ML20147G924

Reg Guide 1.7,Revision 2, Control of Combustible Gas Concentrations in Containment Following a Loca

ML20147G924
Person / Time
Issue date:	11/30/1978
From:	NRC OFFICE OF STANDARDS DEVELOPMENT
To:
Shared Package
ML20147G921	List: ML20147G917 ML20147G924
References
REGGD-01.007, REGGD-1.007, NUDOCS 7812270049
Download: ML20147G924 (6)

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Revision 2 U.S. NUCLEAR REGULATORY COMMISSION November 1978 fr %,%

' REGULATORY GUIDE bf e eee e OFFICE OF STANDARDS DEVELOPMENT P7ULATORY GUIDE 1,7 CGNTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS OF-COOLANT ACCIDENT i

l A. INTRODUCTION 1. Metal-water reaction involving the zirconi- l uin fuel cladding and the reactor coolant.

Criterion 05, " Emergency Core Cooling " of Appendix A, " General Design Criterta for Nu- 2. Radiolytic decomposition of the postacci-elear Pow-r Plants , ' to 10 CFR Part 50, "Do- dent emergency cooling solutions (oxygen will mestic Licensing of Production and Utihzation also evolve in this process). ,

Facilitie s , ' requires that a .5ystem be provided 1 to p rovide abundant emergency core coohng. 3 Corrostun of metals by solutions used for Criterion 50. "Contamment Design Basis, as emergency coolmg or containment spray.

amended, requires that the reactor contamment structure be designed to accommodate, without if a sutTicient amount of hydrogen is gener-exceeding the design leakage rate, conditions ated , it may react with the oxygen present tn that may result from degralation, but not total the containment atmosphere or, m the case of fanure, of emergency core cooling functiontng. merted contatnments , with the oxygen gener-Criterion 41, " Con t ainmen t Atmosphere Clean- ated fullowmg the accident. The reaction could up," requires that systems to control hydro- take place at rates rapid enough to lead to high gen, oxygen, and other substances that may temperatures and significant overpressurization be released in to the reactor containment be M the con tainmen t , which could result in a provided as necessary to control the concen- breaching of con tainmen t or a leakage rate tra tions of such substances following postu- above that specified as a limiting condition for lated accidents and ensure that con tainment operation in the Technical Specifications of the integrity 13 mam tained . hetnse. Damage to systems and components es-sential to the continued control of the post-

• - In addition . the Commission has published Lor A conditions could also occur.

amendments to Part 50 in which a new s 50.44,

" Standards for Combustible Gas Control Sys- The extent of metal water reaction and asso-l tems in Ligh t-Wa te r-Cocied Power Reactors. ' crated hydrogen production depends strongly

was added. This guide describes methods that on the course of even ts assumed for the I would be acceptable to the NRC staff for imple- accident and on the effectiveness of emergency

'menting these' regulations for light-water reae. cooling systems Evaluations of the perform-tor plants with cylindncal, ztrealoy-clad oxtde ance of emergency core cooling systems (ECCS) fuel. Light-water reactor plants with statnless meluded as engmeered safety featu res on steel cladding and those with noncylindrical current light-water-cooled reactor plants have claddmg will continue to t;e considered on an been made by reactor designers us tn g individual basis , analytica! models described in the "In t ermi Acceptance Criteria for Emergency Core Cool-B. DISCUSSION ing Systems for Light-Water Power Reactors" published in the Federal Register on June 29, Following a loss-of-coolant accident (LOC A ), 1371. and as amended on December 18, 1971 "

hydrogen gas may aceumulate within the con-tainment as a result of:

tanes .ndic.te substanuve enantes frcm previous Lssue. 16 FR 13 Ana J4 FR OW USNRC REGULATORY GUiOES Comments snoua d ce sent to me sacreta<v of me Comm>soon u S. wciear Requwtory Commission Wastungton O C. 2c556. Attenton Cocneting 74!

Requietory Guides ave saued to descree and etwas avanecie to the puunc Service Branen.

memods accontante to me NRC staff og emogmennng soucihc parts of the Cammession s segulatons to desineare techniques used by the sta*t m evefu. The guides are issued in the foisowmq ten Droad divis or's etmg specific prootema or postuered scorfents or to provide gu.cance to r

appa: Cants Regulatory csuices are not su)stitutes for reqWefions and Com. t Power 8teactors 6 Products chance eith team e not requeed Wmoos and sosutons $t'erent from those 2. Research and Test Aeactors E Transportation set out in the guides *,H be acceptacie if they provme a Deses for me fmdings 3 Fueis and Wteriars facilit4es 8 occupatonai Herm reouisate to the .ssuance or conttrwence of a permit or aconse by the 4 Environmentat ano sitmg 9 Anotrust and Fmancial Review Commismon. 5 Marews and Plant Protecton 10. Generai Reauests 'or smgie coones of :ssued guides <wnich may be encroducem or +or N,* Cornments and suggestens Nr imorevements m these guides we encournoed at osacement on an automatic destreuten hst for single CoDies Of 'Uture gusops all nmes, and guides edi be rev sed as appropriate to accommooate comments in specific amssons snound De mede m writmg to the U S Nucteer Requietory and to rettect new mtormatinn or 'tuperience Thas Jume *ds revesed as a result CommtsshA Washmgton DC 20565 Artention Deector Oms.on af of substantpre comments reCelved t'em me DvChC and additional Staff review TeChneCal mforreation and OOcument Confrot 7 812 2 7 0 0 </rj  !

. c These calculations are further discussed in the 5 50.16), it is consistent with the consideration staff's concluding stateraent in the rulemaking of the potential for degraded ECCS perform-hearing on the Acceptance Criteria , Docket ance discussed above to establish also a lower RM 1.

• The result of such evaluations is limit on the assumed amount of hydrogen gen-that, for plants of current design operated in erated by metal-water reactions in establishing conformance with the Interim Acceptance Cri- combustible gas control system requirements.

te ria , the calculated metal-water reaction In establishing this lower limit, the staff has amounts to only a fraction of one percent of the considered the fact that the max: mum metal-fuel cladding mass. As a result of the rule- water reaction permitted by the ECCS perform-making hearing (Docket RM-50-1). the Commis- ance criteria is one percent of the cladding sion adopted regulations dealin g with the mass. Use of this "one percent of the mass" effectiveness of ECCS (10 CFR Part 50, value as a lower limit for assumed hydrogen 9 50.46, " Acceptance Criteria for Emergency production, however, would unnecessarily pe-Core Cooling Systems for Light-Water Nuclear nalize reactors with thicker cladding, since for Power Reactors"). the same thermal conditions in the core in a 1 postulated LOCA, the thicker cladding would i The staff believes it is appropriate to con- not, in fact, lead to increased hydrogen gen-sider the experience obtained from the various eration. This is because the hydrogen genera- <

ECCS-rela ted analytical s t udies- and test tion from metal-water reaction is a surface )

programs, such as code developmental efforts, phenomenon.  !

fuel denstfication, blowdown and core heatup i studies, and the PWR and BWR FLECHT tests, A more approp riate basis for settin g the and to take account of ,he increased con. lower limit would be an amount of hydrogen as-Servatism for plan ts with ECCS evaluated sumed to be generated per unit chdding area.

under 6 50. LG in setting the amount of initial it is convenient to specify for this purpose a metal-water reaction to be assumed for the hypothetical uniform depth of cladding surface purpose of establishing design requirements reaction. The lower limit of metal-water reac-for combustible gas control systems. The staff tion hydrogen to be assumed is then the hypo-has always separated the design bases for thetical amount that would be generated if all ECCS and for containment systems and has the metal to a specified depth in the outside required such contatnment systems as the com- surfaces of the cladding cylinders surrounding bustible gas control system to be designed to the fuel (excluding the cladding surrounding withstand a more degraded condition of the the plenum volume) were to react.

reactor than the ECCS design basis permits.

The approach is consistent with the provisions In selecting a specified depth to be assumed j of General Design Criterion 50 in which the as a lower limit for all reactor designs, the y need to provide safety margins to account for staff has calculated the depth that could cor-the effects of degraded ECCS function is respond to the "one percent of the mass" value no ted . Although the level of degradation con- for the current core design with the thinnest sidered might lead to an assumed extent of cladding. This depth (0.01 times the thickness metal-water reaction in excess of that calcu- of the thinnest fuel cladding is used) is lated for acceptable ECCS performance, it does 0.00023 inch (0.0058 mm).

not lead to a situation involving a total f ailure of the ECCS, in summary, the amount of hydrogen to be i generated by metal-water reaction in determin- j The staff feels that this " overlap" in protec- ing the performance requirements for combus- l tion requirements provides an appropriate and tible gas control systems should be five times prudent safety margin agatns t unpredicted the maximum amount calculated in accordance evenN during the course of accidents. with 3 50.46, but no less than the amount that would result from reaction of all the metal in A ccord tn gly ,' the amount of hydrogen as- the outside surfaces of the cladding cylinders sumed to be generated by metal water reaction surrounding the fuel (excluding the cladding in estabbshing combustible gas control system surrounding the plenum volume) to a depth of performance requirements should be based on 0.00023 inch (0.0058 mm).

the amount calculated in demonstrating com-l pliance with 9 50. 4ts , but should include a It should be noted that the extent of initial margin above that calculated. To obtain this metal-wa ter reaction calculated for the first l m irgin , the assumed amount of hydrogen core of a plant and used as a design basis for '

should be no less than five times that calcu- the hydrogen control system becomes a limiting lated in accordance with $ 50.46. condition for all reload cores in that plan t unless the hydrogen control system is sub-Since the amounts of hydrogen thus deter- sequently modified and reevaluated.

mined may be quite small for many plants (as a result of the other more stringent requirements for ECCS pe rformance in the criteria of The staff beheves that hydro tems in plants receiving opera'ting genlicenses control on sys-S any or :he 4xan m. my t exam.a .n m. Nac pub. the basis of ECCS evaluations under the a amm.nt twm. "In te rim Acceptance Criteria" should continue 1.7-2

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to be designed for the 5 percent initial metal- turbulence, chemical additives in the coolant, e water reaction specified in the origtnal issuance impu rities , and coolant temperature can all of this guide (Safety Guide 7). As operating exert an influence on the gas yields from radi-

./ plants are reevaluated as to ECCS performance oly sis . The hydrogen production rate from under 5 50.46, a change to the new hydrogen corrosion of materials within the contatnment.

control basis enumerated in Table 1 may be e.g. , aluminum depends on the corroston rate.

made by appropriate amendments to the Tech- which in turn depends on such factors as the nical Specifications of the license. For plants containment coolant chemistry, the coolant pH. !

receiving construction permits on the basis of the me' il and coolant temperatures , and the ECCS evaluations under the Interim Acceptance surface area exposed to attack by the coolant.

Criteria, the applicant would have the option of Accurate values of these pa rameters are difficult to establish with certainty for the con-using either a 5 percent initial metal-water reaction or five times the maximum amount ditions expected to prevail followtng a LOC A.

' calculated in accordance with $ 50.46, but no less than the amount that could result from Table 1 defines conservative values and as-reaction of all the metal in the outside surfaces sumptions that may be used to evaluate the of the cladding cylinders surrounding the fuel production of combustible gases following a (excluding the cladding surrounding the ple. LOCA.

num volume) to a depth of 0.00003 inch (0.0058 mm). If these assumptions are used to calculate the concentration of hydrogen (and oxygen) within No assumption as to rate of evolution was as. the con tainmen t structures of reactor plants sociated with the magnitude of the assumed following a LOCA. the hydrogen concentratton metal-water reaction originally given in Safety is calculated to reach the flammable Limit within Guide 7. The metal-water reaction is of signifi- reriods of less than a day af ter the accident cance when establishing system performance tur the smallest containments and up to more requirements for containmen t designs that than a month for the large<t ones. The hydro-employ time-dependent hydrogen control fea- gen concentration could be maintained below its tures. The staff recognizes that it would be lower flammable limit by purging the contain-unrealistic to assume an instantaneous release ment atmosphere to the environs at a controlled of hydrogen from an assumed metal-water reac- rate after the LOCA: however, radioactive lion. For the design of a hydrogen control sys- materials in the containment would also be  ;

tem, therefore , it should be assumed that the released Therefore, purging should not be the '

prunary means for controlling combus tible

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initial metal-water reaction would occur over a short period of time early in the LOCA tran- gases following a LOCA. It is advisable,

(_ stent, i. e. , near the end of the blowdown and however, that the capability for controlled core refill phases of the LOCA transient. Any purgtng be provided to aid in containment at-hydrogen thus evolved would mix with steam mosphere cleanup, and would be rapidly distributed throughout the containmen t compartments enclosing the The Bureau of > lines has conducted experi-i reactor primary coolant system by steam flow- ments at its facilities with initial hydrogen ing from the postulated pipe break. These com- volume concentrations on the order of 4 to 12 l partments include the "drywell" tn typical boil- volume percent. On the basis of these experi-  ;

ing water reactor con tainments , the " lower ments and a review of other reports, the NRC <

volume" of ice conden=er containments, and the staff concludes that a lower flammability limit of I full volume of " dry" containments. The dura. 4 volume percent hydrogen in air or steam-air tion of the blowdown and refill phase is gen- atmospheres is well established and is ade-erally several minutes. Therefore, the assump- quately censervative. For initial concentrations tion of a two-minute evolution time , w hich of hydrogen greater than about 6 volume represents the period of time during which the percent. it is possible in the presence of suffi-maximum heatup occurs, with a constant reac- cient ignition sources that the tutal accumu-tion rate is appropriately conservative for the lated hydrogen could burn in the containment.

design of hydrogen control systems, even with For hydrogen concentrations in the range of 4 the additional assumption that the resulting to 6 volume percent, partial burning of the j hydrogen is uniformly distributed in the con. hydrogen above 4 volume percent may occur, tainment compartment enclosing the primary However, in this range of 4 to 6 percent, the i ,

coolant system. The effects of partial pressure rate of flame propagation is less than the rate i l of steam within the subcompartments and con- of rise of the flammable mixture. Therefore '

.tainment should be considered in the evaluation the flame can p ropa gate upward, but not' of the mixture composition. horizontally or downward. In this case, only a fraction of hydrogen will burn in the contain- -

The rate of production of gases from radioly- ment and complete combustion will not- occur sis of coolant solutions depends on (1) the until the hydrogen concentration is increased

• amount and quality of radiation energy ab- above 6 volume percent. The staff believes that sorbed in the specific coolant solutions used a limit of 6 volume percent would not result in ,

r and (2) the net yield of gases generated from effects that would be adverse to containment l

, .the solutions due to the absorbed radiation systems Applicants or licensees proposing a I

' energy. Factors such as coolant flow rates and design limit in the range of 4 to 6 volume 1

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percent hydrogen should demonstrate through C. REGULATORY POSITION supporting analyses and experimental data that }

. con tainmen t features and safety equipment 1. Each boiling or pressurized light-water required to operate after a LOCA would not be nuclear power reactor fueled with uranium made inoperative bv the partial burning of the xide pellets within cylindrical zircaloy clad-hydrogen .

ding should have the capability to ta) measure the hydrogen concentration tn the containment.

In small containments, the amount of metal- (b) mix the atmosphere in the containment, and water reaction postulated in Table 1 may result (c) c ntrol combustible gas concentrations in h yd rogen concen trations above acceptable without relying on purging and/or repressuri-l l limits. The evolution rate of hydrogen from the zation of the containment atmosphere following, metal-water reaction would be greater than that a LOC A .

from either radiolysis or corrosion. Since it is difficult for a hydrogen control system to pro. 2. The continuous presence of redundant cess large volumes of hydrogen very rapidly, combustible gas control equipment at the site an alternative approach is to operate some of may not be necessary provided it is available the smaller containments with inert (oxygen. on an appropriate time scale. However, appro-deficient) atmos pheres . This measure, the priate design and procedural provisions should

, "inerting" of a containment, provides sufficient be made for its use. These provisions should

, time for combustible gas control systems to include consideration of shielding requirements become effective following n LOCA before a to permit (a) access to the area where the flammable mixture is reached in the contain- mobile combustible gas control system will be; ment. Hydrogen recombiners can process the coupled up and (b) the coupling operation tot l

. containment atmosphere at a limited rate of 100 be executed. In addition, centralized storage j l150 sefm per recombiner. Therefore, an facilities that would serve multiple sites may be I morrimately large number of recombiners would used. provided these facilities include provi- '

be required to centrol the hydrogen concentra. sions such as maintenance, protective features, l l tton that is postulated to be generated in the testing, and transportation for redundant units i first 2 minu tes of the LOCA. There are to a particular site.

presently no other methods of combustible gas control except for purge systems that release 3. Combustible gas control systems and the a

radioactive materials. provisions for mixing, measuring, and sampling should meet the design, quality assurance, re-4 dundancy, energy source, and instrumentation For all containments. it is advisable to pro- requirements for an engineered safety feature.  ! i vide means by which combustible gases result-in g from the postulated metal-water reaction, In addition, the system itself should not intro-duce safety problems that may affect contain-

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radiolysis, and corrosion followtng a LOCA can ment in tegrity . The combustible gas control be mixed, sampled, and controlled without re- system should be designated Seismic Catego-leasing radioactive matertals to the environ- ry I (see Regulatory Guide 1.29, " Seismic De-ment. sign Classification"), and the Group B quality standards of Regulatory Guide 1.26, " Quality Smee any system for combustible gas control Group Classifications and Standards for Water ,

is designated for the protection of the public in Steam , and Radioactive-Waste-Containing Com-the event of an accident, the system should ponents of Nuclear Power Plants," should be meet the design and construction standards of ap plied .

engmeered safety fea t u re s . Care should be taken m its design to ensure that the system water-cooled power reactors should itself does not introduce safety problems that is have the installed capability for a con-may affect containment integrity. For example, trolled purge of the containment atmosphere to tt 1 flame recombiner is used, propagation of aid m cleanup. The purge or ventilation system flame mto the con tainmen t should be pre- may be a separate system or part of an existing l Ven ted , system. It, need not be redundant or be desig- I nated Setsmic Category 1 (see Regulatory '

In most reactor plants, the hydrogen control Guide 1.29), except msofar as portions of the system would not be required to be operated system constitute part of the primary contain-for 7 days or more followmg a postulated de- ment boundary or contam filters.

sign basis LOCA. Thus it is reasonable that 5. The parameter values listed in Table 1 hydrogen control systems need not necessarily should be used in (a) calculating hydrogen and be Lns talled at each reac to r . Provision for oxygen gas concentrations in containments and either onsite or offsite storage or a shared (b) evaluating ~ designs provided to control and arrangement between licensees of plants in to purge combustible gases evolved in the I reasonably close proximity to each other may course of loss-of-coolant accidents. These val-be developed . An example of an acceptable ues may be changed on the basis of additional arrangement would be to provide at least one experimental evidence and analyses.

h yd rogen control system per site with the l proviston that a redundant unit would be avail- 6. Matertals within the containment that able from a nearby site. would yield hydrogen gas due to corrosion from /

1.7-4

the emergency cooling or containment spray applicant or licensee proposes an acceptable j solu tions should be iden tified , and their use alte rnative method of complying with thesej should be limited as much as practical. standards, the method described heretn will be i applied by the NRC staff in accordance with '

s 50 44. Where this may involve addition, ehm-D. IMPLEMENTATION inaticn, or modtfication of structures, systems ,

or components of the facility af ter the con-The applicability of the standards for com- struction permit, manufacturing license, or de-bu s tible gas control systems to existing and sign ap p roval has been issued. backfitting i f utu re facilities is set forth in 6 50.44. decisions will be determined by the staff on a l I Therefore, except in those cases in which the case-by-case basis . l l.7-5

TABLE 1 x ACCEPTABLE ASSUMPTIONS FOR EVALUATING THE PRODUCTION OF COMBUSTIBLE GASES FOLLOWING A LOSS OF. COOLANT ACCIDENT (LOCA)

Parameter Acceptable Value Fraction of fission product radiation a. Beta energy absorbed by the coolant

1. Betas from fission products in the fuel rods: 0

2. Betas from fission products intimately mixed with coolant: 1.0 b- Gamma

1. Gammas from (5sion products in the fuel rods, coolant in core region : 0.l**

2. Gammas from fission products intimately mixed with coolant, all coolant: 1.0 llydrogen yield rate G(ll,; )* 0.5 molecule /100 eV Oxygen yield rate GlO. )* 0.25 molecule /100 eV 3

Extent and evolution time of initial core Hydrogen production is 5 times the extent metal-water reaction hydrogen production of the maximum calculated reaction under from the cladding surrounding the fuel 10 CFR Part 50, g 50.46, or that amount that would be evolved from a core-wide average depth of reaction into the original )

cladding of 0.00023 inch (0.0058 mm), '

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whicheveris greater, in 2 minutes.

Aluminum corrosion rate for aluminum 200 mils /yr. (This value should be adjusted exposed to alkaline solutions upward for higher temperatures early in the accident sequence.)

Fission product distribution model 50% of the halogens and 1% of the solids present in the core are intunately mixed with the coolant water.

All noble gases are released to the con tainmen t .

All other fission products remain tofuel rods.

flydrogen concentration limit 4 v/o***

Oxygen concentration limit 5 v/o. (This limit should nc. oe exceeded if more than 6 v/o hydrogen is present.)

For water, borated water, and borated alkaline solutions; for other solutions, data should be presented.

This fractton is thought to be conservative, further analysis may show that it should be revised .

The 4 v/o hydrogen cone'entration limit should not be exceeded if burning is to be avoided and if more than 5 v/o oxygen is present in the containment. This amount may be increased to 6 v/o, with the assumption that the 2 v/o excess hydrogen would burn in the containment (tf more than 5 v/o oxygen is present). The effects of the resultant energy and burning should not create conditions exceeding the design conditions of either the containment or the safety equipment necessary to mitigate the consequences of a LOCA. Applicants and licensees '

should demonstrate such capabthty by suitable analyses and qualification test results. -

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l n- g 7 1973 VALUE/ IMPACT STATEMENT ON THE EFFECTIVE 3 50.44 STANDARDS FOR CCMSUSTISLE GAS CCffi'ROL SYSTEMS .iND -

THE REVISION 2 0F THE REGULATORY GUIDE 1.7 I. The Procosed Action A. Descriotion ,

The proposed action is to clarify the Ccmmission's original intent regarding the degraded Emergency Core C: cling System (ECCS) functioning contained in Appendix A, General Design Criterion 50, " Containment design basis," to provide clear stand- l i

ards for combustible gas centrol systems, and to resolve a l l

complex policy and technical issue by way of rulemaking. I B. Need for the Procesed Action The present wording of Criterion 50 states that the containmen:

design must have sufficient margin to accommeda:e, withcut exceeding the design leakage rate, calculated pressure and temperature conditions resulting from any less-of-ccciant acciden:

(LOCA). This margin must reflect, accng other considerations ,

the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generaters and energy frem metal-water and other chemical reactions that may result frem degraded emergency core cooling functioning.

1 Encicsure C

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For the' purpose of containment design, the Commission has assumed

, that degraded emergency core cooling functioning may ccour beyond the amergency core cooling acceptance criteria but not to the point of total failure. The assumed performance degradation could result in localized fuel element failure frem such causes as localized ficw blockage, localized hot spots, popped cans and local distortion in core geometry. However, the degraded ECC functioning has been subject to various interpretation in the related industry. There is an apparent need to clarify cur position on this issue.

On the other hand, the design of ccmbustible gas control systems requires evaluations of hydrogen generated frcm all sources.

Among them the extent of metal-water reaction and associated hydrogen production depends strongly en the course of, events assumed for the accident and on the effectiveness of emergency cooling systems. Recent development on the effectiveness of

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ECCS performance resulted in a rulemaking (10 CFR Part 50,!50.46) that made a number of changes to the previous c:nservative modell . These changes have increased the margin o'f 5005 evalua-tion model should a design basis LOCA ever occur. The increased conservatism of 150.46 would limit the fuel duty, with the result that calculated metal-water reaction rates would be significantly reduced. Additionally it is appropriate to consider 2 . Enclosure C

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the expe'rience obtained from the various ECCS-related analytical studies and test programs, such as code developmental efforts, fuel densification, blowdown and core heat-up studies, and the PWR and SWR FLECHT tests, and to take account of the increased conservatism for plants with ECCS evaluated under 5 50.46, in setting the amount of initial metal-water reaction to be assumed for the purpose of establishing design requirements for combustible gas control systems.

For these reasons, it is proposed that the combustible gas control system standards be relaxed for plants operating under 1 50.46. To resolve this complex technical issue, a definite need exists for a new regulation as well as revisions to Regulatory Guide 1.7 for combustible gas control systems.

C. Value/Imaact of the Dracosed Action It should be noted that the following value/ impact assessment

. is prepared based on the current staff technical position.

1. NRC Coerations No negative impact will be expected for new clants in terms of NRC time and resources required for implementation, licensing review, inspection and enforcement, since the proposed action is involved only with revisions of technical data, not the methodolo~gy. For older plants (primarily BWRs with Mark I and Mark II containment), a reevaluation 3 Enclosure C 7-- . _

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l of the metal-water reaction and. hydrogen concentration must be performed to determine whether de-inerting of those inerted containments is permitted. However, these reviews and 1 4

reevaluatiens have been done with the exception of Peach Bottom 2 & 3 which are pending on the submittal of 10 CFR 50.46 ECCS analyses.2 In summary, there would be no significant impact on NRC operations except (1) that the proposed action will clarify NRC positions on the inerting requirement and )

(2) that the proposed rule will resolve a complex policy and

, l technical issue to expedite future licensing review precesses.

2. Other Government Acencies Not applicable unless the goverr. ment agency is an applicant

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as TVA.

3. Industry The proposed action has little or no effect upon ?WRs and BWRs with Mark III containment, while for EWRs with Mark I and Mark II contairment, the propcsed changes in the me:al-water reaction calculatiens wculd greatly reduce the e number of inerted containment and even pessibly to none.

Excluding Peach Bottem 2 & 3 which are pending en the sub-mission and review of 10 CFR 50.46 ECCS analyses, all of the other 34 plants listed in Table 1 3 could be de-inerted.

4 Enclosure C A

As a result, an annual saving of approximately $12,000,000 could be expected by the industry. In addition, restrictions imposed by inerting on inspections and maintenance works could be removed. The potential hazard associated with transportation and storage of liquid nitregen could be minimized. ,

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4. public As a result of the proposed changes, inerting is being i eliminated in most cases with the possible exception of peach Bottcm 2 & 3. In these cases, the level of public safety is maintained because the safety margin afforded by inerting is more than ccmpensated for by the more restrictive

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limitatians placed on operating conditions of the fuel.

C. Decision on the Precosed Action j l

Guidance should be furnished on standards for c:mbustible gas

. control systems. ,

1 II. Technical Accreach A. Technical Alternatives Because the proposed action has not changed the actual methodology, there is no need to review new technical alternatives. The preposed values fer the parameters set forth in the proposed 5 Enclosure 0

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action are based on conservative staff's judgements and may be revised if further study show that other values are more appropriate.

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8. Discussion and Cemcarisen of Current Technical Alternatives Inerting the centainment is the only practical mean for appropriate hydrogen c' .itrols.during a very short time period after a LOCA, if this is needed, withcut releasing radioactive materials to the environment.

For long time periods after the LOCA, usually in several days, two methods have been proposed which are acceptable to the i

l staff: containment air dilutien systems (CAD) and recembiner systems. .

l CAD systems, 3:ch censist of a repressuri::atien system and a

! purge system, are to repressuri::e by pumping air or inert gas

! into the containment until the pressure reaches 50". of the design value, folicwing which the centainmen: is purged through an appropriate filtering system. Mcwever, this system would only be used for older plants because of the ccmplexity of backfitting other systems such as a reccmeiner which would require a larger penetration.

For new plants it is advisable to employ a closed system such as a reccmbiner which is able to remove the hydrogen prcduced frem several sources without purging the containment. Hewever, i l

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Enclosure C e

it is advisable'to require the capability of controlled purging as an aid in cleanup of containment atmosphere. Although re-combiners should be designed as engineered safety features, it would not necessarily be so for the purging system.

C. Decision on Technical Accroach For older plants, containment air dilution (CAD) systems, i.e. ,

repressurization and purging, are the most appropriate means for hydrogen control with or without inerting. Hewever, a limit on the incremental dose due to purging should be established in order that the total doses after the LOCA remain below the .

limits set forth in 10 CFR 100.

For new plants, recombiners are closed systems which remove the hydrogen frem the containment without significant radioactive material release. A controlled purge system is required as a back-up system.

III. Procedural Accroach A. Procedural Alternatives Potential SD procedures that may be used to promulgate the proposed action and technical apercach include the folicwing:

Regulation Regulatory Guide 7 Enclosure C 3 .

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ANSI Standard, endorsed by a Regulatcry Guide

- Branch position NUREG B. Value/Imoact of Precedural Alternatives It.was a Commission's decision to resolve the technical problem by the way of rulemaking process. To comply with the proposed regulation, revision of the' former Safety Guide 7 is thereby undertaken.

1 C. Decision on Precedural Acercach As a result, a new paragraph, 3 50.44, has been prepared to add to 10 CFR part 50. Revisions of the Regulatory Guide 1.7 have )

also been prepared.

1 IV. Statutory Considerations ,

A. NRC Authority

• The new 5 50.44 paragraph wculd fall under the au:hori y and safety requirements of the code of Federal Regulatienr.,

Tit!e 10 ENERGY and in particular, un' der General Cesign Cri:eria,

' Appendix A of 10 CFR 50 which requires, in part, that the effects of potential energy sources which have not been included in the determination of pressure and temperature conditions shall be included for the containm .it design. The new ! 50.44 wculd require the consideration of the energies fr m metal-water and 8 Enclosure C

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Other chemical reactions that may result frem degradation, but not total failure of, the emergency core ecoling functioning.

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8. Need for NEpA Assessment

.The proposed action is not a major action, as defined by 10 CFR Sl.5(a)(10), and does not require an environmental impact statement. .

V. Relationshio to Other Existinc or procesed Regulatiens or Policies No potential conflicts or overlaps with other agencies, e.g. , EPA, CE, FPC, FEA, 00T, that may have parallel or opposing requirements to the proposed action are kncwn or expected.

VI. Summary and Conclusions As a consequence of the Commissions decision of resolving the technical problem concerning the combustible gas control after a LOCA by way cf the rulemaking process, a new paragraph 5 50.14 to be incorporated to 10 CFR 50 has been prepared. A revision of Regulatory Guide 1.7 has accordingly been prepared. It is concluded that a sufficient and appropriate safety margin is posed with these actions. ,

9 Enclosure C

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REFERENCES

1. Interim Acceptance Criteria for Emergency Core Cooling Systems *

. published June 29,1971, (36 FR 12248).

2. Memorandum to V. W. Panciera from S. Buckley/C. Grimes, "Deinerting of Operating SWR Containments," June 7,1977.

3. Enclosure "F", "Value Impact Statement," the Ccomissioners' paper on proposed amendments to 10 CFR Part 50, June 25,1976.

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10 Enclosure C A, -