ML25196A392

04-16-76 Status of Generic Items Relating to Light-Water Reactors-Report No. 4

ML25196A392
Person / Time
Issue date:	04/16/1976
From:	Moeller D Advisory Committee on Reactor Safeguards
To:	Rowden M NRC/Chairman
References
Download: ML25196A392 (1)

Text

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION Honorable Marcus A. R:>wden Acting Chairman WASHINGTON, D. C. 20555 AprH 16, 1976

u. S. Nuclear Regulatory Conmission Washington, OC 20555

Subject:

STA'IUS. OF GENERIC ITf.HS REIATJ:K; 'ID LIGll'l'-WMER RFAC'IDRS:

REPORl' NO. 4

Dear Mr. R:

>wden:

'lhe Advisory Corrmittee on Reactor safeguards reported on the "Status of Generic Items Relating to Light-water Reactors* in its letters of December 18, 1972, February 13, 1974, and March 12, 1975.

'Ibis is the fourth such report. Since the Coomittee limits its definitions of generic items to those cited specifically in its letters pertaining to projects and related matters, the attached listing is not all-inclusive, the Nuclear Regulatory Conmission Staff list bas additional generic items.

Group I of the attachment is a reiteration of the generic items oonsidered resolved at the time the Conmittee issued its first report. Group IA includes those items resolved between December 1972 and February 1974; Qoup IB includes those items resolved between February 1974 and March 1975; Qoup IC includes those items resolved since March 1975. Following each res()lved item is a brief statement of the specific action that resulted in the reso-lution. Group II lists those items included in the original report for which resolution on a generic basis is still pending. Groups IIA and IIB include generic items that were added in the second and third reports, Group IIC inclmes those added in the present report. 'lhe ACRS and the NIC Staff will continue to consider the safety significance of ~oup II, IIA, IIB and IIC items on a case-by-case basis tmtil generic resolution is reached. Formal actions such as issuance of Regulations or Regulatory Guides are anticipated for many of the Group II, IIA, IIB and IIC items.

2240

lbnorable Marcus A. R:>wden April 16, 1976

'Ihe NRC Staff considers several of the items on the attached tD1resolved lists to be "resolved*, based on specific positions taken by them; however, these "resolutions" do not meet the ACRS criteria for adequate doct.111entation and for COlrmittee awroval so they are considered to be pending mtil formally discussed and accepted by the ACRS.

It:ell5 in the above category are:

(1)

II-2 "Effective Cperation of Containment Sprays in a IJXA*;

(2)

II-4 "Instruments to Detect Fuel Failures";

(3) rr-6 "CC>nloon l-'bde Failures*;

(4) rr-9 "'!he Advisability of Seismic Scram*;

(5)

IIA-1 "Pressure in Containment Following IJXA";

(6)

IIA-2 "Control a::>d Drop Accidents (BIR)";

(7)

IIA-5 "Rupture of High Pressure Lines Oltside Containment*;

(8)

IIA-8 "Isolation of IDw Pressure fran High Pressure Systems";

(9)

IIB-3 "Behavior of BWR Mark III Containments."

Owing to questions raised concerning the scope and intent of various generic issues, the Ccmnittee has incorporated into the attachnents a brief description for all items mresolved now or at the time of the third report.

"Resolved" as used in the Generic Items reports refers to the following:

In some cases an item has been resolved in an administrative sense, recognizing that technical evaluation and satisfactory implementation are yet to be canpleted. Anticipated Transients Without Scram represents an example of this category. In other instances, the resolution has been accanplished in a narrow or specific sense, recognizing that further steps 2241

Honorable Marcus A. R:>\\<rlen April 16, 1976 are desirable as practical or that different aspects of the problem require further investigation. Examples are the possibility of improved methods of location of leaks in the primary system, and of improved methods or augmented scope to inservice inspection of reactor pressure vessels.

Attachments:

1)

Group I

2)

Group IA

3)

Group IB

4)

Group IC

5)

Group II

6)

Group IIA

7)

Group IIB

8)

Group IIC Sincerely yours, Dade w. M:>eller Olainnan 2242

GENERIC I'1D1S Group I - Resolved Generic Items

1. Net R:>sitive Suction Head for occs Pllllps:

covered by Regulatory Guide 1.1.

2.

Emergency R:>wer : Covered by Regulatory Guides 1. 6, 1. 9, and 1. 32 and portions of IEEE-308 {1971).

3.

Hydrogen Control After a IDss-of-COOlant Accident {U£A):

N:RS concurred in proposed Staff position, covered by me Standard Review Plan for Nuclear R:>wer Plants.

4.

Instrument Lines Penetrating Containment:

Covered by Regulatory Guide 1.11 and Supplement.

5.

Strong M:>tion Seismic InstrllTlentation: Covered by Regulatory Guide 1.12.

6. Fuel Storage R:>ol Design Bases: Covered by Regulatory Guide 1.13.

7. Protection of Primary System and Fngineered Safety Features l!gainst PUmp Flywheel Missiles: covered by Regulatory Guide 1.14.

8. Protection l!gainst Industrial Sabotage: Covered by Regulatory Guide 1.17.

9. Vibration ftt>nitoring of Reactor Internals and Primary System:

covered by Regulatory Guide 1.20.

10.

Inservice Inspection of Reactor Coolant Pressure Boundary: covered by ASME lk>il~r and Pressure vessel {BPV) COde,Section XI and Regulatory Guide 1.65.

11. QJality Assurance niring Design, Construction and ~ration:

covered by 10 CFR so, Appendix B; ASME BPV COde,Section III; ANSI N-45.2-1971, Regulatory Guides 1.28, 1.33, 1.64, 1.70.6 and Proposed Standard ANS-3.2.

12\\. Inspection of BWR Steam Lines Beyond Isolation valves: Covered by ASME BP\\T COde,Section XI.

13.

Independent Oleck of Primary System Stress Analysis: covered by ASME BP\\T COde,Section III.

14. ~rational Stability of Jet PlinpS: ~st and operating experience at Dresden 2 and 3 and other jet punp BNRs have satisfied the ACRS concerns.

2243

Group I Continued

15. Pressure vessel Surveillance of Fluence and N!1.l' Slift: Covered by 10 CFR SO,.Appendix A and Appendix H; and AS'ffl Standard E-185.

16. Nil n.ictility Properties of Pressure vessel Materials: Covered by 10 CFR so,.Appendix A and Appendix G: ASME BPV Code, Section III:

ACRS Pressure vessel Report.

17. Operation of Reactor With less 'Ihan All IJ:x>ps In Service: Covered by ACRS-Regulatory Staff position that manual resetting of several set points on the control roan instnunents under specific conditions and procedures is acceptable in taking one primary loop out of service.

'!his position is based on the expectation that this mode of operation will be infrequent.

18. Criteria for Preoperational 'lesting: Covered by Regulatory Guide 1.68.

19. Diesel Fuel capacity: Covered by ACRS-Regulatory Staff position requiring 7 days fuel.

20. capability of Biological Shield Withstanding D:>uble-Ehded Pipe Break at Safe Ends: Covered by ACRS-Regulatory Staff position cited in several letters that such a failure should have no unacceptable consequences.

21. Operating 01.e Plant While Other (s} is/are U'lder Construction:

Specific requirements have been established by ACRS-Regulatory Staff.

R>sition will be prepared.

22.

Seismic Design of Steam Lines: Covered by Regulatory Guide 1.29.

23. ~ality Group Classifications for Pressure Retaining Components:

Covered by Regulatory Guide 1.26.

24.

Ul ti.mate Heat Sink: Covered by Regulatory Guide 1. 27.

25.

Instnunentation to Detect Stresses in Containment Walls: Covered by Regulatory Guide 1.18.

2244

Group IA - Generic Items Resolved Since December 18, 1972 I.

USe of Furnace sensitized Stainless Steel: Covered by Regulatory Guide 1.44.

2. Primary System Detection and Iocation of leaks: Covered by J.egulatory Guide 1. 45.

3. Protection.Against Pipe l'flip: Covered by Regulatory Guide 1.46.

4. Anticipated Transients Without Scram: Covered by Regulatory R>sition

Dxmnent, 0 Technical Report on Anticipated Transients Without Scram for water-cooled R:>wer Reactors," WASB-1270, september 1973.

s.

OCCS Capability of Current and dl.der Plants: Covered by Rulemaking as a general policy decision, although acceptable detailed implementation remains to be developed. J))cket RM-50-1, 0 Acceptance Criteria for Emergency Core Cooling Systems for Light-water-Cooled-Nuclear R>wer Reactors, 0 December 28, 1973.

2245

Group IB - Generic Items ~solved Since February 13, 1974

1.

Positive ftt>derator Coefficient:

IWRs presently have or expect to have zero or negative coefficients. 'illere sane Technical Specifications allow a slightly positive coefficient, the accident and stability analyses take this into acco1.mt.

&lrnable poison provisions have been designed into IWRs to reduce otherwise excessive positive coefficients to allowable values.

2.

Fixed Incore Detectors on High R:>wer IWRs:

Fixed incore detectors are not required for IWRs since reviews of potential p:,wer distribution ananalies have not revealed a clear need for continuous incore monitoring.

3.

Performance of Critical Cc7nponents (Pl,111PS, cables, etc.) in post-IDCA Ehvironnent: Qualification requirements of critical caip>nents are now covered by Regulatory Guides 1.40, 1.63, 1. 73 and 1.89 and IEEE standards 382-1972, 383-1974, 317-1972, 323-1974.

4.

vacumt ~lief valves Controlling Bypass Paths on BNR Pressure Suppression Contaimlents: 01 designs prior to GE Mark III oon-taimnent, resolution lies in surveillance and testing of vacut1n relief valves. For Mark III containments, an additional require-ment is that the design be capable of accamDdating a bypass equivalent to one square foot for a given flow conditiOli.

5.

Emergency Power for '1W or M:>re ~actors at the same Site: ~solved by issue of Regulatory Guide 1.81.

6.

Effluents fran Light-water-Cooled-Nuclear R:Jwer ~actors: ~solved by issue of Appendix I to 10 CFR 50.

7.

Control R>d Ejection Accident: ~solved for IWRs by Regulatory Guide 1. 77.

2246

Group IC - Generic Items Resolved Since March 12, 1975

1.

Main Steam Isolation valve leakage of BWR's:

Covered by Regulatory Guide 1.96.

2.

.niel Densification: covered by 10 CPR 50 Appendix K plus case--by-case review of vendor fuel DDdels.

3.

Rx! sequence control Systems: covered by me Staff Review and Approval of ~10527 and Presentation to ACRS.

4.

seismic category I ~irements for Auxiliary Systems:

Covered by Regulatory Guides 1.26 and 1.29.

2247

IC-I -

MAIN S'l'FAM ISOIATICII VALVE LF.AKAGE CF BNR;

'!he BWR main steam isolation valve (MSIV) leakage problem relates to a loss of coolant accident condition where radioactivity levels are postulated to exist in the coolant. 'lhe anticipated leakage levels in this class of valve may result in excessive releases of activity to the environment.

R:>ssible solutions include another valve in series to decrease leakage to acceptable levels, sealing mechanisn.s such as water seals to trap and retain radioactivity, or another type of MSIV capable of meeting the low levels of leakage. 'lhis issue has been resolved through issuance of Begulatory Guide 1.96.

2248

IC FUEL DmSIFICATim Fuel Densification is a facet of Behavior of Reactor Fuel Older Al:xlomal.

Conditions (II-7) and OCCS capability of CUrrent and Older Plants (IA-5).

'!he densification of fuel changes fuel-cladding gaps, contained sensible heat, and fuel temperatures, leading to tmpredictable fuel behavior throughout life. Newer fuel has been JOOdified in seveal ways to minimize densification, particularly with regard to increasing initial densities of fuel pellets. Existing results indicate that these changes have eliminated densification as a problem. '!his issue is considered resolved on the joint bases of conformance to Appendix K of 10 CFR 50 and case-by-case reviews of vendor fuel JOOdels.

2249

IC ROO S~ cnnroL SYS'JUtS lbJ sequence control systems or rod pattern control systems are designed to prevent a control rod pattern to exist where a control rod accident could result in peak fuel enthalpies in excess of 28() calories/gram for the entire raDJe of plant operations and oore exposure.

'!be problem is one of inadequate experience or analytic evaluation to confirm the design conservatism. 'lbpical reports have been sli:mitted and NlC reviews are ca11?lete, so the issue is considered resolved.

2250

IC SEISMIC CA'ffXDRY I~

POR AUXILIARY SYSTEMS various auxiliary systems provide continuous or intermittent functions insofar as control of radioactivity, safe plant operation including startup and shutdown, or essential safety actions under accident oondi tions.

'lhese systens are designed to various codes and standards and nay or nay not be seismic category I. 'lbese auxiliary systems have been evaluated on the basis of such factors as potential release of retained radioactivity, disruption of operation, and failure to provide vital safety functions to determine what canponents, if any, need to be designed and constructed to meet seismic category requirements. 'lbese factors have been incorporated into Regulatory Guides 1.26 and 1.29.

2251

Group II - Resolution Pending

1.

'1\\Jrbine Missiles: '.l'\\Jrbine failures for past 16 years have been evaluated and a statistical probability analysis has been ocmpleted.

An ACRS letter (April 18, 1973) discusses the problems.*

2.

Effective ~ration of Containnent Sprays in a IOCA:

Extensive documentation in topical reports. Review and evaluation are required.

3.

i:ossible Failure of Pressure ~

R>st-IOCA By 'lhennal Shock:

IeJulatory Guide 1.2 covers current infor:mation. Ultimate position as to significance of thermal shock requires input of fracture mechanics data on irradiated steels fran the Heavy Sectian Steel Technology Program.

4.

Instruments to Detect Fuel Failures: Instrumentation exists to detect fuel failures. Continuing work is required..

s.

R>nitoring for Excessive Vibration or IDose Parts Inside the Pressure vessel: State-of-the-Art results appear pranising. lt>re work may be required prior to decisian as to installation of equipnent.

6.

Cl:mll>n ~

Failures: Baquirements for diverse caupouents should be established.

7.

Behavior of Reactor Fuel 'Older Abnormal Conditians: 'Jhis includes:

flow blockage1 partial melting of fuel assemblies as it affects reactor safety1 and transient effects on fuel integrity. 'Jhe PBF program will address sane of these items.

8.

BJR R:!circulation Pllip OVerspeed Dlring IOCA:

Decision required by ACRS-Regulatory Staff.

9.

'.Ibe.Advisability of seismic scram: Further stmies required to establish need.

10. F.mergency Core Cooling System capability for Future Plants: Partially resolved by amendnents to 10 CFR 50 [50.34(a) (4), 50.34(b) (4), 50.46, and ~ix K].

IOCA evaluation IIX>del OC11Plete.

ACRS feels new cooling awroaches should be explored.

11. Instrumentation to Follow the Course of an Accident: A leJulatory Glide to be issued should resolve the issue.

• Regulatory Glide ism preparation.

2252

II TURBINE MISSIU:S

'I\\trbine failures for the past 16 years have been evaluated and a statistical probability analysis has been ooopleted.

An ACRS letter (April 18, 1973) discuses the problem.

'lhree issues require answers to resolve the turbine missile problem:

(1)

'!he first relates to the ar;p-opriate failure probability value;

-4 based on historical failures the probability is about 10.

Industry predicts a much lower failure probability based on improvements in materials and design.

'lb date the ACRS has accepted the 100re conservative value; ( 2)

'!he second issue is strongly dependent on turbine orienta-tion with respect to critical safety structures. strike probabilities fran high angle missiles are acceptably low for single mits and may be acceptable for multi-lD'lit plants, depending on plant layout; however, lower angle missiles with ncn-optiml.lD (tangential) turbine orientation have unacceptably high strike probabilities; (3)

'lbe third issue is one of penetration and damage of structures housed in the oontainnent. 'lbe limited experimental data pertaining to penetration of large irregularly shaped missiles are not sufficient to determine structural response to impingement of turbine disc segments. Jt>st missile penetration formulas are not relevant to this case. sane experiments with irregular missiles might resolve this issue, particularly for older plants with ncn-optinun turbine orientations.

2253

II BFFECJ.'lvE OPERATIOO CF CCNmJ>>1Em' SPRAYS IN A 1/XA Review and evaluation are required of the variety of experiments which have been conducted on the effectiveness of various oontainnent sprays on the removal and retention of airborne radioactive materials anticipated to be present within contaimtent following a IOCA.

such review should consider adequacy or definition of the piysical and chemical forms of the anticipated airborne radionuclides, and quality or evaluative tests of the removal efficiencies of various sprays Older the conditions of temperature, pressure, and radiation doses expected to exist under IOCA conditions. A desirable extension might be benefit/risk analyses of the use of sprays containing chemicals (steh as NaOO) which have the potential for damaging equipnent within oontaimtent. Stones usi.BJ other spray additives, such as hydrazine, have been conducted. If canpounds, sooh as this, have distinct advantages, action should be taken to encourage their use.

2254

II POOSIBLE PAIUJRE CF PRESSURE VESSEL ~WCA BY "l'BEllW, sane Farlier nuclear reactor pressure vessels slt>jected to fluences of 19 1-4 x 10 nvt, which are anticipated in the last 20 years of a 4()-year life, may suffer severe radiation damage denoted by pronotmeed shift in impact transition temperature at the inner surface. 'lhere will be a damage gradient which decreases sharply, so that the properties halflay through the wall are essentially those of the as-fabricated material.

If a IOCA occurs near end-of-life, the injection of cold water on the region of degraded properties may initiate and propagate a crack because of high local stresses near the surface. Analytic procedures indicate the stresses drop rapidly with distance through the wall so the flaw should not propagate beyond sane limiting point.

'lhe lack of experimental evidence and the relative width of the error band in the analytic results are such that sane experiments are required to validate the analytic model. 'Ibese are planned under the HSS'l' progran.

2255

11 ms'l'1Umfl'S '10 IE'l'fCJ.' PUEL FAIUJRBS In the event of local fuel overbeatinj that leads to clad failure and to melting of sections of one or mre irradiated fuel elanents, relatively large ano\\D'lts of fissi011 products could be rapidly released to the reactor coolant. Although the course of events sd:>sequent to su:h fuel melting may not have serious safety significance, there bas, as yet, been no experiemental research progran to investigate such situations and there is little applicable reactor experience. Also, the sd:>sequent course of events may depend, in part, on the initiating cause of fuel overbeati.DJ.

Farly warninJ and timely response may avert an incident tecxning an accident.

Instrumentation related to such diagnostic purposes is being used 011 a,st power reactors. A stmy to define the clH)ropriate mininun requirements and judicious responses to various signals is needed.

2256

II ~rromg; FOR EXCESSIVE VIBRATICB 00 UD;E PAR1'S INSIDE 'l'BE PRESSURE VESSEL IDose parts monitoring can provide early warning of potential mechanical problems or failures within the pressure vessel and throughout the primary coolant circuit. Ieactor vendors have developed mnitoring systans; however, general requirements remain to be established.

2257

11 NCfi-RANI01 IIJLTIPIE -FAIWRES (F'OIIERLY "CXIIDf KIE PAIUJRB")

'lhe term *<XIIIIDll nDde failures* has, in many instances cane to aean multiple failures of identical caapx..ents exposed to identical or near1y identical conditions or environnents, and the use of diversity in canponents has been proposed or required to avoid sldl failures. Die concem of the.ACRS is better expressed by the term *nan-randcn ml.tiple failures*, which is intended to incltxle not anly the type of "<<11+1'\\ lll0de failure" discussed above but other types of multiple failures for which the consequences and probabilities cannot be predicted by application of the single-failure criterion. F.Xantples incltxle the use of the sane sensors or oanponents for both control and protection systems (a reso1ved matter); sequential ml.tiple failures due to a *domino effect*, and simultaneous multiple failures due to a single fault. Since designs usually do not knowingly incorporate features susceptible to S1X:h failures, techniques and criteria need to be developed to detect and avoid them in all systems important to safety.

2258

II BEHAVIOR OF RF.AC'lOR FUEL CH>ER AIHlRMAL CCH>ITI(R;

'!he Behavior of Reactor Fuel tmder Amormal COnditions is still considered llllresolved due to the limited experimental data available.

Partial melti.n:J of fuel assemblies due to flow blockage might lead to autocatalytic effects leadi.n:J to mre extensive fuel failure, pressure pulses, etc. Similar behavior might occur in the case of reactivity transients. '!he ACRS encourages analytic mdeli.n:J but believes appropriate experimental data are necessary. It is anticipated that tests in the Iower Burst Facility (:EBF) should supply much of the required data.

2259

II BNR J?O<<> OJE1S>Em OORDC A IOCA It is p:,ssible for a BNR recirculation Pll1'? to overspeed if a large break occurs at the awropriate p:>5ition in specific pipinJ. O>n-servative estimates indicate substantial overspeed and p:>ssible failure of canponents with the generation of missiles. 'Jhe problem is beinl aa:,roached analytically and experimentally with scaled pmps. 'Jhe reliabiity of such protective measures as the use of decouplers between pmip and DDtor is under stmy.

2260

II '1BE ADVISABILITY CF SEISMIC SCRAM

'!he ACRS has reocmnended that sb.x:lies be made of techniques for seismic scram and of the potential safety advantages and potential disadvantages of pranpt reactor scram in the event of strong seismic mtion, say more than one-half the safe shutdown earthquake.

various suitable techniques have been identified and exist, but thus far only liinited stmies have been reported on the pros and cons of seismic scram. '!he principal potential advantage identified arises fran the greatly improved coolability of a core in the mlikely event of a seismically induced IOCA, smuld scram precede the IOCA by several seconds. A principal reason given in OEP)Sition to seismic scram relates to a stated interest in keeping power stations on the line to provide power offsite should a severe earthquake occur.

2261

II ECCS CAPABILl'l'Y FOR FU'lURE PIAN'l'S

'!he ACRS has placed considerable eq>basis on :EX:'CS safety R&D so that the extent of the conservatism in the :EX:'CS licensing requirements could be made roore precise. With roore experimental data a realistic and quantitative awraisal of ECCS systems 1110uld lead to valid judgments on the changes in licensing which could be put on a firm basis.

Parallel approaches that seek to improve the reliability of ECCS systems, to improve the roonitoring of low power peaking, and to inprove those fuel assembly designs which lower peaking factors are encouraged. Further, changes in plant design which improve the reflooding of the reactor core should be sought and evaluated.

R&D efforts on analysis of core blowdown and reflood should be increased and canbined with the results of the standard problems and the associated experiments.

Improved analytical methods 1110uld provide a basis for optimized ECCS.

2262

II IN.5TRJMENTS 'ID FOLIOJ 'DIE CXXJRSE CF AN ACCIDENT Instnnnentation for determining the nature and the course of potentially serious accidents, on a time scale that will permit appropriate emergency action, should be provided at power plants and appropriate calibration methods and calculated bases for interpreting instr1.1t1ent responses should be available. '!be diversity of the installed instrl.lllentation should be adequate to assure that assessments can be made of accidents and appropriate action taken if they follow the predicted course.

In addition, the raoJe of the instrumentation should be adequate to cover various potential radioactivity releases, particularly the t.g;>er range limits so as to fully encanpass the spectra of all possible accidents.

2263

Group IIA - Resolution PendiBJ - Itans Since Decanber 18, 1972

1. Pressure in Containnent FollowiBJ IOCA:

PU.rtber criteria and methods are needed to better evaluate local dynamic pressures in a UlCA to establish :mre definitive design margins.

2. Control R:>d Ikop Accident (BIIRS):

calculations indicate that the reactivity response differs fran earlier values.

New analyses are required, includiBJ three-dimensional effects.

3. Ice Condenser Containments: Additional analyses are required to establish response duriBJ a IOC.A, and to establish design margins.

4.

lhlpture of High Pressure Lines OJtside Containnent: "lhe possibility exists that failure of a high pressure line such as a steam pipe can prevent operation of critical safety catp:>nents.

s.

RiR runp Overspeed ruriBj a IreA: Problem arises in similar manner to that of BWRs ( Item 8 Group II).

6. Isolation of IDw Pressure Fran High Pressure Systems: Jlssurance required that 1ow*pressure systans cannot inadvertently be inter-connected with a high pressure system leadiBj to failure. "lhere are potential interaction problems between Class 1 and Class 2 or Class 3 pressure connections.

7. steam Generator 'l'llbe leakage: Partially resolved by issue of Regulatory Guide 1.83 which addresses the ccnoem fran a pre-ventative point of view.

a..ACRS/NIC Periodic l~Year Review of all :RJwer :Reactors: A nore effective, continuous alternative ag;,roach to periodic reviews is bei.BJ proposed. Pendi.BJ ACRS review, this item is still considered tmresolved.

2264

IIA PRESSURE IN ~

POLIOml; UlCA A potential problem is the overpressurization of sd:x:arpartments following a ux::A.

A suitably conservative analytic pressure model should be developed to determine that adequate safety margins exist with regard to subcanpartment loads.

2265

IIA COll'H)L IO> DIOP ACCDDT (llms)

Sane mcertainties have arisen in previous calculations of this postulated accident, inclooing the choice of negative reactivity insertion rate due t.o scram and the potential differences between a two dimensional and a three dimensional calculation. Particularly for the latter point, more precise theoretical amparisons may be required t.o resolve the matter although probabilistic oonsiderationss may be relevant.

2266

IIA ICE <XHENSER <DmU1fiJft'S

'lbe ice condenser 0011tainnents have sd:>stantially smaller volme an the asst.nption that the ice will ccncJense the steam during a UXA, thus preventing systen overpressurization. 'Jhe rate of ocndensatian is critical in the initial stages of the blOldown and is influenced by interaction of vap:,r with the ice. If the current analyses prove that the condensation nlldel is suitably canservative, the problem ney be resolved.

2267

IIA RUP1'URE CF HIGH PRF.SStJRE LINES 00.l'SIDE CXlrrADl'1EN'l'

'lhe fmctional or structural integrity of CXIIP)nents required for safe shut.down and the maintenance of cold shutdown may be endaDjered by a failure of high pressure pipin:J at certain locations outside of containnent. Specific design and sin:Jle failure criteria need to be developed to resolve this problem.

2268

IIA IWR PCltP 0JERSPEED IIJRDE A ux:A It is possible for a lWR primary coolant pmp to overspeed if a large break occurs at the clR)ropriate positian in specific piping.

Conservative estimates indicate s1.mtantial overspeecl and possible failure of canponents such as flywieel.s with the generation of missiles. 'lbe problem is being approached analytically and experimentally with scaled pmp. ibe reliability of such protective measures as electrical braking of the pmp motor is amder stmy.

2269

IIA ISOIM'ICB CF IDll PRESSURB PR:ft HIGH PRBSSORE St&W Assurance is required that low pressure systans cannot inadvertently be interconnected with a high pressure system leading to a failure of the low pressure systems due to overpressurizatian. a>th Class 2 systans required for EX:CS and Class 3 systans used for nany auxiliary coolant systems provide vital fmictions during nomal and accident conditions.

'/my conditions leading to interocnnectian could trigger a ux=A. It is recognized that these systems must be interccnnected because of the functions they meet. 'lberefore, particular attention is required in valve reliability and reliability of valve actuatiDj circuits to assure that these valves will not open while the primary system is at pressure.

2270

IIA STFAM GENERMtR 'l'llBE LEAKAGE

~rrnally the stean generator is not a critical oarp:>nent durinJ a IOCA-EX:cS.

R:>wever, a special case exists where the stean generator tubes have been degraded due to corrosion, wastage, etc. If the shock loads imposed by the IOCA cause a critical nt.lllber of td:>es to fail, say by a double-ended (guillotine) break the inflow £ran the secondary side can cause chokinJ of flow durinJ EX:cS, preventinJ adequate cooling of the core. '!be critical nllllber of tubes is relatively small. A position such as one specifyinJ a statistically significant level of nondestructive examination (NDE) might resolve this issue. '!he purpose of NDE would be to confirm that damage is not excessive on the assl.lIIPtion that damaged tubes may fail catastrophically.

2271

IIA PERI<DIC { 10-YFAR) REVIEW OP ALL POfER REACims In its report of June 14, 1966, the ACRS reoamended that periodic catprehensive reviews be conducted of operating licensed power reactors by the NIC staff. ~

reviews 1110uld be preceded by a canprehensive report by the operator which evaluated the past experience and the safety of future operation of the plant.

'!he NIC staff has maintained a cantinuin] review of the safety of operatin] plants.

In particular, as generic mtters of potential safety significance arise, the appropriate operatin] reactors are asked to assess the relevance of the matter to each particular reactor. 1his is a necessary but different aspect of the cantinuin]

surveillance and review of the safety of operatin] reactors than was envisaged by the ACRS in its recmmendation of June 1966.

1he Conmittee cantinues to believe both awroaches are desirable and awaits the developnent of a progran of periodic canprehensive reviews.

2272

Group Im - lesolution Pending - Items Added Since February 13, 1974

1. Hybrid leactor Protection Systen: Systems should be qualified for reliability, particularly through in situ tests and under various environnent.al conditions, prior to use in reactor system.

2. ()Jalification of new fuel geanetries: 'lhe 16x16, 17xl7 lWR and 8x8 BWR fuels should undergo testing to meet Item 2 in Group IC and Item 7 in Group II.

3. Behavior of BWR Muk III Containnents: various aspects, inclming vent clearing, vent ooolant interaction, pool swell, pool strati-fication, pressure loads and flow bypass should be resolved. ~is is an extension of Item 1 in Group IIA.

4. Stress Corrosion Cracking in BWR Piping: several failures have occurred in operating BWRs.

'lhe ACRS letter of February 8, 1975, discusses possible actions that should lead to generic resolution and extensive programs are underway by industry, ERllt\\, and NIC.

2273

IIB HYBRID ~

Pl011!Cl'ICB SYS'.lBE

'!he proposed systems would contain sane types of ocmponents and subsystems not previously used for reactor protection. It is necessary that the required system reliability, both during nor:mal operation and under postulated atnor:mal oonditions, be established through an awropriate cart>ination of tests and analyses.

2274

IIB ~ICATICE CF Nmi FUEL GBMETRIES New fuels proposed for both BWRs and IWR; include the 8x8 (BIIR) and 16xl6 and 17xl7 I:wR fuels.

'!he Omnittee recognizes that these fuels are intended to operate at power densities lower than earlier fuel designs. lt>wever, testing programs are ocnsidered necessary to establish their densification behavior (IC-2) as well as their behavior under al:normal oonditions (II-7). JIH)ropriate experimental programs should be developed dealing with flow blockage, behavior of fuel after partial melting and fuel response mder transient conditions. It is anticipated that the solution of this item will include a synthesis of FBP data, experiments on earlier fuel types, behavior of fuel in ccmoercial reactors, and oonfinnatory experiments on these fuel designs.

2275

IIB BEHAVIOR OF 11m MARK III COffl\\Dle1ENl'

'!he BWR M:lrk III O:>ntaimtent differs in many respects fran the Mark I and II designs. various aspects such as vent clearing, vent/coolant interaction, pool swell, pool stratification, pressure loads, and flow bypass must be evaluated and awrOYed; OBJOing experimental tests should develop much of the necessary data to confirm the con-servatism in design.

2276

IIB S'l'RE$ OORIO:UCB ClWlCIR.; IN BIR PIPIR.;

Several failures have occurred in operating BNRs.

An ACRS letter of February 8, 1975, discusses p>ssible actions that should lead to generic resolution, and extensive programs are underway by Industry, ERDA. and NlC.

'!he austenitic stainless steels are C<imDnly used as piping material in many of the smaller BWR lines. A canbination of weld sensitization, residual stresses, supeqx>sed loads, and oxygen equal to or greater than 0.2 ppn in the BWR coolant can lead to cracking, initiating on the inner surface and propagating through the wall.

In mst cases there will be a leak well before pipe failure so there is adequate warning i however, one can postulate a I.OCA caused by a guillotine break with minimal prior warning. OJrrent efforts are to minimize stress corrosion by using other materials.

2277

Group IIC - :Resolution Pending - Items Added Since March 12, 1975

1. I.ocking OJt of ECCS R:>wer q>erated valves: "lhe camdttee suggests that further attention be given to procedures involving locking out electrical sources to specific mtor-operated valves r~red in the engineered safety functions of ECCS.

2. Fire Protection: '.llle Omnittee £ecn1111::,ds review of design features intended to prevent the occurrence of damaging fires and to minimize the consequences to safety-related equipnent should a fire occur.

3. Design Features to control Sabotage: Attention should be given to aspects of design that could inprove plant security.

4. Deoontanination and Deoamdssioning of :Reactors: ~ific plans should be developed, including definitive codes and standards covering plant decamlissioning. Also experience should be gained in reactor decontanination so that stX:h information is available when needed.

s. vessel SuWort Structures: ()JeStions that have arisen ocnceming the loads on pressure vessel SUEP)rt structures due to certain postulated loss-of-coolant accidents should be resolved.

6. water Harrmer: several cases of water slugging or water hamler have occurred in both :EWR; and BNRs. corrective measures should be taken to minimize stX:h events.

7. Maintenance and Inspection of Plants: Provisions should be included in the design of future plants which anticipate the maintenance, inspection and operational needs of the plant throughout its service life.

a. Behavior of BWR Mark I O>ntainnents: various aspects relevant to the BWR Mark I containment should be resolved. Included are stX:h items as relief valve restraint, cohtrol of local dynanic loads in the torus, vent clearing and establishnent of torus water tanperature limits during a IOCA.

'.lllis is an extension of Itan 1 in Group IIA.

2278

IIC LCXlUt<<; WI' CF EO;S POJER-OPERATED VALVES

'!be '(ilysical locking out of electrical sources to specific lll'.>tor-operated valves required in the engineered safety fmctions of ECrS has been required, based on the assm,ption that a spurious electrical signal at an i.no,worttme time could activate the valves to the adverse position; e.g., closed rather than open, or opened rather than closed. lllile su:::h an event has a finite probability another probability exists that the valves might be adversely positioned due to operator error.

'!be ACRS believes the matter smuld be stmied using a systems aR)roach, and considering soch items as (1) the evaluation of the probability of a spurious signal; (2) time required to reactivate the valve operator; (3) status of signal lights,men the ciruit breaker is open1 (4) can the valve be locked out in an improper position due to a faulty indicator; (5) are there other designs improving reliability without lock-out; (6) what are the advantages and disadvantages of corrective action by an alert operator in case of incorrect positioning vis-a-vis a systen with power locked out.

2279

IIC FIRE HOJ.fCl'IOf

'lbe incidence of fires having the potential of damage to safety-related equipnent has been sufficiently large to warrant ooncem.

Measures should be taken to review design features intended to prevent the occurrences of damaging fires and to minimize the oonsequenoes to safety-related equipnent should a fire occur. 11lere feasible, nmifications should be made in older plants in addition to incorporating stX:h changes into plants now being designed. Particular attention should be devoted to regions where vital systems are in near juxtaposition since an event such as a fire could affect redundant and diverse engineered safety f mctions.

2280

IIC DESI~ FEMURES 'IO CDl.lK>L SAOO'mGE Considerable attention has been devoted to control of industrial sabotage of nuclear power plants, particularly with regard to control of mauthorized access, and potential DDdes of sabotage by individuals or groups external to the operating organization. ihe ACRS believes that deliberate attention stould be given to aspects of design that could improve plant security. With the enphasis being placed on standardized plant designs, it becanes especially important to introduce design measures that could protect against industrial sabotage, or mitigate the consequences thereof.

2281

IIC DmN.l'AMINATIOO AND DEXD1Ml$ICIITR; CF ~

'!he Coomittee believes that liell developed plans, confirmed by appropriate experiments where necessary, should be available to cover such events as the decontaminatioo of primary reactor systems and the decarmi.ssiooing of reactors. At this time the information on full scale decontamination is quite limited, particularly for BWRs.

Examples of p:>tential problems include such itens as handling of decontamination solutions, p:>tential hideout of radioactive products, enhanced corrosion and crud formation following decontamination, and the p:>ssible imcanpat-ibility of the different alloys in the pressure boundary to the decontamination solutions.

Experience is limited with regard to decannissioning operations including rules for dismantling and for mthballing. Definitive plans and standards should be developed covering such items as adequacy of action, problens in restitution of site, mutual responsibility of State and Federal Q>vermient, etc.

2282

IIC VESSEL SUPK>Rl' S'l'R[x::T(JRES A possible consequence of the instantaneous double-ended pipe break postulated to occur in certain large pipes of l:WBs is the asynmetric loading of the reactor pressure vessel support structures. '!he magnitooe and effects of such loads on the pressure vessel should be determined to establish if such loads adversely affect the predicted course of a u:x::A. If analysis indicates that the results are tmacceptable, appropriate corrective action should be taken. A potential effect is pressure vessel m:>vement due to blowaown jet forces at the location of the rupture, transient differential pressures in the annular region between the vessel and the shield, and transient differential pressures across the core barrel within the reactor vessel.

2283

IIC WATER HAMMER several instances of water slugging or.water hanmer have occurred in both BWRs and EWRs due to causes such as the trapping of water between two valves. '!his slug of water is accelerated by steam or water once the valves are opened. '!he stored energy is sufficient to damage piping, bend or break pipe restraints, and damage support structures. Nater harcmer may occur due to flow instabilities in steam generators in conjunction with water flowing into the feedwater inlets, resulting in comparable damage.

Corrective measures should be taken to minimize such occurrences after canpletion of analytic and experimental stoUes directed to an understanding of the causes.

2284

IIC MAI~E AND INSPErl'IOO OF PIANTS Experience with older plants has verified that appropriate modifications in piping layout, with respect of walls and structures, type of insulation used, and weld joint design, to cite sane obvious items, lead to improved maintenance, more reliable inservice inspections, and a better meeting of the operational needs of the plant throughout its service life, including decontamination and eventual deoallnissioning.

An additional benefit is the reduction in personnel exposures in plants making then mre amenable to maintenance and inspection.

Appropriate changes should be considered in future designs to meet these criteria.

2285

IIC BEHAVIOR OF BNR MARK I CCNrAnl>1EN'lS Recent tests on the BWR Mirk I Contairnnent design revealed phenanena not anticipated on the basis of earlier tests where pressure loads were imposed by insertion of air. Specific problems sanewhat omparable to those mder review for the Mark III Contairnnent, include relief valve discharge pipe restraints in the torus, local dynamic loads on the torus, vent clearing, and influence of torus temperature on the IOCA.

Olgoing experiments are expected to develop the necessary data to confirm the adequacy of the existing design or to permit necessary modifications.

2286