ML20215D953

Discusses CRGR 850603 Meeting Re App J Revs & Upcoming Meeting to Discuss Origin of Need & Technical Basis for leak-tight Containment.Relevant Info,Including App a & J Citations,Encl

ML20215D953
Person / Time
Issue date:	06/06/1985
From:	Ross D NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Committee To Review Generic Requirements
Shared Package
ML20215D926	List: IA-86-483, Lists 15 Excerpts from 23-yr Old TID-14844,w/comparisons to Such Sources as NUREG-0956 & Aps,Ans & Ucs Repts & Questions Progress Made ML20215D923 ML20215D930 ML20215D933 ML20215D940 ML20215D944 ML20215D953 ML20215D961 ML20215D965 ML20215D971 ML20215D978
References
FOIA-86-483 NUDOCS 8610140372
Download: ML20215D953 (14)

Text

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UNITED STATES I %p, NUCLEAR REGULATORY COMMISSION t r wasWNGTON, D. C. 20f,55

% e.... June 6, 1985 NOTE TO: CRGR

[

FROM: A f. Ross, Deputy Director Office of Nuclear Regulatory Research

SUBJECT:

APPENDIX J PHILOSOPHY ,

At the June 3, 1985, CRGR meeting on App. J revision, we decided to have another meeting to explore the underlying philosophy. If we are going to discuss the origin of the need and technical basis for a leak-tight containment, we should "look at the record." I am attaching the following as a partial record:

.- Enclosure A - Part 100 relevant pages Enclosure B - App.' J. citation Enclosure C - GDC (Appendix A to Part 50) citations Enclosure D - The ACRS report on Engineered Safeguards of 1964 Enclosure E - Some Thoughts on TID-14844 (Memo by Ross of April 3,1985)

I have the following points to make on these enclosures.

Enclosure A.

This portion of our regulations links the " expected demonstrable leak rate from the containment" to a postulated fission product release. Footnote 1 to 10 CFR 100.11 provides general guidance as to severity of core damage. The note at the end of this part further references some methods in TID-14844, mostly in context of site evaluation (which is the purpose of paragraph 100.11).

Enclosure B. .

Definition A under Part II, Explanation of Tenns, states that the primary reactor containment serves as an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment.

i.

Enclosure C.

GDC-50 relates the containment design basis to the pressure-temperature conditions resulting from any loss-of-coolant accident. (NB: of course, these pressure-temperature conditions are weakly related, at best, to severe accident pressure-temperature values.)

Enclosure D.

The section on containment notes that the containment is designed to prevent the escape of injurious amounts of fission products resulting from any failure in a nuclear system. Increased leakage rate is stated to be an indication of l

deterioration. Even then (p. 4 of their Addendum #1) we see: " Continuous 8610140372 860926 TB 6-483 PDR

s monitoring of leakage where possible is desirable as a check against misoperation."

i Enclosure E.

TID-14844 in many places emphasizes the importance of the containment as a factor in mitigating off-site doses.

Discussion The most vexing problem that we have is the paradox between the pressure load used as a basis for containment design (the LOCA) and the pressure loads associated with severe accidents.

To illustrate this point I have enclosed (Enclosure F) some pressure-time histories for a variety of severe accidents. For each of these accidents there is an associated fission product inventory in the containment (available for release) and, for each accident, there is a variety of containment failure g modes. In no case does the containment pressure go up to the design basis, and-then sit there for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, with a high load of fission products, and leak at tech. spec. values.

The escape from this paradox may be found in a reading of the future of App. J, when the full benefits of the reappraisal of severe accidents is in hand. One outcome is illustrated blackout (TMLB') by examining)the for Surry containment (Table 1. Consider event treeprobabilities the conditional for the station for the central columns. This is one perception of how containments behave, given TMLB'. If one sunned the probabilities, weighted with dose, it is likely ,

that the dose contribution due to non-failure leakage (so-called Appendix J 1eakage) will be small.

And, the pressure trace for TMLB' is characteristic (in large PWRs) of slowly-davelnninn eccidents wherein App. J 1eakage would presumably be relevant.

Perhaps of more relevance is that unseemly leakage from a containment with '

damage fuel inside would thwart recovery operations by on-site personnel.

IreadportionsofNUREG/CR-4220(PNL-5432),ReliabilityAnfysisof l Containment Isolation System. That report uses 10 10 for "large-leak" unavailability (e.g., the/3 failure mode. From Table 1 (the Surry tree), the value for 14 3 between 10 and

',rge 10isoittion failure

-4 . If Case 1 oforCentral preexisting columnleak is 2x10-3 of Table which is 1 is inspected, .- '

then there are four non-zero failure modes:

Basemat melt (9%)

Late overpressure failure by pressure relief (3%)

Late overpressure rupture (29%)

Late H2 Burn Failure (20%)

Also, in this column, 38% of the time the containment stays intact (presumably leaking a' la' App. J).

If power and cooling were restored in a reasonable time period (a few hours) the 38% non-failure probability should increase. For large, stronger containments the non-failure conditional In such cases, the value of a tight containment (probability vs. one thatisleaked even higher.

at, say,1%)

would be relatively high both on and off site. Also note from Table 1, that suppressing the /3 mode barely changes the 60% conditional probability that the containment fails in some manner (albeit hours later). Consider the

, following arithmetic:

Risk is somehow proportional to the product of failure probability multiplied by the leak rate. For the /3 -mode, assume (for illustration) a pre-existing four-in. dia. hole. T,he risk of App.jtypeleakageis.1%perday. The risk of fj is(Of about q 2x10- multipled by 100%/ day which is .2% per day. course, -

risk is not %/ day units, but this is a comparative illustration.)

This shows that suppressing g would be effective, but could eliminate around half the severe accident risk, given that the containment otherwise remains intact.

I j Conclusions

1. Pressure-time histories prescribed by GDC-50 and fission product time histories prescribed by Part 100 (and, by extension, TID-14844 and Reg Guides 1.3, 1.4) are technically inconsistent. App. J cannot be modified to resolve any paradox therein.

2. An essentially leak-tight containment will continue to be of importance in coping with severe accidents, although probably of more importance on-site, in the first few hours.

3. Risk to the public, measured in probability of exceeding a given dose at a given distance, will eventually prove to be weakly dependent on La, for low ( /_1%) values of La.

, 4.Pressuremanagementcansuppressthegfailuremodemuchas augmented ISI suppressed the V failure mode. This was

recognized over 20 years ago (Enclosure D).

5. Possibly of greater importance then type B or C tests in terms of risk to the public is pressure-temperature-radiation qualification assurance of penetrations and isolation values, especially when subjected to severe accident time-histories as depicted in Enclosure F.

I

't  ;

S. The CRGR should not that a) some notion of the foregoing should be reflected in the SOC, and b) it should recommend to the EDO that an integrated study of the role of containment performance in preventing injurious radiation to the public be expedited during the application phase of severe accident policy implementation.

D. . Ross, Deputy Director Office of Nuclear Regulatory Research

Enclosures:

As stated s

O

TA3LE I RESULTS FOR SURRY THLD' (Station Blackout, Including Loss of Auxiliary Feedwater)

CONTA!8NIENT PAILUBE N008 pa08482L17388 Opt. CENTRAL ptSS.

CASB le CASE 2* CASE 1* CASE 2+

us contagneont pailure .75 .38 .1d .02 ..

sesemat meltthrough only .13 .09 .33 .002 Late Overpressurasation lyressure Relief) .31 .03 .03 .. ..

Late Overpressurasatten (Rupture) .01 .29 .25 .58 .04 .

Late mydrogen turn .. .20 .42 .30 .03 I tarly stese Spite and/or Vessel Depressurisation .. .. .10 .. '

terly Steen Spits /Depres. e Nydrogen Burn =.

. Estly stee e

.005

.g2

)

. , e. in-vessei .ntea. spite8 / iosion (.sarect s ,e. - 1 i.arge isolation ailure or p,.esassing

,3,Depres. asse 11.,cn> . steating

.. .. .. .. 01 praction of Above Affected tyr (a) preesisting small Containment Leak ** .01 .03 .03 .05 .05 Iti Induced small steam Generator Tune Leak'* .. .001 .001 .10 .30 I!!. PRINC! PAL CONTAINntNT PATNWAYS n An induced failure of the het leg or induced seal LOCA usually occura. In the astrastic walethreugh, spell SCTas are possante, vessel failure modes are pressure e3ectaen er seltthrough accespanying induced hot leg ruptures, er in. vessel steam emplessons faalang the bottom head an the central and possamastat weltthroughs. Vapotasataan release almos

. always occurs and estaatton release almost always occurs in the central and possamaatte

. walkthrougas.

SM!.2104 CALCULA!!ON

11) No Contst**ent Faf get Sprays recovered lates a permanently No. tut consequences too.anae deoras sea otten forma.

samalar to selow.

(2) Bese*at weltteroueht Sprays recovered in approsisately half of Yes. (no steer esplossens, tne casess a permanently coolable doorts bed sa not formed. anduced RC8 faalures, or direct heasangi.

(3) Late overcressurination: Sprays either do not recover er are so.

tasaes ey a 4ater nycrogen surn that does not feal containment.

A doorts ted may to temperataly cooled but drnes out siter spray faalute. The auss11ery Dualdang is typassed.

(4) Late uvdrosen turnt sprays are recovered before the turn, but No.

say real agaan arter containment rupture. Ausilasty buildang ,

typassed. I

15) Stea* Seite/Depressuritation/ Direct Westinet spreys never operate Yes, (no steen emplessons, pract to contaanevnt rea4ure. Medfrate prosettlity of later dateet heatang, or anduce:

tecovery. The ausaltery saaldang as sypassed. RCS faalutess.

e l (6) foolation Fa11eret Either no grose containment failure occurs er No.

Enere as a daract hosting failure at vessel areach. Sprays may or may not recover. The teolation failure is usually aitagated by the ausalasty building. But later gross contaanment faalutes typass the ausiliary bualdang.

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• Case 2 for central and possimistie weltthroughs includes the effects of an. vessel steam esplosions and es vessel direct heating. Case 1 does not include these effects. Street heetang includes hydrogen turning af the atmosphere is fleRPaale. See test fer a discussion of limatatsens associated wath the results for steam esplossons and direct testing.

i ** small centsineent leets include those producing lestage greater than containment desten eut I not large enough to preclude gradual overpressurisation from steam and/or mentendenstales.

( Likewise, small steem generator tube loans include those producang pramary to-secondary leanage greater than desagn but not large enougn to depressurase tne primary system selow the i

accumulator Gespeint before vessel areacn.

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e EHCLOSURE A 10 CFR 100.11 e

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total radiation dose in excess of 300 rem to the thyroid from lodine expo-8

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accidentally flow into nearby streamsor rivers or might find ready

.:"- access to sure.

(2) A low pop stion zone of such .

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underground water (d) Where tables, unfavorable physical size that an indiv dual located at any j s.

- , %J characteristics of the site exist.point on its nevertheless outer boimdary who be ;sresult-ex- ,

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proposed site posed to the radioactive cloud N"'

ing from the postulated fission prod- ..

found compensating to be engineering acceptableuct if release the design (during the ofthe facility entire period of includes appropriate '4 and adequate safeguards.

Its passage) would not receive a totalradiation W' .. dose to (21 FR 3509. Apr.12.1962. as amended excess dose inat of 25 rem or a total 38 of 300 rem to the thy-excess radiation 5 49 yR 31261. Nov.13.19'l3) '

rold from todine exposure.(3) A population center"% .

distance*Y' of g100.11 Determination of emeluelon meta.  ?

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at least one and one-third times the' /-

Iow population zone. and population ^$%

eenter distance. distance from the reactor to the outer f

.l (a) As an aid in evaluating a pro- boundary of the low population sone. -

1 .

' posed site. an applicant should assume In applying this guide. the s fission produce release' from thetermined upon consideration of popu-boundaryof the population 34 -

ce core, the expected demonstrable leak -

.d V i rate from the containment lationanddistribution.

themeteorological Politicalconditions ide. Where very large bound-triespertinent are ?' not controll to his site to derive an exclusion cities arestes may be necessary alow involved, population a greater because distance of totalinte-zone and population 1; e A

. catio center distance. For the purpose of this analysis, which shallgrated set forth population thebasis dose for consideration.(b) the numerical values For%siles -

used. for m the applicant should determine the cilities comideration should be given -

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following- to the folloving:

(1) If the reactors are independent e

(1) An exclusion area of such size that an individual located at any point m to the extent that an accident in one

• on its boundary for two hours reactor in another, would immedi-stely not initiatefollowing the size of the exclusion an accident onset of the postulate c- -

r '

e area, low population zone and popula-fission product release would not re-eeive a total radiation dose to the whole body in excess of 25 rem

• or a tion center distance shall be fulfilledwith respect to e 3 _

g Ce

• ually. The envelopes of the plan over-m d

'The fission product release assumed for these cateulations should be based upon a lay of the areas so calculated shallthen be taken a .

Et major accident, hypothestzed for purposes boundaries.

Gf of site analysts or postulated (2) from If theconsider-attons reactors are interconnect- of posa!ble gy g

would result in potential hazards not ex- ed to the extent that an accidentaccider inone reactCr tal events, Couth Ste teeded br those from any accident consid- operation of any other, the size of the 4 ered credible. Such accidents have sencrally exclusion area. Iow population zone th- been assumed to result in substantial and population melt,down center of the core with distance shallbe subsequent based l- relea upo ser tnd appreciable quanttttes of flaston products,'The whole body dose of 25 rem referred to above corresponda numericaHy to theinterconnected reactors emit their pos-tulated fission lth2 once in a lifetime accidental or emersenerdose far radiation workers which, accord!ns (rea ---

to NCRP recommendations may be disre- dent body value conditions.

and the 300 Rather.

rem thyroidthisvalue 25 rem whole l

M* sarded in the ation exposure determination status (see NBS Handbook of havetheir radt- been set forth in these sutdes as refer-ence valu nr.y 69 dated June S,1959). However neither its s cf use nor that of the 300 rem value for thy-untion of reactor sites with respect to poten-r"m itns told exposure as set forth in these site crite-ria sutdes are intended to imply that these tial reactor accidents of exceedingly lowprobability o public exposure to radiation.

O be cant numbers constitute acceptable limits foremergency doses to the public under acci- .

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Part 100, App. A 10 CFR Ch. I O-1-85 Edition) taneously. This requirement may be proposed sites for the suitability nuclear of the plant power plantses-design bases and reduced in relation to the degree of tablished in consideration of the seismic and c characMaues of the proposed coupling between reactors, bility of concomitant accidents theand proba, [e[o the probability that an individual These criteria are based on the limited would not be exposed to the radiation geophysical and geological information effects from simultaneous releases. available to date concerning faults and The applicant would be expected to earthquake be revised occurrence as necessaryand effect.

when moreThey will complete justify to the satisfaction of the Com- information becomes available.

' mission the basis for such a reduction , ,

in the source term.

(3) The applicant is expected to These criteria, which apply to asuelear show that the simultaneous operation power plants, describe the nature of the in-of multiple reactors at a site will not vestigations required to obtain the geologic result in total radioactive eifluent re- and seismic data necessary to determine site n

s leases beyond the allowable limits of suitability and provide reasonable assurance that a nuclear power plant can be construct.

y applicable regulations, ed and operated at a proposed site without M.

• Novs: For further guidance in developing undue risk to the health and safety of the G. = . the exclusion area, the low population sone, public. They describe procedures for deter-94 and the population center distance, refer- mining the quantitative vibratory ground

? enee is made to Technical Information Doc- motion design basis at a site due to earth-A..a ument 14844. dated March 23.1962, which quakes and describe information needed to c ..r contains a procedural method and a sample determine whether and to what extent a nu-

• y . calculation that resu?t in distances roughly clear power plant need be designed to with-reflecting current siting-practices of the stand the effects of surface faulting. Other

. Commission. The calculations described in geologie and seismic factors required la be Technical Information Document 14844 taken into account in the siting and den

. may be used as a point of departure for con- The of nuclear power plants investigations describedare identified.

in this ap-

.w alderation of particular site requirements pendix are within the scope of invastigs-which may result from evaluation of the

.a characteristics of e particular reactor, its tions permitted by l 50.10(cM11 of this chap-ter.

purpose and metted of operation. Each applicant for a construction permit W >* . Copies of Technical Information Doc 5 shallinvestigate all seismic and geologic fac-

.. *

• ment 14844 may be obtained from the Com*

. mission's Public Document Room.1117 H tors that may affect the des!gn and oper.

. Street NW., Washington, D.C., or by writing ation of the proposed nuclear power plant )

~ the Director of Nuclear Reactor Regulation. trrespective of whether such factors are ex- '

~

U.S. Nuclear Regulatory Commission. piteitly included in these criteria. Additional Washington. D.C. 20555.' . ' investigations and/or more conservative de-terminations than those included in these 127 FR 3509. Apr.12,1962, as amended at 31 criteria may be required for sites located in al. ' '

FR 4670. Mar.19,1966; 38 FR 1273. Jan.11, areas having complex geology or in areas of 1973: 40 FR 8793. Mar. 3.1975: 40 FR 26527 high seismicity. If an applicant believes that

,'9

June 24,19751 the particular seismology and geology of a f site indicate that some of these criteria, or

- portions thereof, need not be satisfied, the J *. . -

AFFENDix A-Sttsnt!C AND Grotocic specific sections of these criteria should be SITINo CRITERIA yoR NUCLEAR' POWER identified in the license application, and ,

supporting data to justify clearly such de- l PIANTS -*

• partures should be presented.

1.FUaPoss These criteria do not address investiga-b '. -

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General Design Criterion 2 of Appendix A tions of volcanic phenomena required for to Part 50 of this chapter requires that nu- sites located in areas of volcanic activity. In-

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clear power plant structures, systems and vestigations of the volcanic aspects of such components important to safety be designed sites will be determined on a case-by-case

.. g to withstand the effects of natural phenom- basis.

ena such as earthquakes, tornadoes, hurri-

' a. g -

canes, floods, tsunami, and seiches without Irr.Dzrtwrrtows

, loss of capability to perform their safety As used in these criteria:

v. - functicns. It is the purpose of these criteria (a) The " magnitude" of an earthquake is a W" -

- to set forth the principal seismic and geo- measure of the size of an earthquake and is

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logic considerations which guide the Com- related to the energy released in the form of

.. mission in its evaluation of the suitability of

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l EllCLOSURE B CURREilT 10 CFR 50 APPEilDIX J DEFINITIONS i.

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10 CFR Ch. I d.1-45 Edition) Nuclec Port 50, App. J

0. " Type B Tests" means tests intended close s I. Irraonverrow to detect local leaks and to measure leakage undera vented One of the conditions of all operating II. across each pnasun-containing or leakage. O.*i censes for water cooled power reactors as limiting boundary for the following primary ard at I

specified in 5 50.54to) is that primary reac. reactor containment penetrations: pared-I. Containment penetrations whose design abillt) ment leakage test requirements set forth in incorporates resilient seals. gaskets or seal- iting t

' videtor containments shall expansionmeet bellows, andthe this appendix.These test requirements pro. ant componds, piping penetrations fitted for preoperational and periodic verift. with electricalcontain.

cation by tests of the leak. tight integrity of trations fitted with flexible metal seal as.

pene. II the primary reactor containment, and sys, Ar semblies. condt tems and components which penetrate con- 2. Air lock door seals. including door oper-devel

~

tainment of water. cooled power reactors. ating mechanism penetrations which are tor c<

and establish the acceptance criteria for part of the containment pressure boundary. comp such tests. The purposes of the tests are to 3. Doors with resilient seals or gaskets ment assure that (a) leakage through the primary except for seal. welded doors. Up

- reactor containment and systems and com. 4. Components other than those listed in prim:

.\ ponents penetrating primary containment 11.0.1.11.0.2. or 11.0.3 which must meet stalk shall not exceed allowable leakage rate the acceptance criteria in III.B.S. fluid values as specified in the technicsi specifica. H. " Type C Tests" means tests inteMed to tems tions or associated bases and (b) periodic measure containment isolation valve Icak. talm surveillance of reactor containment penetra. age rates. The containment isolation valves any:

tions and isolation valves is performed so included are those that; and thgt proper maintenance and repairs are I. Provide a direct connection between the ble.

made during the service life of the contain. Inside and outside atmospheres of the pri. the*

ment, and systems and components pene. mary reactor containment under normal op. A-trating primary containment. These test re. eration. such as purge and ventilation. Con-quirements may also be used for guidance in Vacuum relief. and instrument valves; V.A establishing appropriate containment leak. 2. Are required to close automatically the age test requirements in technical specifica. upon ' receipt of a containment isolation the

- tions or associated bases for other types of .alsnal in response to controls intended to tair

! nuclear power reactors. etfect containmentisolation; the

3. Are required to operate intermittently shs II. EEFLaN ATIoM oF TsMis under postaccident conditions; and '

p ,'

• A. " Primary reactor containment" means 4. Are in main steam and feedwater piping ,

as I the structure or vessel that encloses the and other systems which penetrate contain. 8

,{

components of the reactor coolant pressure ment of direct cycle boiling water power re.

boundary, as defined in 5 50.2(v). and serves actors. sut as an essentially leak tight barrier against I. Pa (p. sis.) means the calculated peak ,,

the uncontrolled release of radioactivity to containment internal pressure related to les.

eg the environment. the design basis accident and specified B. " Containment isolation valve" means either in the technical specification or asso. idl l any valve which is relied upon to perform a ciated bases. ch containment isolation function, J. PS (p.sig.) means the containment IIJ }

C. " Reactor containment leakage test pro. . vessel reduced test pressure selected to ps gi gram" includes the performance of Type A. measure the integrated leakage rate during wj Type B. and Type C tests. described in JLP. periodicType Atests, II.O. and II.H. respectively. K.14 (percent /24 hours) means the maxi-I th D. "Isakage rate" for test purposes is that mum allowable leakage rate at pressure Pa l II.

leakage which occurs in a unit of time, as specified for preoperational tests in the i n:

at stated as a percentage of weight of the origi. technical specifications or associated bases, nal content of containment air at the leak. and as spectfled for periodic tests in the op-fr.

. ec age rate test pressure that escapes to the erating license. '

p.

~ outside atmosphere during a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test L. Ld (percent /24 hours) means the design t1

. i period. leakage rate at pressure. Pa. as specified in E. "Overall integrated leakage rate" N 11 the technical specifications or associated t' means that leakage rate which obtains from l t.

bases. -

a summation of leakage through all poten. M. Lt (percent /24 hours) means the maxi. I

• tial leakage paths including containment mum allowable leakage rate at pressure Pt

. welds. valves, fittings. and components derived from the preoperational test data as 1 k

which penetrate containrpent. specified in III.A.4.(a)(111). 3

/

• F. " Type A Tests" means tests intended to N. Imm. Ltm (percent /24 hours) means measure the primary reactor containment the total measured containment leakage

. overall integrated leakass rate'(1) after the rates at pressure Pa and Pt. respectively. ob-g containment has been completed and is tained from testing the containment with 4 .

ready for operation. and (2) at periodic in. components and systems in the state as r tervals thereafter.

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ENCLOSURE C

i. 10 CFR 50 APPEllDIX A GDC 50-57 i

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o~- - 1 Ao Nuclear Regulatory Commission Part 50, App. A

' ..uv detection. and isolation capabilities shall be boundary material during operation, main- " '

provided to assure that for onsite electric tenance, testing, and postulated accident -

po71r system operation (assuming offsite conditions, and the uncertainties in deter-powIr is not available) and for offsite elec- mining (1) matertal properties, (2) residual, .

tric power system operation (assuming steady state. trid transient stresses, and (3) (

Cnsite power is not available) the system -size of flaws. '

safIty function can be accomplished, assum- Criterion $2-CapcHlity for containment ing a single failure. frakape rate testing. The reactor contain- 4 Criterion ($-Inspection of cooling water ment and other equipment which may be .

system. The cooling water system shall be subjected to containment test coaditions -

designed to permit appropriate periodic in- shall be designed so that periodic integrated spection of important components, such as leakage rate testing can be conducted at

, h%t exchangers and piping, to assure the containment design pressure.

integrity and capability of the system.

1 Criterion ds-Testing of cooling water Criterion $J-Provisions for containment testing and inspection. The reactor contain-j system. The cooling water system shall be ment shall be designed to permit (1) appro- *-

i designed to permit appropriate periodic priate periodic inspection of all important -

pressure and funct19nal testing to assure (1) areas, such as penetrations, (2) an appropri- *

-t

', the structural and leaktight integrity of its ate surveillance program, and (3) periodic-components (2) the operability and the per- testing at containment design pressure of fIrmance of the active components of the the leaktightness of. penetrations which g . system, and (3) the operability of the have resilient seals ahd expansion bellcws*

system as a whole and, under conditions as Criterion $d-Piping systems penetrating

) close to design as practical, the performance containment Piping systems penetrating

' Cf the full operational sequence that brings primary reactor containment shall be pro-th2 system into operation for reactor shut, vided with leak detection, isolation, and con-down and for loss-of-coolant accidents. in-cluding opcation of applicable portions of tainment , capabilities having redundancy, th? protection system and the transfer be, reliability, and performance capabilities which reflect the importance to safety of tween normal and emergency powee sources.

isolating these piping systems. Such piping F. Reactor containment systems shall be designed with a capability to test pettodically the operability of the criterion 30-Containment desian bo4(s. Isolation valves and associated apparatus -

Th2 reactor containment structure. Includ- andto t ine if valve leakage is within ing access openings, penetrations, and the p containment heat removal system shall be Criterion $$-Reactor coolant pressure .

designed so that the containment structure boundary penetrating containment Each and its internal compartments can accom-modate, without exceeding the design leag. I!ne that is part of the reactor coolant pres-age rate and with sufficient margin, the cal. sure boundary and that penetrates primary cut ted pressure and temperature condi. reactor containment shall be provided with tion 2 rm!&r fvn any less of coolant accl. contain nent isolation valves as follows, dent. This margin shall reflect consider, unless it can be demonstrated.that the con-C.tlin of (1) the effects of potential energy tainment isolation provisions for a specific .

sources which have not been included in the class of lines, such as instrument lines, are detrrmination of the peak conditions, such acceptable oJi some other defined basis-as Energy in steam generators and as re. (1) One locked closed isolation valve insite quired by l 50.44 energy from metal water and one locked closed isolation valve outside and other chemical reactions that may containment; or result from degradation but not total fa!!ure (2) One automatic isolation valve inside gf Emergency core cooling functioning. (2) and one locked closed isolation valve outside th2 limited experience and experimental containment; or data available for defining accident phe- (3) One locked closed isolation valve inside nomina and containment responses, and (3) and one automatic isolnle veve outside th? conservatism of the calculational model containment. A simple cf eck valve may not and input parameters. be used as the automatic isolation valve out-Criterion $f-Fracture prevention of con- side containment; or f inment pressure boundary. The reactor (4) One automatic isolation valve inside '

cont nmentiboundary shall be designed and one automatic isolation valve outatie cith sufficient. margin to assure that under containment. A simple check valve may not ,

cerating, maintenance, testing, and postu' be used as the automatic isolation valve out- I litid accident conditions (1)its ferritic ma- side containment.

• tert'is behave in a nonbrittle manner and ys s

(2) the probability of rapidly propagating Isolation valves outside containment shall 3 frncture is minimized. The design shall re- be located as close to containment as practi-flect consideration of service temperatures cal and upon loss of actuating power auto- g and other conditions of the containment matic isolation valves shall be designed to g 1

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P 10 CFR Ch. I (1-1-85 Edition),

I j Port 50, App. A t take the position that provides greater means to control suitably the release of ra-dioactive materials in gaseous and liquid ef.

safety.

Other appropriate requirements to minl- fluents and to handle radioactive solid

• m!ze the probability or consequences of an wastes produced during normal reactor op.

accidental rupture of these lines or of lines erstion, including anticipated operational connected to them shall be provided as nec- occurrences. Sufficient holdup capacity essary to assure adequate safety. Determl- shall be provided for retention of gaseous nation of the appropriateness of these re- and liquid effluents containing radioactive quirements. such as higher quality in materials particularly where unfavorable design. fabrication, and testing, additional site environmental conditions can be expect-provisions for inservice inspection protec* ed to impose unusual operational limita.

tion against more severe natural phenom- tions upon the release of such effluents to ena, and additional isolation valves and con- the environment.

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' tainment, shall include consideration of the Criterion 81-Fuel storspe and handling

. population density, use characteristics, and and radioactivity control The fuel storage physical characteristics of the site environs. and handling. radioactive waste, and other Criterion 58-Primary containment (sola- systems which may contain radioactivity tion. Each line that connects directly to the shall be designed to assure adequate safety containment atmosphere and penetrates under normal and postulated accident con-primary reactor containment sha!! be pm- ditions. These systems shall be designed (1) vided with containment isolation valves as with a capability to permit appropriate pert.

follows, unless it can be demonstrated that odic inspection and testing of components the containment isolation provisions for a important to safety (2) with sultable shield-specific class of lines, such as instrument ing for radiation protection. (3) with appro-I es. are acceptable on some other defined priate containment, confinement, and !!!ter.

.ing systems. (4) with a residual heat remov-(1) One locked closed isolation valve inside al capability having reliability and testabt.

and one locked closed Isolation valve outside lity that reflects the importance to safety of

. containment; or decay heat and other residual heat removal, (2) One automatic isolation, valve ins!de and (5) to prevent significant reduction in and one locked closed isolation valve outside fuel storage coolant inventory under acci-containment;or, dent conditions.

(3) One locked closed isolation valve inalde Criterion #2-Presention of criticality in

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- and one automatic isolation valve outside fuel storspe and handling. Criticality in the

- containment. A simple check valve may not fuel storage and handling system shall be

• be used as the automatic isolation valve out. prevent *d by physical systems or prar*===

side containment; or W

(4) One automatic isolation valve inside preferkbly by use of geometrically safe con-

~- and one automatic isolation valve outside figurations.

containment. A simple check valve may not criterion 83-Nonitoring fhel and seaste be used as the automatic isolation valve cut- storspe Appropriate systems shall be pro-vided in fuel storage and radioactive waste side containment. systems and associated handling areas (1) to

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Isolation valves outalde con}ainment shall detect conditions that may result in loss of be located as close to the containment as residual heat removal capability and exces-practical and upon loss of actuating power, sive radiation levels and (2) to initiate '

ap-automatic isolation valves shall be designed propriate safety actions.

to take th'e position that provides greater Criterion $d-Nonitoring radioactivity re-safety. leases. Means shall be provided for monitor.

Criterion 87-Closed system isolation ing the reactor containment atmosphere, valves. Each line that penetrates primary spaces containing components for recircula-reactor containment and is neither part of tion of loss of. coolant accident fluids, efflu-the reactor coolant pressure boundary nor ent discharge paths, and the plant environs connected directly to the containment at. for radioactivity that may be released from mosphere shall have at least one contain. norma) operations. Including anticipated ment isolation valve which shall be either

- automatie, or locked closed, or capable of operational occurrences, and from postulat- j remote manual operation. This valve shall ed accidents.

be outside containment and located sa close (Bec.181. as amended. Pub. L 83-703, 68 to the containment as practical. A simple 8 tat. 948 (42 U.S.C. 2201); sec. 201, as check valve may not be used as the auto- amended. Pub. L 93-438,88 Stat.1242. Pub.

g matic isolation valve. L 94-79. 89 Stat. 413 (42 U.S.C. 5841))

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VL Fuel and Jtadioactivity Control 136 FR 3256. Feb. 20.1971. as amended at 36 FR 12'l33 July 7,1971; 41 FR 8258. Feb.12.

i Crtfeffon 60-Control c/ releases of radio- 1976; 43 FR 50163. Oct. 27,1978) cetive materials to IAe environment The nuclear power unit design shall include

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ENCLOSURE D 1964 ACRS REPORT .

ON ENGINEERED SAFEGUARDS 1

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