasonable assurance of mitigation features can be obtained.

The short studies required by the rule will add nothing substantial to the record while wasting industry and NRC resources.

COMMENT 3 (Category 8):

The assumed atmospheric dispersion of 25% of the-core equi horium nalogen inventory should be reassessed as a function of the type of transient or accident.

Iodine releases are likely to be i

smaller due to retention in water or steam.

i 13.

Ethos Research Group - Non-Profit Group I

COMMENT 1 (Category 1):

Hydrogen conditions following a real LOCA may be much more severe than expected from the " postulated LOCA." This phrase should be replaced by "any LOCA" in Section 50.44(c)(1).

COMMENT 2 (Category 3):

All reduced-strength or small containments of the ice condensor variety should have inert atmospheres or a system for inerting the atmosphere in the event of a LOCA.

Ignitors in ice con-densor plants can result in sizeable pressures during operation and also cause local temperatures which could coabust ice chest foam resulting in possible containment failure.

Inerting should be required for all LWRs.

l COMMENT 3 (Category 4):

The design analysis limit of 75% metal-water reaction should be increased to 100% since the addition of water to a degraded core may occur at any time during a LOCA or post-LOCA sequence.

I COMMENT 4 (Category 8):

To be conservative, a value on the potential release of radioactive material from fuel to coolant or containment should be not less than 100% of the noble gas inventory,100% of the halogen inventory, and 100% of the fission product inventory (or residual fuel inventory) should be used.

Unusual reaction of the degraded core with coolant and residual steam may exceed normal design expectations.

COMMENT 5 (Category M):

Mixing is essential to many hydrogen control systems.

Therefore all air circulation components must be designed to function during any accident.

15.

Atomic Industrial Forum - Non-Profit Association COMMENT 1 (Category G):

The intention to use the interim rule as the basis for licensing approval should be stated in the interim rule.

COMMENT 2 (Category G):

An integrated approach is :> commended for rulemaking including the development of a safety goon and methodology, degraded core considerations and establishment of minimum required engineered safety factors, and reactor siting.

The existing emergency planning rule should be re-examined after completion of the above.

COMMENT 3 (Category G):

In some cases, the rule is insufficiently flexible to allow for plant-to plant variability.

1 l

5 Enclosure "B-3"

o

/

COMMENT 4 (Caieocry G):

The rule should have provisions to prevent nullification of compliance agreements already reached by the Commission with licensees.

COMMENT 5 (Category 4):

NRC staff positions taken in the SECY-107 series along with 50.44(c)(3)(iii) and 50.44a provide a practical basis for licens-ing.

The regulation should not require design studies.

COMMENT 6 (Category 3):

Inerting all Mark I and II BWRs does not appear to be justified.

This should be reviewed to assure the correct cost /

benefit decision has been made and that the result is an overall reduction in risk.

COMMENT 7 (Category 4):

If design analyses are required, they should be based on those accident sequences that are major contributors to the risk.

Implementation of design changes must await the conclusion of generic rulemaking.

The interim rule should only set dates for design study schedules to be submitted.

COMMENT 8 (Category 5, 6):

Internal recombiners or other alternate means of hydrogen control should be allowed as well as external recombiners.

COMMENT 9 (Category 10):

50.44a(c)(3)(1) - There is no set of radioisotope measurements that can determine the degree of core damage.

Rephrase this to say "to indicate the extent of core damage" or "to confirm that significant damage has occurred."

COMMENT 10 (Category 13, 14):

Training should be addressed in 10 CFR 55 only.

COMMENT 11 (Category 13):

No good water level indicators are currently available for PWRs.

Installation date should be established after an NRC l

approved system is commercially available.

It is not necessary to require readouts for the total range of uncovering.

16.

Yankee Atomic Electric Company - Utility COMMENT 1 (Category G):

Codification of existing specific requirements is unnecessary.

Detailed requirements should be handled by documents other than rules.

COMMENT 2 (Category 3):

Inerting may increase the overall risk due to decrease in containment accessability.

The MIT Study (enclosed as IAEA-CN-39/212) and the SECY documents support this.

Inerting costs are not small; downtime cost due to inaccessability should be considered.

COMMENT 3 (Category 4):

Requesting hydrogen control studies via the regulations is inappropriate, untimely, and wasteful.

Recommend that RES handle this.

COMMENT 4 (Category 10):

Chloride sampling per 50.44a(c)(3)(v) is difficult, expensive, and unnecessary.

l l

6 Enclosure "B-3"

4 COMMENT 5 (Category 12):

Vendor information says January 1,1981 implementation date for accident monitoring information can not be met.

COMMENT 6 (Category 13):

No good reactor vessel water level indicators exist.

Delete this from rule.

17.

General Electric - Vendor COMMENT 1 (Category G):

The interim rule as presently formulated is unnecessary and should not be implemented.

Should be revised to accou'it for already agreed-upon positions and impl3 mentation measures.

l COMMENT 2 (Category 3):

Several unique inherent design features in BWRs prevent substantial H2 generation.

Inerting is thus unnecessary and not recommended due to risk to plant personnel and reduction in plant opera-tional safety.

Differing views expressed by the NRC Probalistic Assessment Staff and this interim rule should be resolved.

COMMENT 3 (Category 4):

Performing analyses in response to the interim rule is premature and wasteful in light of the longer term rulemaking i

l proceeding.

Design criteria needs to be established.

Use of PRA evalua-l tions for various design modifications could measure overall safety.

Recommend this section be deleted.

COMMENT 4 (Category 5, 6):

Internal recombiners or other alternative means should be allowed.

Recommend implementation date to be two years l

after the rule's effective date.

COMMENT 5 (Category 7):

Item (2) in the last line of 50.44a(a) should be reworded to preclude the inadvertent opening of a S/RV which provides venting in the BWR, since this event is analyzed as part of the design basis.

COMMENT 6 (Category 10):

Items (iii), (iv), and (v) of 50.44a(c)(3) whould not apply to BWRs, per Eisenhut's 9/5/80 Clarification of the TMI Action Plan.

l COMMENT 7 (Category 12):

Recommend implementation date for containment j

water level monitors be changed to "first scheduled outage."

Plants with Mark I containment would require 3 - 4 weeks to drain torus to install l

water level taps.

l COMMENT 8 (Category 13, 14):

Training requirements should be in 10 CFR 55 l

only.

Post-TMI emergency operator guidelines should be fully implemented in plant specific procedures and assessed relative to the interim rules training requirements before implementation is set.

COMMENT 9 (Category 1~1,:

It is not reasonable or necessary to cover the complete range of c.e uncovering either for safety or operator information.

7 Enclosure "B-3"

o f

18.

Stone and Webster Corporation - AE COMMENT 1 (Category G):

Interim regulations should not be issued until rulemaking proceedings have been completed.

They may mullify previous agreements.

Recommend formal hearing procedures should be provided for the interim rule.

' COMMENT 2 (Category G):

Schedule dates for modifications in 50.44a should agree with those in the Commissioners' 10/28/80 Clarification to the TMI Action Plan.

COMMENT 3 (Category 3):

An adequate basis is not given for inerting BWRs.

Inerting may increase overall risk.

Application of inerting should be reviewed to assure cost / benefit decision is correct and reduction of risk is significant.

COMMENT 4 (Category 4):

The amount of hydrogen generation should be based on plant-specific evaluations.

If the analyses show that this amount can not result in containment failure, no further evaluations are necessary.

The interim rule should establish dates only for submitting design analyses schedules.

COMMENT 5 (Category 5, 6):

Modify sections to allow use of alternate means of dealing with H, other than external recombiners.

2 COMMENT 6 (Category 10):

No readily identifiable radioisotope measurements to determine the degree of core damage.

Modify rule to require only con-firmation of significant damage.

COMMENT 7 (Category 10):

50.44a(c)(3)(iii) and (iv) are inconsistent with the positions on baron and dissolved gases sampling stated in the 9/5/80 NRC Clarification letter of TMI Action Plan.

The rule should be modified to reflect the letter's positions.

19.

Consumers Power Company - Utility COMMENT 1 (Category 4):

Evaluation of mitigative features should only be required if the results of the baseline analysis are unsatisfactory.

Evaluation of mitigation features should be deferred to the long term rule-making.

l COMMENT 2 (Category 7):

The rule should only require the capability to l

vent noncondensible gases to the extent that will insure core coolability.

Reactor vessel head vents may not be necessary.

l COMMENT 3 (Category 12):

Requiring the quantifying of concentrations of radioiodines and radioactive particles in airborne effluents at each anti-cipated release point is impractical, as reflected in Revision 2 of Regula-tory Guide 1.97.

COMMENT 4 (Category 13):

Reactor vessel water level indicators do not exist at present.

Inadequate core cooling can be determined with ex' sting instrumentation and proper operator training.

l i

8 Enclosure "B-3'

s 20.

Georgia Power - Utility COMENT 1 (Category G):

Promulgation of the interim rule is premature and imprudent.

There is a lack of sound technical bases for the design parameters.

The results of the long term rulemaking should form the i

technical bases for additional prevention and mitigation mequres.

Ade-quate provisions are being incorporated through Commission Orders, Reg.

Guides, and Action Plans.

21.

Westinghouse Electric Corp. - Nuclear Vendor COMENT 1 (Category G):

Urges NRC to formulate integrated generic rule-making proceedings for consideration of safety goals and risk assessment methodology, minimum required engineered safety features, degrac'ed core cooling, siting and emergency planning.

COMENT 2 (Category G):

Interim acceptance criteria is essential to the j

resolution of generic safety issues.

Interim criteria should form the basis for licensing and should not be challenged in individual licensing proceedings.

COMMENT 3(Category 4J:

TMI amounts of H2 generation (30-50% clad-water reaction) are appropriate for the interim rule.

The SECY-80-107 series of letters indicate that dry containment systems will have adequate safety margins for this amount of H.

The design study requirement is not justi-2 l

fied.

COMENT 4 (Categories 5 and 6):

The option cf installing internal recom-biners should be in the rule.

COMMENT 5 (Category G):

An exemption should be applied to licensees who have committed to changes which were approved by the NRC staff on the basis of the previously established requirement.

This was the approach in issuing the new Appendix R on fire protection.

COMMENT 6 (Category G):

NUREG-0737 will contain all THI related items that have been approved by the Commission.

The interim rule should be consistent with or merged with this.

COMMENT 7 (Category 8):

The accident sequences to be considered in 50.44a(b)(1) should be specified.

COMMENT 8 (Category 10):

There is no set of radioisotope measurements that can be used to determine the degree of core damage.

Should be re-placed to say "to confirm that significant damage has occured."

COMMENT 9 (Category 11):

In the discussion and text for Leakage Integrity Outside Containment, the phrases "every reasonable effort" and " maximum extent practicable" are used.

These need clarification.

COMMENT 10 (Category 12):

The upper bounds for containment pressure moni-toring given in the rule's discussion are excessive and need justification.

COMMENT 11 (Category 12):

H concentration, containment pre,sure and water 2

level should be available on demand, not " continuously."

9 Enclosure "B-3"

d COMMENT 12 (Category 13):

No good PWR vessel water level indicator is available.

Implementation date should be established after the NRC approves a system.

It is not reasonable to require a readout coverirg the total range of level indications.

COMMENT 13 (Category 14):

Training should be addressed solely in 10 CFR 55.

22.

C-E Power Systems - Nuclear Vendor COMMENT 1 (Category G):

This interim rule should be the basis for licensing approval until a final rule is issued.

However, this rule contains excessive details and specifies redundant and unnecessary implementation dates.

COMMENT 2 (Category 13):

The rule should not preclude new technical advances for meeting the safety criteria.

The detailed criteria for instrumentation for detection of inadequate core cooling is inappropriate at this time.

COMMENT 3 (Category 8):

It is overly prescriptive to designate the level of equilibrium halogens in light of concerns raised in the 8/14/80 letter to Ahearne from Stratton, et al.

COMMENT 4 (Category 4):

The interim rule is not an appropriate mechanism for implementation of design analyses to evaluate measures to mitigate the consequences of large H2 generation.

23.

Florida Power - Utility COMMENT 1 (Category 7):

Venting of the reactor vessel head is not necessary for establishing or maintain natural circulation in a raised loop B&W reactor if a means of venting the hot legs is available.

COMMENT 2 (Category 13):

Based upon the existing instrumentation all necessary operator actions are taken in the event of inadequate core cooling and the additional hardware prescribed by the rule adds nothing.

The NRC stated at the SMUD hearing that incore thermocouples are adequate for detecting inadequate core cooling.

24.

Omaha Public Power District - Utility COMMENT 1 (Category 8):

The rule's source terms for personnel shielding conflict with the commissions 9/5/80 clarification of TMI Action Plan requirements which says that recirculated depressurized reactor coolant may be assumed to contain no noble gases.

The operation of Fort Cal-houn's containment spray recirculation system will release all noble gases in the containment sump into the containment atmosphere.

COMMENT 2 (Category 5 and 6):

The District is willing to consider new regulations concerning revisions of the assumed rate of the H2 genera-tion, but questions the need for a hydrogen recombiner until analyses verify gains in plant safety.

10 Enclosure "B-3"

i s

25.

Duke Power Company - Utility COMENT 1 (Category 13):

Having a reactor vessel level instrument is unneccessary and may be unsafe.

No such instrument is commerically available and thus the implementation date is unrealistic.

26.

Virginia Electric and Power Company - Utility COMENT 1 (Category G):

Recommend integrated rulemaking of a safety goal and methodology. degraded core considerations, establishment of minimum required engineered safety features, siting and emergancy planning.

COMENT 2 (Category G):

Formal codification of existing NUREG-0660 requirements is not recommended.

The rule involves hardware that is not available and thus implementation dates are premature.

COMENT 3 (Category 4):

Previous NRC and industry studies show that safety margins exist.

Regulations are not an appropriate way to under-take research.

The 6 month implementation date for analysis is not appropriate.

COMENT 4 (Category 12):

Sampling and analysis of gaseous effluents may not be possible for all accident conditions due to radiation exposure limi-tations.

COMENT 5 (Category 12):

NRC technical requirements for hydrogen concentration monitoring, particularly sample locations, need to be established before setting an implementation date.

COMENT 6 (Category 12):

NRC technical requirements for containment water level need to be established before an implementation date is set.

COMENT 7 (Category 13):

Implementation date is unrealistic for reactor vessel water level indicator.

NSSS vendors must design and test systems and the NRC must approve these before installation.

31.

Washington Public Power Supply System - Utility COMMENT 1 (Category G):

The Commission should not establish interim requirements which are higly probable to change during the long-term rulemaking process, unless absolutely necessary to protect the public.

1 COM ENT 2 (Category 4):

It is unreasonable to demand separate studies of methods to control H.

The related value/ impact statement estimated 2

"several hundred man years" would be required.

l COMMENT 3 (Category M):

Any interim requirements dealing with the degraded core issue should include the residual ATWS risk, to assist in bringing that issue to an orderly close.

l COMMENT 4 (Category 8):

The proposed design requirement for radioactive l

material release should not be promulgated until the appropriate level has been determined in the long-term rulemaking.

l 11 Enclosure "B-3"

?

COMMENT 5 (Category 10):

The sampling for chlorides in the reactor coolant should be required only for those facilities using saltwater cooling.

COMMENT 6 (Category 11):

Inference that operators or original designs are deficient in maintaining leakage integrity outside containment is incorrect and should be deleted.

COMMENT 7 (Category 11):

The terminology of " integrated leak tests" should be clarified to mean only individual system leak tests which need not be concurrent as long as each test is performed within the normal refueling cycle.

32.

Alabama Power - Utility COMMENT 1 (Category G):

(The wording and content of this letter is nearly identical to that of letter 20, Georgia Power) 32A. Philadelphia Electric Company - Utility COMMENT 1 (Category 7):

Endorse the BWR Owner's position which is that currently installed power operated safety grade relief valves, the HPCI system, and the RCIC system provide the necessary vent capability and that the imposition of a single failure criterion for prevention of inadvertent operation of this vent capability is not applicable to BWRs.

33.

Public Service Indiana - Utility COMMENT 1 (Category 4):

The 75% value of clad-water reaction is excessive.

The NSAC document indicates 45% and the Rogovin Report indicates 31-35% occurred at TMI.

COMMENT 2 (Category G):

Implementation dates should be set on a plant-by plant basis taking scheduled outages into consideration.

COMMENT 3 (Category 13):

Existing devices are sufficient to indicate inadequate care cooling.

Implementation date of new instruments is set before such devices can be developed.

34.

County of Suffolk - Government Agency COMMENT 1 (Category 3):

Recommended that alternatives to inerting for hydrogen control be seriously evaluated for BWRs, prior to adoption of the interim rule.

Inerting decreases the frequency of inspection and could cause a greater frequency of heat removal systems failures.

Improved public health and safety has been demonstrated, and thus the rule's position on inerting is premature and probably unnecessary.

35.

Northeast Utilities - Utility COMMENT 1 (Category G):

Implementation dates do not reflect the input solicited and received by the NRC at their regional meetings the week of 9/23/80 or the relief given in the 10/31/80 Eisenhut letter.

l l

12 Enclosure "B-3"

i

,t COMMENT 2 (Category 3):

There is significant cost due to the amount of nitrogen needed to inert plants.

Inerting decreases safety by restricting surveillance of equipment in containment and causing a signficant risk of asphyxiation.

COMMENT 3 (Category 5):

The existing Haddam Neck Plant purge system was deemed acceptable in meeting NUREG-0578 requirements.

The interim rule should not mandate additional requirements where agreement had been reached.

COMMENT 4 (Category 6):

Requirements for hydrogen recombiner capability would lead to costly plant modifications that may not meet the final NRC H generation criteria.

Schedule is unrealistic.

2 COMMENT 5 (Category 8):

The requirements in NUREG-0588 for the release of radioactive material are different and more realistic than those of the interim rule.

COMMENT 6 (Category 10):

Clarification of the time to "promptly" obtain and analyze samples if needed.

Recommend NUREG-0578 time limits.

COMMENT 7 (Category 10):

Eisenhut's letter of 9/5/80 limits exposure to 5 Rem whole body and 75 Rem to extremeties.

The interim rule's

" equivalent dose to any part of the body" is not well defined.

COMMENT 8 (Category 10):

The rule gives no credit for offsite sample analysis.

NUREG-0578 and the 9/5/80 Eisenhut letter allow offsite analysis within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

COMMENT 9 (Category 10):

Requiring quantification of both total in the reactor coolant is in conflict dissolved gases and dissolved H2 with NUREG-0578 which requires quantification of only one of these.

NUREG-0578 requires boron analysis only for PWR's.

COMMENT 10 (Category 12):

The discussion section on accident monitoring instrumentation references Rev. 2 to Reg. Guide 1.97.

This is inappropriate since the guide is unapproved.

1 COMMENT 11 (Category 12):

The requirement that hydrogen concentration be i

continuously monitored should allow for the fact that H2 sampling is not continuous and that each sample requires real time to evaluate.

COMMENT 12 (Category 13):

The NRC has infermally approved of the use of self powered neutron detectors in conjunction with incore thermocouples to indicate water level within the core.

Hence, additional instrumenta-tion is required to cover the range above the core only.

New level monitoring devices will not be ready in time to meet the implementation date.

36.

Long Island Lighting Company - Utility COMMENT 1 (Category 3):

The requirement to inert BWRs is without justification.

Adverse safety impacts are caused by backfitting designs, adding new equipment, limiting access, and increasing the number of l

13 Enclosure "B-3"

s 4

containment penetrations.

Adverse health impacts include the potential for pockets of trapped inert gas and increased occupational exposure when the need for breathing apparatus increase work time.

Cost of inerting Shoreham is $4 million which is large in relation to the admittedly small benefit gained.

COMMENT 2 (Category 7):

The negative safety impacts of requiring high point vents in the reactor coolant system have not been adequately considered.

37.

Arkansas Power and Light Company - Utility COMMENT 1 (Category G):

It is improbable that an accident of the magnitude addressed in the interim rule would occur during the period preceding the formulation of the final rule.

The implementation of the interim rule is costly and will not obviously improve safety.

COMMENT 2 (Category 4):

The implementation dates for design analyses should be staggered to accommodate the limited number of people who can do this type of work and the limited number of NRC reviewers.

COMMENT 3 (Category 7):

Many plants can adequately cool the core without offsite power or natural circulation.

Vessel head vents are unnecessary for many plants and could be dangerous.

COMMENT 4 (Category 8):

The implementation date for " Protection of Safety Equipment and Areas" is not practical for many plants.

There appears to be no basis for the low dose rate.

COMMENT 5 (Category 10):

The low dose rate for sampling appears to have no basis and is inconsistent with that in 50.44a(b)(1).

Knowledge of reactor coolant chloride levels is irrelevant.

COMMENT 6 (Category 12):

The interim rule's positions on accident monitoring instrumentation appear to involve no determination of what control room decisions would be affected and is inconsistent with efforts to improve the human interface with control room instrumentation.

COMMENT 7 (Category 12):

Only anticipated significant release points should be monitored for radiofodines and radioactive particles.

l COMMENT 8 (Category 13):

Reactor vessel water level has only a very tenuous relation to core cooling (i.e., core temperature).

The core exit thermocouples permit a rough pre-correlation to fuel temperature.

38.

Department of Energy - Government Agency COMMENT 1 (Category 7):

Creating high point vents should be avoided since they can provide additional sources for LOCA.

Other methods than additional primary coolant openings should be found.

COMMENT 2 (Category 4):

There is no basis for clad-water reacti n limits higher than TMI values.

14 Enclosure "B-3"

i I'

s COMMENT 3 (Category 6):

The interim rule's discussion section infers con-struction of venting facilities may be required at each site.

Recommend capability to hook up to a mobile unit instead.

40.

Consumers Power Company - Utility COMMENT 1 (Category 7):

The criteria for the higri point vents should be made consistent with the 9/5/80 Eisenhut letter, i.e., the single failure criterion is unwarranted.

COMMENT 2 (Category 13):

The existing instrumentation is sufficient to indicate the existence of inadequate core cooling for BWRs.

t i

l 15 Enclosure "B-3"

/

GROUPING OF COMMENTS General Comments 1.

Comments 2/1, 6/1, 7/1, 8/1, 16/1, 17/1, 18/1, 20/1, 31/1, 32/1 and 37/1

~

state that the interim rule is inappropriate and should be withdrawn.

Reasons cited were:

a)

NRC needs to first establish its policies regarding hydrogen control, degraded core cooling, plant accident instrumentation, etc., before issuing specific technical specifications.

b)

A formal hearing process is necessary.

Implementation should wait for final rule.

c)

Provisions of the interim rule should be implemented instead through other documents such as reg. guides.

Codification of existing specific requirements is unnecessary.

d)

The cost resulting from implementing the interim rule is not justified since it will not obviously improve safety and the requirements are likely to change during the long term rulemaking.

2.

Comments 1/1, 11/1, 15/1, 21/2, and 22/1 state that the interim rule (subject to recommended changes) should be the basis for licensing and should not be challenged in individual licensing proceedings.

This should be stated in the rule.

l 1

Enclosure "B-4"

i s

3.

Comments 15/2, 21/1 and 26/1 recommend an integrated approach for rulemaking including the development of a safety goal and methodology, degraded core considerations, establishment of minimum required engineering safety features, reactor siting and emergency planning.

4.

Comments 6/2, 15/4 and 21/5 state that the rule should have provisions to prevent nullification of compliance agreements already reached by the Commission with licensees.

5.

In addition to several coments objecting to specific implementation dates, the following general comments were received:

l a)

Comment 10/1 recomends omission of statutory deadlines or add provisions that give flexibility for missed deadlines, b)

Comment 18/2 stated that the dates should agree with those in the Commissioner's 10/28/80 clarification to the TMI Action Plan.

c)

Comment 26/2 stated that since needed hardware is not now available, j

implementation dates are premature.

1 d)

Comment 33/2 stated that implementation dates should be set on a l

plant-by plant basis, taking scheduled outages into consideration.

e)

Comment 5/3 stated that an implementation schedule should be l

required to prevent stalling.

2 Enclosure "B-4" n

o e

f)

Comment 35/1 states that the implementation dates do not reflect the input given at regional meetings or the relief given in the 10/31/80 Eisenhut letter.

6.

Comment 7/2 states that the requirements of the interim rule should be integrated with those of the 9/5/80 Eisenhut letter.

7.

Comment 21/6 states that the rule should be merged with the forthcoming NUREG-0737.

8.

Comment 15/3 states that in some cases the rule is not flexible enough to allow plant-to plant variability.

9.

Comment 5/4 notes that the pre-TMI NRC staff is still in place and states that a new approach to nuclear safety is needed.

Comments 5/2 and 5/3 state that the proposed changes have as much potential to increase risk as to reduce it and that the best way to eliminate degraded core accidents is to shutdown all plants.

3 Enclosure "B-4"

l I

Recommended Additions to Interim Rule 10.

Comment 2/6 states that it would be desirable to have a direct indication of whether or not natural circulation flow has been established.

11.

Comment 5/5 states that workers and operators must have means to determine their personal safety, or else they may panic and leave their posts.

12.

Comment 13/5 states that mixing is essential to many hydrogen control systems and thus all air circulation components must be designed for accidents.

13.

Comment 31/3 states that the residual ATWS risk should be included in any interim requirement on degraded core cooling.

l 4

Enclosure "8-4"

- = _...

o Section 50.44(c)(1) i 6

9 14.

Comment 13/1 state that hydrogen conditions following a real LOCA may be much more severe than from the " postulated LOCA." This phrase should be changed.

i 4

.1 J

i 1

l 5

Enclosure "B-4"

t Section 50.44(c)(3)(i) 17.

Comments 2/2, 6/3, 15/6, 12/1, 15/6, 16/2, 17/2, 18/3, 34/1, 35/2, and 36/1 states that the requirement to inert Mark I and II BWR containments should be removed and/or reevaluated.

The rule's stated justification of decreased risk at a small cost has been challenged by the following:

a)

Inerting increases risk to workers by adding the possibility of asphyxiation and increased occupational exposure when the need for breathing apparatus increases work time.

/

b)

Inerting decreases containment accessability and thus decreases system safety through restrictions to equipment surveillance and maintenance.

c)

Inerting costs are not small.

The cost of increased downtime needs to be considered.

d)

No public domain analysis exists to show that hydrogen combustion is possible or will exceed ultimate containment capacity.

18.

Comment 5/6 states that actual in place test data to measure the effectiveness of inerting systems needs to be generated.

1 l

19.

Comment 13/2 states f4. di.ecLced strength or small containment of the l

l ice condensor variety shoulo wyve inerted atmospheres and notes problems that could be caused by using ignitors.

l l

6 Enclosure "B-4"

i 4

50.44(c)(3)(ii) 20.

Comments 4/1, 4/2, 6/4, 9/1, 11/2, 11/4, 12/2, 15/5, 15/7, 16/3, 17/3, 18/4, 19/1, 22/4, 26/3, 31/2 and 37/2 obi et to the interim rule's requirement that design analyses be made to evaluate measures of hydrogen management.

The following reasons were cited:

a)

Acceptance criteria should first be established by the NRC.

b)

NRC staff positions taken in the SECY-107 series should be used in lieu of new analyses.

c)

Performing analyses in response to the interim rule is premature and wasteful in light of the longer term rulemaking proceedings.

d)

An interim rule is an inappropriate mechanism to implement design analyses or research.

e)

Plant-specific evaluations of H2 generation and containment capacity should be used instead.

Realistic evaluations and PRA methods are recommended.

f)

Evaluation of mitigation features should only be requested if the baseline analysis is unsatisfactory and only for the features proposed by '. 4e applicant.

l 7

Enclosure "B-4"

e s

g)

The implementation date should be extended, staggered (to accommodate the limited amount of analyses and reviewers), or changed to require only schedules of analysis.

~

21.

Comments regarding the interim rule's design limit of 75% fuel cladding-water reaction were a)

Comments 21/3, 33/1, and 38/2 indicate that this limit should be lowered to values estimated for the TMI accident, b)

Comment 13/3 stated that 100% clad water reaction should be used.

c)

Comment 11/3 asked for clarification that studies need not consider hydrogen generation above the 75% limit (the discussion section is unclear).

22.

Comment 5/7 says that analysis is insufficient and that in place tests are necessary.

Long term H2 generation, blocking of cooling lines, and erroneous instrument reading must be considered.

i 1

l l

8 Enclosure "B-4"

6 o

Sections 50.44(c)(3)(iii) and (iv) 23.

Comment 5/9 states that the number of containment penetrations should be minimized, so " dedicated penetrations" should be changed to " penetrations with non-conflicting use."

24.

Comments 15/8, 17/4, 18/5 and 21/4 state that internal recombiners or other alternate means of hydrogen control be allowed as well as external recombiners.

Comment 38/3 suggests that the capability to use external mobile units be allowed.

25.

Comments 3/1, 24/2 and 35/4 state that external recombiners should not be required until final decisions on H generation rates and plant safety 2

are made.

26.

Comments 3/2 and 17/4 request an extention of the implementation dates for recombiners.

Supplies are limited.

27.

Comment 5/8 states that the rule should require that recombiners actually be installed, rather than just the " capability" to install.

28.

Comment 35/3 states that additional requirements should not be added to a purge system deemed acceptable in meeting NUREG-0578 requirements.

9 Enclosure "B-4"

s Section 50.44a(a) 29.

Comments 2/3, 6/5, 6/6, 19/2, 23/1, 32a/1, 36/2, 37/3, 38/1, and 40/1 object to the requirements for high point venting.

Reasons given are:

a)

Many plants currently have sufficient venting schemes.

b)

The negative impacts of requiring high point vents have not been adequately considered.

c)

Reactor vessel head vents are unnecessary and may added risk.

d)

Single failure criteria for high point vents is an unjustified escalation of requirements.

l 30.

Comment 5/10 states that there are circumstances where the migration of l

i gases would be stopped short of a high point.

Also, in place testing is necessary.

31.

Comment 17/5 states that "(2) inadvertent actuation of a vent" should be reworded to preclude the inadvertent opening of a S/RV which provides 1

venting in a BWR.

10 Enclosure "B-4" l

o Section 50.44a(b)(1) i 32.

Comment 3/3 states that the rule should address how or if airbor"1 radioactivity outside containment should be handled.

33.

Comments 6/7 and 24/1 imply that source term levels as given in the 9/5/80 clarification of the TMI Action Plan should be used instead.

Comment 35/5 states that NUREG-0588 requirements should be used.

34.

C>mment 11/5 asks for clarification of the rule's discussion section statement that systems of "significant value" must be assured to operate.

35.

Comment 22/3 says that designation of the level of equilibrium halogens is overly prescriptive.

Comment 12/3 states that iodine releases are likely to be less due to retention in water or steam.

36.

Comment 13/4 states that to be conservative, a 100% release rate of the noble gas, halogen and fission product inventories should be used.

37.

Comment l1/7 requests specification of the accident sequences for which the body dose limit applies.

38.

Comment 31/4 states that implementation should not occur before appropriate radioactive material release levels are determined by the long term rulemaking.

11 Enclosure "B-4"

s 39.

Comment 37/4 states that the implementation date is not practical and there appears to be no basis for the low dose rate.

l l

l i

l l

i l

12 Enclosure "B-4"

J

/

Section 50.44a(c)

40. Comments 6/8, 17/6, 18/7, 35/7, 35/8 and 35/9 imply that this section should adopt positions set in NUREG-0578 and Eisenhut's 9/5/80 clarification letter of the TMI Actica Plan.

Specifically:

a)

Boron analyses should be required only for PWRs.

b)

Items 50.44a(c)(iii) and (v) should not apply to BWRs.

c)

Quantification should be for total dissolved gases or dissolved hydrogen, not both, d)

Exposure limit of 5 Rem to body and 75 Rem to extremities should be adopted.

e)

Credit for offsite sampling should be given.

11.

Comments 15/9, 18/6 and 21/8 stated that there exists no set of radioisotope measurements that can determine the degree of core damage.

Rephrase this to say "to confirm that significant damage has occurred."

42.

Comments 16/4, 31/5 and 37/5 state that sampling of chorides is unnecessary.

Should be required only for facilities using saltwater cooling.

13 Enclosure "B-4"

4 43.

Coment 37/5 states that the low dose rate for sampling has no basis and is inconsistent with 50.44a(b)(1).

l 45.

Comment 10/2 says the rule should be revised to reflect EPRI's recent report on post-accident sampling.

j 46.

Comment 35/6 states that time limits for sampling should be given, not i

just "promptly."

l l

l l

I I

i l

l

)

14 Enclosure "B-4"

O 4

Section 50.44a(d) 47.

Comment 21/9 states that clarification in the rule's text and discussion of Leakage Integrity Outside Containment is needed for the phrases "every reasonable effort" and " maximum extent practicable."

48.

Comment 31/6 states that inference that operators or original designs are deficient in maintaining leakage integrity outside containment is incorrect and should be deleted.

49.

Comment 31/7,tates that " integrated leak tests" should be clarified to mean only individual system leak tests which need not be concurrent as long as each test is performed within the normal refueling cycle.

15 Enclosure "B-4"

s 4

Section 50.44a(e) 50.

Comments 3/4, 21/11 and 35/11 object to the requirement of a " continuous" indication of containment H because:

2 a)

In BWRs the combustible gas control system is already in operation.

b)

H should be monitored only by batches and on demand.

2 l

51.

Comments 3/5, 10/3, 16/5, 17/7, 26/5, and 26/6, state that implementation dates for Accident Monitoring should be removad or extended because:

i a)

Not sufficient time to install radiciodine, particulate, and noble gas effluent monitors.

b)

Revision 2 to Reg. Guide 1.97 should be issued first.

l c)

Containment water level monitors implementation should be changed to "first scheduled outage." Takes 3 to 4 weeks to drain a Mark I

torus, d)

Technical requirements for containment water level need to be i

established first.

l monitoring, particularly sample e)

Technical requirements for H2 locations, need to be established first.

16 Enclosure "B-4"

4 i

\\

52.

Comment 6/9 states that requirement for continuous noble gas effluent monitoring is inconsistent with Eisenhut's 9/5/80 clarification letter.

53.

Comments 19/3 and 37/7 state that quantifying concentrations of radioiodines and radioactive particles in airborne effluents at each anticipated release point is impractical (as stated in Rev. 2 of R.G. 1.97).

Only significant points should be considered.

54.

Comment 21/10 states that the upper bounds for containment pressure monitoring (in the discussion section) are excessive and need justification.

55.

Comment 26/4 states that sampling and analysis may not be possible for all accident conditions, due to exposure limits.

56.

Comment 35/1G ',tates that reference to the unissued Rev. 2 of Reg.

Guide 1.97 in the discussion section is inappropriate.

57.

Comment 37/6 states that the rule's position on accident monitoring appears to have no determination of what control room decisions would be affected and is inconsistent with efforts to improve human interface with l

control room instrumentation.

17 Enclosure "B-4"

s Section 50.44a(f) 58.

Comments 15/10 and 17/8 imply the training requirements in this section should appear in 10CFR55 only.

59.

Comments 15/11, 16/6, 17/9, 19/4, 21/12, 22/2, 23/2, 25/1, 26/7, 33/3, and 35/12 objected to the requirement for reactor vessel water level instrumentation and/or the implementation date given.

Reasons are:

a)

No gcod water level indicators are commercially available and approved by the NRC.

b)

It is not necessary to require readouts for the total range of uncovering and thus current instrumentation (e.g., incore thermocouples) can be used.

c)

Such instrumentation may increase risk.

l 60.

Comment 2/5 states that reactor water level indication would be valuable during a small LOCA, but that the implementation date is optimistic.

61.

Comment 37/8 states that the reactor vessel water level has only a very tenuous relation to core cooling (i.e., core temperature).

The core exit thermocouples permit a rough pre-correlation to fuel temperature.

18 Enclosure "B-4"

/

62.

Comment 2/4 states that the subcooling meter is of marginal value:

a) during a small break LOCA the reactor system is in a saturated state.

b) the plant operators don't use it since it is not used during normal operations.

63.

Comment 40/2 states that the existing instrumentation is sufficient to indicate the existence of inadequate core cooling for BWRs.

19 Enclosure "B-4"

4 Section 50.44a(g) 64.

Comments 15/10, 17/8, and 21/13 state that training should be addressed solely in 10CFR55.

65.

Comment 17/8 states that post-TMI emergency operator guidelines should be fully implemented in plant specific procedures and assessed before the rule's implementation is set.

.I l

l l

l l

l l

20 Enclosure "B-4"

~

CROSS INDEX OF COMMENTS Organization Letter No.

1 2

3 4

5 6

7 8

9 10 11 12 13 Edison Electric Institute 1.

2 L. P. Leach 2.

1 17 29 62 60 10 Philadelphia Electric Co.

3.

25 26 32 50 51 Conner &~ Moore 4.

20 20 M. I. Lewis 5.

9 9

5 9

11 15,18 22 27 23 30 Bechtel Power Corp.

6.

1 4

17 16 29 29 33 40 52 TWR owner 5 Gr.sup 7.

1- ~~6-~

L. J. Ybarrendo 8.

1 l

J. F. Kowa.iski 9.

20 Fluor Power Services 10.

5 45 51 Offshore Power Systems 11.

2 20 21 20 34

__ Commonwealth Edison 12.

17 20 35 Ethos Research Corp.

13.

1r 19 -~21- ~36 - F Atomic Industrial Forum 15.

2 3

8 4

20 17 20 24 41 58,63 59 Yankee Atomic Electric Co.

16.

1 17 20 42 51 59 General Electric 17.

1 17 20 24,26 31 40 51 58,64 59 Stone & Webster 18.

1 5

17 20 24 41 40 Consumers Power Co.

19.

20 29 jl3_

59 Georgia Power 20.

1 Westinghouse Electric Corp.

21.

3 2

21 24 4

7 37 41 47 54 50 59 63

_C-E_ Power Systems 22.

2 59 35 20 _

Florida Power 23.

29 59 Omaha Public Power District 24.

33 25 Duke Power Co.

25.

59 Virginia Electric & Power Co.

26.

3 5

20 55 51 51 59 Washington Public Power Supply System 31, 1

20 13 38 42 48_

49_.

Alabama Power 32.

1 9 Philadelphia Electric Power 32a.

29 EL Public Service Indiana 33.

21 5

59 8 Country of Surtolk 34.

17

~~

E Northeast Utilities 35.

5 17 28 25 33 46 40 40 40 56 50 59

• , _Long Island Lighting Co.

36.

17 29 a5 Arkansas Power & Light Co.

37.

I 20 29 39 42,43 57 53 61 0,

D.O.E.

38.

29 21 24 Consumers Power Co.

40.

29 63 Enamples:

(1) The 2nd comment in letter 38 (denoted 38/2) is addressed in the general comment 21, as is the 3rd comment in letter 13 (denoted 13/3) and others.

(2) The 10th comment in letter 15 (denoted 15/10) is addressed in both general comment 58 and 63.

s

t ENCLOSURE "C" i

. _... _ - -.,. _,.,.... - _. ~.

.s.

[7590-01]

3 NUCLEAR REGULATORY COMISSION 10 CFR Part 50 Interim Requirements Related to Hydrogen Control and Certain Degraded Core Considerations for Nuclear Power Plants AGENCY:

Nuclear Regulatory Commission.

ACTION:

Final Rule.

SUMMARY

The Nuclear Regulatory Commission is amending its regulations to improve hydrogen control capability during and following an accident in light-water reactor facilities and to provide specfic design and other requirements to mitigate the consequences of accidents resulting in a degraded reactor core.

The amendments are applicable to boiling and pressurized light-water nuclear power reactors.

EFFECTIVE DATE:

[30 days following publication in the Federal Register]

FOR FLRTHER INFORMATION CONTACT:

Morton R. Fleishman, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, telephone 301-443-5921.

SUPPLEMENTARY INFORMATION:

On October 2, 1980, the Nuclear Regulatory Commission published in the Federal Register (45 FR 65466) a notice of l

proposed rulemaking inviting written comments or suggestions on the pro-posed rule by November 3, 1980.

The notice concerned proposed amendments to 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facil-1 ities," to improve hydrogen management in light-water reactor facilities and to provide specific design and other requirements to mitigate the consequences of accidents resulting in a degraded reactor core.

1 Enclosure "C"

[7590-01]

r Thirty-five persons submitted comments regarding the proposed amend-ments.

Though the comment period was scheduled to expire on November 3, 1980, comments received subsequent to that date have been considered, with the latest comment letter being dated February 9, 1981.

The comments are part of the public record and may be examined and copied in the Commis-sion's Public Document Room at 1717 H Street NW., Washington, D.C.

A summary of the comments along with a comment analysis and a value/ impact assessment are also available for inspection and copying in the Public Document Room.

The comments that were submitted have been carefully reviewed and evaluated during preparation of this notice of final rulemaking.

The final rule contains revisions to the proposed rule that reflect these comments.

The commenters were about equally divided between those in favor of and those opposed to publishing the interim amendments.

Whether or not the commenter favored publishing a final rule, additional detailed comments were generally provided on specific aspects of the pro-posed amendments.

The NRC's Office of Nuclear Reactor Regulation sent a letter on September 5, 1980 to all nuclear power plant licensees, applicants and construction permit holders providing a " Preliminary Clarification of the TMI Action Plan Requirements." This was followed by a series of four regional meetings, noticed by publication in the Federal Register (45 FR 60508) and held during the week of September 22, 1980, in order l

to provide a more detailed explanation of the requirements and to obtain industry comments.

Based on the discussions at the meetings and other comments received, the NRC revised the requirements and notified the applicants, licensees and construction permit holders to this effect by 2

Enclosure "C"

s

[7590-01]

s a letter dated October 31, 1980.

The letter and revised requirements are included in NUREG-0737, " Clarification of TMI Action Plan Requirements."1 The Commission has also been considering the ability of pressurized light-water reactor facilities with ice condenser type containments and of boiling light-water reactor facilities with Mark III type containments to withstand a TMI-2 type of accident with the concomitant generation of large amounts of hydrogen.

The final rule contains revisions to the proposed rule that reflect all of the applicable comments including those (a) given in response to the notice of proposed rulemaking, (D) 9.'erated during the regional meet-6 ings and in response to the clarification letters of September 5, 1980 and October 31, 1980, and (c) the Commission's considerations with respect to Mark III BWRs and ice condenser PWRs.

Before discussing the comments and the specific revisions resulting from the comments, it should be noted that, while 5 50.44 has applied only to nuclear power reactors with zircaloy fuel cladding, the new amend-ments in the interim rule are not as limited.

The only new amendment limited in applicability to reactors with zircaloy fuel cladding is 6 50.44(c)(3)(v).

All the other amendments apply to reactors with either stainless steel or zircaloy fuel cladding.

Several commenters have expressed the concern that the various rule-makings currently being pursued by NRC should be integrated, i.e., safety goal, degraded core considerations, minimum engineered safety features, 1

1 1 Copies of this report may be obtained from GPO Sales Program, Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555.

3 Enclosure "C"

. ~.

[7590-01]

siting and emergency planning.

The NRC has also had this concern and on October 15, 1980 the Exe utive Director for Operations established a Degraded Cooling Steering Group to coordinate degraded cooling and re-lated rules.

This group has completed its work and prepared a plan to ensure future integration of these activities.

Numerous commenters have questioned many of the implementation dates,.

specified in the rule, indicating that they cannot be met and citing a variety of reasons, such as procurement lead time, need for the design studies, availability of acceptable equipment, etc.

As indicated previ-ously, during the development of NUREG-0737 on clarification of TMI action plan requirements, many discussions were held between the NRC staff and industry personnel.

The dates in the final rule have been revised to be consistent with those contained in NUREG-0737.

It is believed that the revised dates now provide a reasonable compromise considering both public health and safety and techr.ical feasibility.

Several commenters voiced the opinion that the interim rule should be delayed until NRC sets policy regarding hydrogen control, degraded core cooling, etc., such as will be done during the long term rulemaking.

As was indicated previously in the background information associated with the proposed interim rule, the Three Mile Island, Unit 2 (TMI-2) " accident revealed design and operational limitations that existed relative to miti-gating the consequences of the accident and determining the status of the facility during and following the accident" (45 FR at 65466).

The specific items being covered by this interim rule "have been determined to be of such safety significance that they should be codified by regula-tion in order to provide assurance that the public health and safety will i

be adequately protected" (45 FR at 65466).

That is, they are believed l

l 4

Enclosure "C"

s

[7590-01]

s to be so important that delaying them to the long term rulemaking on degraded core cooling (45 FR 65474) is unwarranted.

Furthermore, the I

Commission intends this interim rule to provide the needed basis for regulatory actions that cover licensing and continued operation of nuclear power plants.

This position has also been recommended by a number of commenters.

By promulgating the interim rule, the Commission intends to remove these items from litigation in individual proceedings and thus eliminate the delay and duplication of effort that occur when the same safety issues are considered in many different proceedings.

Some commenters suggested that in view of the great amount of detail involved, it would be better to promulgate the requirements by using a less formal approach such as providing guidelines set out in regulatory guides.

However, in view of the seriousness of the safety issues involved and the need to implement changes quickly, it was felt to be more appro-priate to affect these changes with a regulation.

The regulation would, of course, be enforceable while guidelines would not.

Additional guidance regarding detailed implementation of specific aspects of the interim rule is provided in NUREG-0737, mentioned previously.

Furthermore, the NRC staff intends to utilize the clarifications and guidance in NUREG-0737 in evaluating applicant and licensee implementation actions.

It has been stated that the interim rule is not flexible enough to allow for plant-to plant variability.

Although the rule is somewhat detailed, the Commission believes it is general enough to account for plant-to plant variability.

The rule will be complemented by NUREG-0738 and Regulatory Guide 1.97 which will provide the additional guidance to account for plant-to plant variability.

5 Enclosure "C"

[7590-01]

Several commenters have indicated that the interim rule covers, to a large degree, previous, less formal agreements made between the NRC and licensees as well as earlier letters to licensees, and that the rule should be made consistent with these agreements.

They suggested that the rule should also have provisions to prevent nullification of previous compliance agreements.

The Commission basically agrees with these sugges-tions and the final interim rule has been made consistent with NUREG-0737 which clarifies the THI Action Plan requirements.

NUREG-0737 includes modifications from earlier letters and NUREG reports in order to account for compliance agreements previously reached.

INERTING OF MARK I & II BWRs [6 50.44(c)(3)(i)]

Some commenters, particularly those associated with Mark I boiling water reactors (BWRs), questioned the advisability of requiring inerting of containments and suggested that other hydrogen control options be per-mitted.

This issue has been extensively reviewed and discussed among the Commission, NRC staff and industry participants.

Numerous reports I

and letters have been written and many meetings held in order to thoroughly air the issue.

Considering the information previously developed, the Commission continues to believe that it would be prudent, l

pending completion of the long term rulemaking on degraded core cooling, to require that all Mark I and II BWR containments be provided with an inerted atmosphere during normal operations.

The original wording of the rule has been revised to delay the imple-mentation date by which the inerting systems must be installed to allow more time for compliance since the date of publication of the rule has l

6 Enclosure "C"

[7590-01]

t been delayed. The rule has also been changed to clarify that the subparagraph applies to Mark I and II BWRs.

HYDROGEN CONTROL FOR MARK III BWRs AND ICE CONDENSER PWRs [l50.44(c)(3)(ii)]

An entirely new paragraph has been added to require that Mark III BWRs and pressurized water reactor (PWR) facilities with ice condenser type containments install hydrogen control systems to accommodate an amount of hydrogen equivalent to that generated from the reaction of 75%

of the fuel cladding with water without loss of containment integrity.

This is a new requirement added as a result of safety issues raised during licensing reviews of new ice condenser and Mark III plants.

In these reviews, it has become clear that additional protection is required to provide assurance that large amounts of hydrogen can be safely accom-modated by these plants. The particular type of hydrogen control system that is selected is left to the discretion of the applicant or licensee; l

however, it must be found to be acceptable by the NRC based upon suitable programs of experiment and analysis. The selection should be supported by comparative analyses of alternative systems to show their relative advantages and disadvantages.

These comparisons are to be submitted as part of the analyses required under B 50.44(c)(3)(iv). At present, a distributed igniter system has been found to be acceptable for the Sequoyah plant with an ice condenser containment, but only as an interim solution while the hydrogen control matter is studied further. A post-accident inerting system has also been discussed for the ice condenser and Mark III containments. Whatever systems are finally proposed and approved for the long term, large amounts of hydrogen must be safely accomodated, and operation of the system, either intentionally or inadvertently, must not further aggravate the course of an accident or 7

Enclosure "C"

[7590-01]

endanger the plant during normal operations. The amount of hydrogen to be assumed in the design of the hydrogen control system is that amount generated by assuming that 75% of the fuel cladding reacts with water. The nominal value of approximately 75% was chosen as a reason-able compromise between the estimated 25 to 50% that occurred at TMI-2 and the 100% that is theoretically possible and which has been proposed by one commenter.

Furthermore, events with metal-water reactions in excess of 75% have a very low probability of termination before core melt. This 75% value also appears to be reasonable because it is suffi-ciently greater than the fuel cladding-water reaction that occurred at TMI-2 to provide a conservative estimate for the cladding reaction that may occur duri ;g a TMI type degraded core accident.

It is expected that the 75% value will permit plants that are either completed or are well along in the construction stage to have a hydrogen control system added without the need for major modifications to their containment structures.

Owners of Mark III BWRs now under construction have been surveyed by the NRC staff to determine the effect on their plant designs of the requirement that they do not exceed ASME Service Level A Limits or the Service Load Category during inadvertent full inerting of a post-accident inerting system. This survey was conducted because a post-accident inerting system rather than a distributed ignition system may be the preferred approach for the Mark III containments. Based on their responses, the Commission has concluded that there would be no significant impact in specifying these requirements for inadvertent full inerting.

Modest deviations from these :riteria will be permitted if good cause is shown. A comparable survey was not conducted for ice condenser plants because tne distributed ignition system apparently is the approach preferred by the owners of these plants.

8 Enclosure "C"

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[7590-01]

There are ongoing programs of research in a number of areas of hydrogen generation, release, burning, and control. These include the analysis of accident sequences, the chronology of hydrogen and steam injection (from the primary system into containment), the analysis of operations to recover coolability, and an assessment of equipment sur-vivability. These studies are expected to reveal the advantages and disadvantages of various hydrogen control systems including those that involve deliberate burning of '.he hydrogen within containment. Based on the state of technology as of April 1981, the Commission believes that control methods that do not involve burning (e.g., inerting) provide protection for a wider spectrum of accidents than do those that involve burning.

Also, as a result of the review of the deliberate ignition systems installed at Sequoyah and McGuire, the staff has identified issues which need to be investigated further. A spectrum of degraded core accident i

scenarios including those which may lead to inadvertent suppression of combustion in the lower compartment, and several hydrogen combustion phenomena, are continuing to be reviewed.

In addition, there is incomplete verification of analytical models and equipment survivability. These issues are being addressed in ongoing research by NRC and the nuclear industry. The Commission concludes that the issues are sufficiently resolved to warrant interim approval of deliberate ignition systems for 1

ice condenser plants. However, the Commission has required in individual licensing proceedings and in the section of this rule on analyses that studies of alternative hydrogen management systems be performed prior to l

l the long-term approval of any particular method.

~

9 Enclosure "C"

[7590-01]

At present, PWRs with dry containments and subatmospheric contain-ments are not required to be provided with additional hydrogen control systems because previous experience and analytical studies, performed by the NRC staff and industry groups, indicate that these containment structures generally can safely handle amounts of hydrogen up to that associated with approximately a 75% fuel cladding-water reaction.

How-ever, they are still required to meet other applicable hydrogen control requirements, including the survivability requirements of 5 50.44(c)(3)

(iii) and the confirmatory analysis requirements of 5 50.44 (c)(3)(iv).

These requirements deal with the need to ensure that certain safety systems will survive degraded core accidents and the need to perform plant specific hydrogen control analyses.

SURVIVABILITY OF CERTAIN SAFETY SYSTEMS DURING AND FOLLOWING A HYDROGEN BURN [6 50.44(c)(3)(iii)]

An entirely new paragraph on safety system survivability has been added.

It applies to all BWRs and PWRs that do not have an inerted con-tainment atmosphere for hydrogen control.

That is, all plants for which there exists the possibility that substantial amounts of hydrogen can be burned in the containment will be covered by the new requirement.

These plants must be provided with certain safety systems that can survive the environmental conditions associated with hydrogen burning.

Thus, for example, if a distributed igniter system is selected for controlling large amounts of hydrogen, the applicants or licensees must demonstrate that the specified safety systems can survive and continue to perform their functions during and following hydrogen burning.

If no new hydro-gen control system is required, as is presently the case for PWRs with large dry containments, these applicants and licensees would still have to demonstrate that the specified safety systems can continue to perform 10 Encic3Gre "C"

t

[7590-01]

s 1

their functions during and following hydrogen burning. This demonstra-tion of survivability for certain identified essential systems is needed because the environmental service temperatures associated with hydrogen burning can be more severe than the conditions for which the equipment l

has been qualified.

ANALYSES [5 50.44(c)(3)(tv)]

The proposed interir rule required that for all PWR and BWR plants, except the Mark I and II BWRs, design analyses must be performed for new hydrogen control measures. Many commenters indicated that the descrip-tion of the destgr. analyses was not precise enough to elicit the desired response.

Furthermore, several commenters have suggested that it is inappropriate to have a regulation requiring hydrogen control design studies in view of the fact that unambiguous event descriptions and acceptance criteria are not supplied. The Commission agrees with these comments in part. As a result, the rule has been modified to clarify the types of analyses required. They can be grouped into four classes, depending upon containment design, as follows:

1.

BWRs with Mark I and II containments are required to be inerted by this rule. There are no further analyses required of these plants.

2.

Analyses are required for PWRs with ice condenser containments to i

fully demonstrate by January 31, 1982, or by the date of issuance of a license authorizing operation at full power, whichever is sooner, that an adequate hydrogen control system is installed and will perform its intended function in a manner that provides adequate safety margins. The analyses are required to include the installed system and alternative systems.

11 Enclosure "C"

[7590-01]

3.

Analyses are required for BWRs with Mark III containments to fully demonstrate by January 31, 1982, or by the date of issuance of a license authorizing operation at full power, whichever is sooner, that an adequate hydrogen control system is installed and will perform its intended function in a manner that provides adequate safety margins. The analyses are required to include the installed system and alternative systems.

4.

Owners of all other containments are required to perform and submit containment integrity analyses and sensitivity studies within six months of the effective date of this rule or the date of docketing of the application for the operating license, whichever is later.

This information is needed to demonstrate with reasonable assurance that safe shutdown and containment integrity will be maintained with adequate safety margins for degraded co,e accidents. These acci-dents will be assumed to produce hydrogen releases to the contain-ment resulting from the reaction of up to and including 75% of the fuel cladding with water for a range of time periods up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The containment integrity analysis will include consideration of local hydrogen detonations.

The analyses required by this section serve two purposes.

First, they support continued reliance on the interim requirements of this rule.

Second, the results will be considered in the longer term rulemaking on degraded cores.

l 12 Enclosure "C" 4

- _. ~,,. -,

,, -. - +,. -,, -

t o

DEDICATED HYDROGEN CONTROL PENETRATIONS [6 50.44(c)(3)(v)]

~

This section was essentially unchanged from that originally proposed except for minor changes of an editorial nature or to clarify tr.1 intended re sning of the regulation.

HYOROGEN RECOMBINER CAPABILITY [9 50.44(c)(3)(vi)]

Several commenters have recommended that S 50.44(c)(3)(vi) be modi-fied to allow the use of alternate means of hydrogen control, such as internal recombiners, rather than restrict the rule to external recom-

-biners.

The proposed rule was not intended to preclude this alternative.

In ib:1, if internal recombiners were present before or will be installed in the future, this section of the rule would not apply since purge /

repressurization systems would not be the primary means for combustible gas control.

This section of the rule only applies to facilities that rely upca purge /repressurization systems.

It should also be noted that this section of the rule does not require that external recombiners be installed where no internal or external recombiners already exist; rather, it requires only the capability for installation.

To avoid confusion in the future, the rule has been clarified to indicate that internal recom-biners are an acceptable alternative to the installation of external recombiner capability.

13 Enclosure "C"

~. ~ -

---we

9 HIGH POINT VENTS IN REACTOR COOLANT SYSTEM [9 50.44a(a)]

A number of commenters have remarked that there is no justification for applying the single failure criterion to the design of the nigh point vents.

Furthermore, it has been suggested that the negative aspect of the high point vents have not been adequately considered and that, in fact, the vents may increase the risk to the public.

The interim rule has been revised to relax the implementation date in response to comments received at the regional meetings in September 1980.

In response to the above comments, the single failure criterion requirement has been deleted except that a single failure within the power and centrol aspects of the reactor coolant vent system should not prevent isolation of the entire vent system when required.

Also a sentence has been added to require that the use of the high point vents not " aggravate the challenge to the containment or the course of the accident."

.~.....

14 Enclosure "C"

[7590-01]

o PROTECTION OF SAFETY EQUIFMENT AND AREAS WHICH MAY BE USED DURING AND FOLLOWING AN ACCIDENT [6 50.44a(b)(1)]

Commenters have suggested that the source term prescribed in this section of the interim rule should be consistent with that given in NUREG-0737.

In particular the source term for recirculated, depres-surized coolant water should not be assumed to contain noble gases.

The rule has been modified to this effect.

Additionally, commenters have suggested that the amount of equilibrium halogens to be included in the source term is overly prescriptive in view of the events which occurred at TMI-2 and some recent experimental data.

The NRC and the Advisory Committee on Reactor Safeguards (ACRS) have given this issue a great deal of attention.

A comprehensive draft report has been prepared by the NRC and its contractors, NUREG-0772, " Technical Bases for Estimating Fission Product Behavior During LWR Accidents,"1 and has been reviewed by the ACRS.

The ACRS indicated that the draft report "provides a good up-to-date summary of knowledge on potential fission product releases under a range of postulated accidents." However, the ACRS indicated that the " report does not contain data or information that would justify changing current regulatory criteria at' this time."

In view of the staff and ACRS conclusions, the Commission intends to make no further change in the radioactive material source term at this time.

IN-PLANT IODINE INSTRUMENTATION [9 50.44a(b)(2)]

i There were no specific comments relative to this section of the rule.

However, as a result of earlier Commission actions on the TMI Action Plan l

l 15 Enclosure "C" I

m

[7590-01]

and NUREG-0737, operating facilities appear to have complied with this item as of January 1,1981.

Hence, the implementation date was deleted and the requirement will be effective on the effective date of the rule.

SAMPLING OURING AND FOLLOWING AN ACCIDENT [6 50.44a(c)]

Commenters have questioned the dose limit of 3 rem for sr.mpling given in 6 50.44a(c)(2) and indicate that it should be increased to 5 rem to be internally consistent with the dose limit in 6 50.44a(b)(1) and with General Design Criteria 19.

As a result of this comment 6 50.44a(c)(2) has been revised to increase the dose limit to 5 rem.

Also, the rule has been revised to clarify that the 5 rem is a design parameter and not an operational limitation.

Operational limitations on the dose permitted to an individual are covered in 10 CFR Part 20, " Standards for Protection Against Radiation."

Several commenters questioned whether radioisotope measurements could

" determine the degree of" core damage and suggested that the rule be revised to simply " indicate the extent of" core damage.

This change has been made.

A number of comments were made suggesting that the rule be made con-sistent with similar items covered by NUREG-0737.

It should be reiterated here that NUREG-0737 will be utilized by the staff in evaluating responses to this rule.

However, some changes in the rule have been made for consist-ency with NUREG-0737.

Thus, 6 50.44a(c)(3)(iii) has been changed to indicate that either total dissolved gases or dissolved hydrogen gas in the reactor coolant must be quantified, not both.

Subparagraph 6 50.44a(c)(3)(v) has been revised to indicate that " Chloride analyses may be performed 16 Enclosure "C"

[7590-01]

offsite and are not required to be done promptly." Furthermore, the rule has been changed to indicate that if in-line monitoring is chosen for j

the sampling, the facility must be provided with the additional capability l

to obtain and analyze grab samples.

1 LEAKAGE INTEGRITY OUTSIDE CONTAINMENT [$ 50.44a(d)]

One commenter felt that the use of the phrase " maximum extent prac-ticable" should be _larified.

Specific values for leakage rate reductions cannot be supplied because they are highly dependent on the particular system being considered.

The use of the term " maximum extent practicable" is intended to mean that a knowledgeable individual would take all rea-sonable measures to reduce leakage based on the available technology, the cost, and the safety and operational benefits to be gained.

A change was made, in response to one comment, to clarify that the integrated leak tests to be performed should be done on the individual systems for which the leakage reduction program is intended, and to con-sider the integrated leakage over a given time interval.

The rule does not require that these tests be done concurrently as long as they are done at intervals that are no longer than that of a normal refueling cycle.

ACCIDENT MONITORING INSTRUMENTATION [6 50.44a(e)]

Numerot.s commenters have recommended that the implementation dates for installation of the accident monitoring instrumentation be delayed, for a variety of reasons.

As was recommended, all of these implementa-tion dates have been delayed based on discussions held during the 17 Enclosure "C"

[7590-01]

September 1980 regional meetings and the comments received.

The revised dates are consistent with those in NUREG-0737.

Several commenters also questioned reference to the use of Regulatory Guide 1.97 since it was not issued at the time the proposed rule was pub-lished.

Regulatory Guide 1.97, Revision 2, "Instrumention for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident,"1 was issued in December 1980.

This regulatory guide and NUREG-0737 should be consulted for guidance as to how the NRC staff intends to evaluate compliance with the interim rule.

Several other changes were also made in the interest of consistency with NUREG-0737 as well as in response to comments.

These involved clari-fying that noble gas concentrations need only be monitored at the "poten-tial accident gaseous effluent release paths." Furthermore, the radio-iodine and radioactive particulates need only be nonitored at " anticipated significant" release points.

DETECTION OF INADEQUATE CORE COOLING [6 50.44a(f)]

Several commenters objected to what they perceived to be a require-ment for a reactor vessel water level indicator and suggested that reli-able indicators were not commercially available and approved by the NRC.

The Commission does not agree and requires augmented water level indica-tion in PWRs by the water level indicators and in BWRs by the incore thermocouples.

The rule states that "the instrumentation system must supply to the control room a recorded, unambiguous, easy-to-interpret, indication of inadequate core cooling," be provided.

This system could include a synthesis of computers, cathode ray tube (CRT) displays and 18 Enclosure "C"

t

[7590-01]

various detectors, etc., and must be capable of meeting these require-ments.

When they are available, reliable reactor vessel water level indicators should be included in the system.

Additional guidance for implementation of this requirement is given in NUREG-0737 and Regulatory Guide 1.97, Revision 2.

TRAINING AND HUMAN ENGINEERING TO MITIGATE DEGRADED CORE ACCIDENTS

[$ 50.44a(g)]

Some commenters remarked that all training requirements should be covered by 10 CFR Part 55, " Operator's Licenses," and suggest that this part of the interim rule, as well as the training requirement in 9 50.44a(f) for inadequate core cooling, be shifted to 10 CFR Part 55.

This position has been previously considered during the development of the interim rule.

Part 55 is concerned primarily with the procedures and criteria for the issuance of licenses to opeators.

This includer licenses for operation of any production or utilization facility as defined in 10 CFR Part 50.

The specific training being required by the interim rule applies only to boiling and pressurized light-water nuclear power reactor licensees.

Furthermore, S 50.44a is primarily concerned with mitigating the conse-quences of degraded core accidents.

Thus, it was felt that, since greater visibi) Sty would be provided for the specific training requirements, and since the regulation was limited in scope to nuclear power plants, it was more appropriate to keep these specific training requirements in l

l 6 50.44a.

The implementation date of April 1,1981 for the training require-ment was deleted because by the time the rule is made effective, all of l

19 Enclosure "C"

9

[7590-01]

e the plants to which it is applicable are expected to be in compliance as a result of the recommendations made in NUREG-0737.

In aadition, the heading and the rule were changed to indicate that " human engineering" data must be utilized to support the development of the training program, emergency procedures and control roem hardware.

The purpose of this addi-tion is to enhance the operator's understanding of the problem and the associated actions required to mitigate the consequen:es of a degraded core accident.

Accordingly, notice is hereby given that, pursuant to the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and section 553 of title 5 of the United States Code, the following amendments to 10 CFR Part 50 are published as a document sub-ject to codification.

PART 50--00MESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES 1.

Section 50.44 of 10 CFR Part 50 is amended by revising para-graph (c) to read as follows:

950.44 Standards for combustible gas control system in light water cooled l

power reactors.

1 (c)(1) For each boiling or pressurized light-water nuclear power l

reactor fueled with oxide pellets within cylindrical zircaloy cladding, it shall be shown that during the time period following a postulated LOCA l

but prior to effective operation of the combustible gas control system, either:

(i) An ur. controlled hydrogen-oxygen recombination would not take l

20 Enclosure "C" y

r-

,,,..---,--n

i t

[7590-01]

place in the containment; or (ii) the plant could withstand the conse-quences of uncontrolled hydrogen-oxygen recombination without loss of safety function.

l (2)

If neither of these conditions can be shown, the containment shall be provided with an inerted atmosphere or an oxygen deficient condi-tion in order to provide protection against hydrogen burning and explo-sions during this time period.

(3) However:

(i)

[As-soon-as practicable-but-not-iater-than-dane-30;-1981] Effec-l l

tive January 1, 1982,* or sooner if practicable, an inerted atmosphere shall be provided for each boiling light-water nuclear power reactor with a Mark I or Mark II type containment; [ facility-for-which-the-application l

for-a-centainment permit-was-decketed-between-March-15 -1964-and-daiy-1; 7

1972;] and (ii) Effective January 31, _1982,or issuance of a license authorizing operation at full power, whichever is sooner, each boiling light-water

(

nuclear power reactor with a Mark III type containment and each pressur-ized light-water nuclear power reactor with an ice condenser type contain-ment shall be provided with an acceptable hydrogen control system justi-fied by suitable programs of experiment and analysis that is capable of handling an amount of hydrogen equivalent to that generated from the reac-tion of 75% of the fuel cladding with water without loss of containment integrity (i.e., steel containments must meet the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Sub-subarticle NE-3220, Service Level C Limits, except that evaluation of

• Comparative text.

Additions shown by underline, deletions by bracket and crossout.

21 Enclosure "C"

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[7590-01]

e instability is not required, considering pressure and dead load alone.

Concrete containments must meet the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Division 2, Subsubarticle CC-3720, Factored Load Category, considering pressure and dead load alone.

These subsubarticles have been approved for incorporation by reference by the Director of the Federal Register on

, 1981.

A notice of any changes made to the material incorporated by reference will be published in the Federal Register.

Copies of the ASME Boiler and Pressure Vessel Code may be purchased from the American Society of Mechan-ical Engineers, United Engineering Center, 345 East 47th Street, New York, N.Y. 10017.

It is also available for inspection at the Nuclear Regulatory Commission's Public Document Room, 1717 H. Street NW., Washington, D.C.)

If the hydrogen control system relies on post-accident inerting, the con-tainment structure must be capable of withstanding the increased pressure (A) during the accident, where it must not exceed Service Level C Limits or the Factored Load Category (as previously specified in this paragraph) and (B) following inadvertent full inerting that may occur during normal plant operations, where it must not exceed either Service Level A Limits (for a steel containment) or the Service Load Category (for a concrete containment).

Furthermore, inadvertent full inerting must be safely accommodated during normal plant operations.

Modest deviations from these criteria will be considered by the Commission if good cause is shown.

The installation and operation of new hydrogen control systems at reactors with operating li~ censes is considered to be an unreviewed safety question as defined in 6 50.59(a)(2).

(iii) Effective January 31,1982, each boiling and pressurized light-water nuclear power reactor that does not rely upon an inerted 22 Enclosure "C" 1

[7590-01]

atmosphere to control hydrogen inside the containment, shall be provided with systems necessary to ensure safe shutdown and containment integrity that are capable of performing their functions during and after being exposed to the environmental conditions created by the burning (or local detonation) of hydrogen.

(iv)[(44)-9esign3 Analyses shall be performed [(A)] for [eaeh boiling] light-water _ nuclear power reactorf [fasility-for-whish-the app 14 sat 4en-fer-a-senstrustion-permit-was-desketed-after-July-1 -1972 3

and -(B)-for-eash-pressurized-light-water-Peaster-fas414ty3 to evaluate i

measures required [that-san-be-taken] to mitigate the consequences of large amounts of hydrogen generated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the start of an accident (hydrogen resulting from the reaction of up to [about] and including 75 percent of the fuel cladding with water). The required analyses will vary in scope, implementation date and acceptable accident scentrios depending on the specific type of nuclear power reactor con-tainment.

Each analysis must include the period of recovery from the degraded condition. The scope and implementation requirements for nuclear power reactors with different types of containments are as follows:

(A) Effective January 31, 1982, or the date of issuance of a license l

l authorizing operation at full power, whichever is sooner, analyses shall be performed for each boiling light-water nuclear power reactor with a Mark III type containment and each pressurized light-water nuclear power reactor with an ice condenser type containment. The analyses must justify the selection o f the hydrogen control system reouired by s 50.44(c)(3)(fi).

23 Enclosure "C"

[7590-01]

(B) For all other light-water nuclear power reactors, except boiling light-water nuclear power reactors with Mark I or Mark II type contairi-ments, the analyses must demonstrate that safe shutdown and containment integrity will be maintained with adequate safety margins, including the effect of local detonations. These [ design] analyses [and-a-proposed de s 4 g n-(e p-d e s ig n s )- to -m i t i ga t e-t he-s e n s e q ue n s e s-e f-hyd ro g en-in-se n t a in -

ment] shall be completed and submitted to the Commission by (6 months from effective date of rule) or the date of docketing of the application for the operating license, whichever is later.

{vl[ffii3--By-dene-38--1981] Effective July 1, 1981, [f.acilities]

nuclear power reactors that rely upon external recombiner_s.or.. purge /

repressurization systems to satisfy the requirements of._3.50.44.shall be provided with containment penetrations for the external recombiners or purge /repressurization systems that either:

(A) are dedicated to that service only, conform to the requirements of Criteria 54 and 56 of Appenaix A of this part, are designed against

(-.

l l

l 23a Enclosure "C" l

I l

[7590-01]

[

[an-assumed] postulated single failures, and are sized to satisfy the flow requirements of the external recombiners or purge /repressurization systems, or l

(B) are of a combined design for use by either external recombiners or purge /repressurization systems and other systems, conform to the l

requirements of Criteria 54 and 56 of Appendix A of this part, are designed against [an-assumed] postulated single failures both for containment isola-l tion purposes and for operation of the external recombiners or purge /

repressurization systems, and are sized to satisfy the flow requirements of the external recombiners or purge /repressurization systems.

(vi)[fiv3--By-danuary] Effective July 1, 1982, [ facilities] nuclear power reactors that rely t purge /repressurization systems as the pri-mary means for controlling ombustible gases following a LOCA shall be provided with either (A) internal recombiners or (B) the capability to install external recombiners, following the start of an accident.

The internal or external recombiners must [that] meet the combustible gas control requirements in paragraph (d) of this section.

The containment penetrations that are used must meet the criteria in paragraphs (c)(3)

[fiii3]{vl(A) and (c)(3)[fiii3]{vl(B) of this section applicable to external recombiners.

2.

A new $ 50.44a is added to 10 CFR Part 30 to read as follows:

$ 50.44a Design require.nents to mitigate the consequences of accidents resulting in a degraded core.

(a) High Poir.t Vents.

[By-danuary] Effective July 1, 1982, each boiling and pressurized light-water nuclear power reactor shall be 1

24 Enclosure "C"

[7590-01]

provided with high point vents for the reactor coolant system [and]

2 3

[high point-vents ] and other systems required to reactor vessel head2 maintain adequate cere cooling if the accumulation of nc.icondensible gases would cause their loss of function, remotely operated from the control room, to provide improved operational capability [for] to main-tain[ing] adequate core cooling following an accident.

High point vents are not required, however, for the tubes in U-tube steam generators.

Since these vents form a part of the reactor coolant pressure boundary, the design of the vents and associated controls, instruments and power sources must conform to the requirements of Appendix A and Appendix B of this part.

In particular, [these] the vent [s] system shall be designed to ensure a low probability that [in-such-a-way-that-no-single failure-could-rescit-in-either] (1) [a-ioss-of-the-capability-of] the vents [to] will not perform their safety functions [or] and (2) there would be inadvertent or irreversible actuation of a vent.

Furthermore, the use of these vents during and following an accident must not aggra-vate the challenge to the containment or the course of the accident.

(b)(1) Protection of Safety Equipment and Areas Which May be Used l

During and Following an Accident.

[By] Effective January 1, 1982, each l

(

boiling and pressurized light-water nuclear power reactor shall be pro-vided with both adequate access to areas [which] that may be used driring and following an accident and protection of safety equipment sr, that an accident [which] that results in the release of large amounts of radio-active material will not limit personnel occupancy or degrade safety equipment by the radiation fields that may exist during and following

[3Except-that-the-tubes-in-B-tube-steam ger.erators-de-not-require-venting-25 Enclosure "C"

[7590-01]

the accident to the extent that required safety functions cannot be accomplished.

(i) The facility design must be based on a release of radioactive material from the fuel to the primary coolant system that is not less than 100% of the core equilibrium noble gas inventory, 50% of the core equilibrium halogen inventory, and 1% of the remaining core fission products.

For equipment and areas affected by the reactor coolant, it shall be assumed that the above distribution of radioactive material is intimately mixed with the coolant water except that recirculated, depres-surized coolant water may be assumed to contain no noble gases.

For equipment and areas affected by the containment atmosphere, it shall be assumed that not less than 100% of the core equilibrium noble gas inven-tory and 25% of the core equilibrium halogen inventory are uniformly dis-persed in the containment atmosphere and an additional 25% of the core equilibrium halogen inventory and 1% of the remaining core fission pro-ducts are uniformly distributed on surfaces exposed to the containment atmosphere.

l (ii) The facility design basis must be such that an individual operator will not receive more than a 5 rem whole body dose, or its equivalent to any part of the body, while performing a necessary safety function during and following an accident.

1 (2) In-Plant Iodine Instrumentation.

[By-daneary-i--1981 2-e]Each boiling and pressurized light-water nuclear power reactor shall be provided with instrumentation, equipment and associated training and procedures for i

determining, under accident conditions, the airborne radiciodine concen-l tration in areas within the facility where plant personnel may be present during and following an accident.

l t

26 Enclosure "C" j

L'7590-01]

(c) Sampling During and Following an Accident.

[By] Effective January 1,1982, each boiling and pressurized light-water nuclear power reactor shall be provided with the capability for personnel to obtain and quantitatively analyze a reactor coolant or containment atmosphere sample during and following an accident.

(1) The facility design must be based on the radioactive material release terms described in paragraph (b)(1)(1) of this section.

(2) The design basis for the plant equipment that provides t[T]he capability to obtain and analyze a sample must be based on the assumption that it will [ineinde-the-capability-for-doing-so-] be done promptly, and without incurring a radiation exposure to any individual in excess of [3]

5 rem to the whole body, or its equivalent to any part of the body.

(3) The capability to quantitatively analyze a sample must be based on the use of either in-line monitoring or an onsite radiological and I

chemical analysis facility.

[--and] If in-line monitoring is chosen, a capability must be provided for backup sampling using grab samples, and must include the capability for analyzing the samples at either an onsite or offsite facility.

The analysis capability must provide, as needed, quantification of the following:

(i) Those radioisotopes necessary to [determina-the-degree] indi-cate the extent of core damage; (ii) Hydrogen in the containment atmosphere; (iii) Total dissolved gases [and],or dissolved hydrogen gas in the reactor coolant; (iv) Boron in the reactor coolant; and

[20r-30-days-after-the effective-date-of-the-ruie--whichever-is-iater-]

27 Enclosure "C"

[7590-01]

s (v) Chloride in the reactor coolant.

Chloride analyses may be performed o'ffsite and are not required to be done promptly.

(d)

Leakage Integrity Outside Containment.

(1) Each boiling and pressur; zed light-water nuclear power reactor

(

licensee shall implement leak reduction measures so that leakage, from l

systems outside containment (systems that would or could contain highly I

radioactive fluids during and following a serious transient or accident),

is eliminated or minimized to the maximum extent practicable to prevent l

the release of significant amounts of radioactive material during and l

l following an accident.

Consideration shall be given to reductions of potential release paths that could result from design or operator deficiencies.

(2) Each boiling and pressurized light-water nuclear power reactor licensee shall establish and implement a program of preventive maintenance i

to eliminate or minimize, to the maximum extent practicable, leakage from systems outside containment.

This program shall include periodic

[ integrated] leak tests of these systems at intervals not to exceed each refueling cycle and also include [as-well-as] the reduction of potential release p&ths by appropriate operator training.

(e) Accident Monitoring Instrumentation.

Each boiling and pres-

~

surized light-water nuclear power reactor shall have the capability dur-1 ing and following an accident for:

(1) Providing and recording in the control room a continuous indica-tion of:

(i) Containment pressure [by] effective January 1, 198[12]2; (ii) Hydrogen concentration in the containment atmosphere [by]

effective [0ctober-i--1981] January 1, 1982; (iii) Containment water level [by] effective January 1, 198[12]2; 28 Enclosure "C"

[7590-01]

4 (iv) Containment radiation level effective [by-Betober-i--1981]

January 1, 1982; (v) Radioactive noble gas concentrations in the plant gaseous effluents at all potential accident release paths effec-tive [by0eteber-i--1981] January 1, 1982; and (2) Quantifying the concentration of radiofodines and radioactive particulates in [the-airborne] plant caseous effluents at [each-anticipated]

all potential accident release [ point] paths effective [by-0ctober-i--1981]

January 1, 1982.

(3) All the instruments and monitoring systems used for accident monitoring shall be designed and qualified (with extended ranges) to perform their function following an accident characterized by the radio-active material release terms described in paragraph (b)(1)(i) of this section.

(f) Detection of Inadequate Core Cooling.

(1) Each boiling and pressurized light water nuclear power reactor licensee shall develop and implement procedures and training to be used by the operators to recognize the existence of inadequate core cooling and low coolant level in the reactor core using available instrumentation.

(2) Each pressurized light-water nuclear power reactor shall be provided with a primary coolant saturation meter (subcooling meter) that provides in the control room a continuous, recorded, on-line indication of the primary coolant saturation condition.

(3)

[By] Effective January 1,1982, each boiling and pressurized light-water nuclear power reactor shall be provided with an instrumentation system such as [a] reactor vessel water level indicators [which-sepplies]

for pressurized water reactors that augment the incore thermocouples; and i

29 Enclosure "C"

)

,-,w, a

v g-,

,,,y,-

4

,,w w

a e+w

--g,,--

y,

[7590-01]

incore thermocouples for boiling water reactors that augment the, reactor vessel water level indicators.

The instrumentation system must supply to the control room a recorded, unambiguous, [direet] easy-to-interpret, l

indication of inadequate core cooling.

The indication must cover the complete range from normal operation to complete core uncovering and give advance warning of the approach of inadequate core cooling.

(4) All instruments used to detect the existence of inadequate core cooling shall be designed and qualified to perform their function following an accident characterized by the radioactive material release terms described in paragraph (b)(1)(1) of this section.

(g) Training and Human Engineering to Mitigate Degraded Core Accidents.

[By-Aprii-i--1981;] [e]E_ach boiling and pressurized light-water nuclear power reactor licensee shall include in its training program for all l

operating personnel training to recognize, control and mitigate the con-sequences of accidents in which the core is severely damaged.

In addi-tion, each licensee shall support the development of its training pro-gram, emergency procedures and control room hardware, with applicable human engineering data.

The training [shali] must include the use of all available structures, systems and components that can control or i

mitigate degraded core accidents.

(Secs. 103, 161b., Pub. L.83-703, 68 Stat. 936, 948; Sec. 201, as amended,

[

Pub. L.93-438, 88 Stat. 1242 (42 U.S.C. 2133, 2201(b), 5841).)

j Dated at Washington, D.C. this day of 1981.

l For the Nuclear Regulatory Commission.

l l

Samuel J. Chilk, l

Secretary of the Commission.

30 Enclosure "C" l

l ENCLOSURE "D"

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