ML20073M885
| ML20073M885 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/1991 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| NUREG-1266, NUREG-1266-V05, NUREG-1266-V5, NUDOCS 9105160056 | |
| Download: ML20073M885 (69) | |
Text
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4 NU R EG-1266 Vol. 5 XRC Safe':y :lesearca in Sumor: 0:' Regu a-ion - FY 1990 e
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AVAILADILITY NOTICE I
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NUREG-1266 Vol. 5 NRC Safety Researc i in Su nort of Regu ation - FY.1990 Manuwript Coinpleted: March 1991 Date Published: Apt ti 1991 i
OITice of Nuclear llegulatory 1(esearch U.S. Nuclear llegulatory Commission Washington,1)C 20555 h
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AllSTilACT lan repois. the snth in a series of annual reports, was for related densions in suppott of Nuc neensing and prepared in resporne to conficssionat inquiries concern-inspection actisities. 'ihis researth is necessary to inale iny how notican terulatory research is used, it sumru certain that the regulations that at e iniposed on licensees r tres the accornplishnients ef the Of hee of Nuclear 1(cru-provide an adequate marrin of safety so as to protect the lator,s 1(esearth during I Y 1940.
health and safety of the public.'lhis repos t describes bof 5 the direct contributions to scientihe and technical
'lhe roat of thn of fite is to ensute that safety related knowledre with terard to nuelcar safety and their regul?-
s escart h proudes the let hnical bases for r ulernating and tory applications.
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CONTENTS IWe Alls 11(M 'l iii INI Ct rilVi! SUN 1N1 Al(Y is i
INTI'GltilY Ul l(1: AUlOlt CON 1PON1:NI'S.
I 1.1 l(cattor Vessel and Pipmp Integrity I
1.1.1 Statement of Problem.
I 1.1.2 Program Stratery I
1.1.3 1(eseanh Accomphshrnents in lY 1990.
2 1.1.3.1 Pressute Yessel Safety.
2 1.1.3.2 Piping Internty 3
1.1.3.3 Inspection Procedures and Technology 3
1.2 Aping of 1(cattor ('o oponents.
4 j
1.2 1 Statement of Problem.
4 1.2.2 l'ropram Stratery 4
1.2.3 1(esearch Aetomplishments in lY 1990...
4 1.2.3.1 Armp1(esearth 4
1.23.2 1(esidual Iare Assessment of hiaj< r 1 Wit Components.....
5 1.2.3.3 Technical liases for laecnse 1(enewal 5
1.2.3.4 Pl( A.llased Prioritization of 1(isk Contributions and hiaintenance.......
5 1.2.3.5 1(crolatory Instrument 1(eview: hianagement of Aging of 1. Wit hiajor Salet) l(elated Components.
b l.2.3.h inspeellon Illtefration b
I.2.3.7 Deptadation blodehng of Components.
b I.2.3.8 Cornponents. System %. and llaCilities....
b l.3 1(cactor l'Qulpmellt Quall[leallon 7
l.3.1 Staternent of Probler 7
l.3.2 Pforrain Strater)..
8 1.3.3 1(esearch Accomplishments in lY 1990.
8 1.4 Seismic Safety 9
l 1.4.1 Statement of Problem.
9 l
1.4.2 Program Strategy 9
l 1.4.3 Itesearch Accomphshments in IT 1990..,,......
10 j
1.4.3.1 liarth Sciences 10 1.4.3.2 Seismic 1:ngineering 1(esearch..............
13 2
Piti!Vl!NTING I)AN1 AGl! TO ltil AUi Ol( Colt!!S......
17 2.1 Plant Performance.
17 2.1.1 Statement of Problem.
17 2.1.2 Propam Stratery.
17 2.1.3 1(escarch Accomplishments in lY 1990...
17 2.1.3.1 Co :)piction of *lhermal llydrauhc Codc Development 17 2.1.3.2 Acci lent hianagement Strateries Transmitted to 1.icensees............
18 2.1.3.3 1(eactivity Accidents....
18 2.1.3.4 Emerpency Core Coolant Tests and Analyses.......,
19 2.1.3.5 Depressuri/ation As Accident hianagement Strategy To hiinimize Consequences of Dir ect Containment lleating.....................
20 v
NUl(EG-1266 l
2.1.3.6 Potential for llecriticahty m IlWits.
20 2.1.3.7 Workshop 1:hutmg Uncertainties in Accident hianapement Strategies 20 2.2 Iluman Performance 21 2.2.1 Staternent ol Problem.
21 2.2.2 Program Stratery 21 2.2.3 ltesearth Accomphshments m IT 1990...
21 2.2.3.1 Iluman I actors liestarch...
21 2.2.3.2 Itchabihty Assessment..
22 3 til:.Act olt CONI AINhilNI Piiltl'Olth1ANCli 23 3.1 Core hielt and Itcactor Coolant System l~ailure....
23 I
3.1.1 Statement of Problem.
23 3.1.2 Program Stratery 23 l
3.l.3 Itcscaub Accomplishments in lT 1990......
23
{
31.3.1 Cor e hielt Progression and liydrogen Generation..
23 i
3.1.3.2 lichavior and Chemical 17orm of 1 ission Products........
24 3.1.3.3 l'ucliCoolant Interactions.
24 5
3.2 lteactor Contairirnent Safety 24 3.2.1 Slalemerit of Problent..
24 3.2.2 Proff am btrategy 25 3.2.3 ltescarth Accomphshments in IT 1900....
25 3.2.3.1 I)irect Containment lleating 25 3.2.3.2 Ilydrogen Transport and Combustion....
25 3.2.3.3 Integrated Codes and Applications 25 3.3 Contamment Structural Internty.
25 3.3.1 Statement of Pmblem.
25 3.3.2 Program Stratcry 26 3 3.3 liescarch Accomplishment $. m IT 1990....
26 3.3.3.1 Structural Tests 26 3.3.3.2 1:quipment liatch Tests.
26 3.4 iteactor Accident itisk Analpis 27 3.4.1 Statement of Problem.
27 3 4.2 Program Stratery 27 3.4.3 ltescart h Accomplishments in IT 1990....................
27 i
I 3.4.3.1 Itesiew of Pit As.
27 3.4.3.2 Compic, on and lleview of 1(cactor Itisk Iteference Document.
28 3.4.3.3 Itisk Model Dctelopment and Application.................
28 4 ASSliSSING sal'l!TY 01: NUCl.l!All WASili DISPOSAL......
29 4.1 Iliph-I n cl Waste 29 4.1.1 Statement of Pioblem.
29 4.1.2 Program Stratery 29 4.1.3 llesearch Accomphshments in IT 1990.
29 4.1.3.1 Ily drorcology 29 4.1.3.2 Stabihty of Underground Openings 30 4.1.3.3 Seahng of lloreholes and Shaf ts in Tuff 30 4.1.34 Waste Package Performance.
30 N Ulti.0 - 1266 vi
4.l.3.5 (ien tche fHi5tly
.IO 4.1.3.h l'cifor mance Auc% ment
.I l 4.l.3.7 ilulemalmp 31 4.2 1A m l.c\\cl W aste 31 4.2.1 Staternent of l'toblem.
31 31 42.2 l'ropiam Stratery 4.2.3 1(esearch Acwmphshments m lY 1990...
32
.l. 2.3.1 1:npincermy rnhancements and Alternatises to Shallow Iand llutial 32 4.23.2 1.1 W W,nte l orms 32 4.2.3.3 Inf0tration el Water.
32 4.2.3.4 l'ettormance Aucksment 32 4.2.3.5 1.1 W Source Term Modeling 32 4.2.3.6 Ilydrolory and Contatmnant Transport.
32 4.2.3.7 1tLlernalmp......
32 I
$ 111 M )1NINt i sal IElY 1%Ul:S ANI) I)lNI'l Ol'ING lt1:01it.ATIONS......
33 51 Genene Safety luucs.
33
%.l.1 Statt ment ol Problem.
33 5.1.2 Program Stratery.
33 l
5 1.3 Itescatch Accomplishments in IT 19n0....
33 5.2 Standatdved and Advanced lteactors.
33 5.2.1 Statement of Problem.
33
$ 2.2 Propram Stratery 33 5.2.3 ltescarch Accomphshinents in IT luva.
35 5.2.3.1 Advanced lteactor Concepts and Standatditation 35 5.3 liuel Uple. I ranspottalion, and Safcpuards 35 5.3.1 Statement of Ptohlem.
35 5.3.2 Propratn Stratepy 35 5.3.3 ltesearch Accomplishments in IT 19n0 36 5.3.3.1 Irutt Cycle 36 5.3.3.2 Transportation and Safeguards..
36 5.4 I)evelopmp and linprovmg liepulations.
36 5.4.1 Statement of Problem.
36 5.4.2 Program Stratery.
3b 5.4.3 ltesearch Accomplishments in IT 1990..
36 5.4.3.1 1)esclop or Modify llegulations 36 5.4.3.2 lleputatory Analysis.
38 5.4.3.3 Independent iteview and Conttol of llulemaking..
38 5.4.3.4 1.iernse lienewal 38 5.4.3.5 timergency Preparedness.....
39 5.4.3.6 Safety Goal Implementation..
39 5.5 Severe Accident implementation.
40 5.5.1 Statement of Problem.
40 5.5.2 Program Strategy,
40 5.5.3 Ilescarch Accomplishrnents in IT 1990..
41 5.5.3.1 Indnidual Plant lixaminations 41 5.5.3.2 lixternal I! vents 41 vii NUltliG-1266
5.5.3.3 Containtnent l'c for mar.cc linprosements 42 5.6 l(adution Protection and llcalth l~lfrets.
42 5.6.1 Statetnent of l'rotilern.
42 562 l'rortion Straterv.
43
$6.3 Itescarth Accornphslunents in IT luul).
43
$ 6 3.1 1(adiation l'rotection li. sues.
43 5 6.3.2 llealth I'.flects 1(esearch 44 5 6.3.3 I)culopinent of 1(utes and 1(erulatory Glades 4$
$.6.3.4 I nonintnental Pohey and I)ecornrnissioning 46 Al'I'liNI)1X A-1 TAlli,ES i
Table 5.1 issues prioriti/cd in IT luull 34 Table 5.2 Generic s.if ety iuocs resobed in lY 199t) 34 Table 5.3 1(ulenuking actions processed during IT 1940.
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EXECUTIVE SUMMAlW l
l Nlte f.afety research is vital foi implementmg a large e
PitlTI;NTING DAh1AGl:'IO IllMCIOI! COltl:S number of the agency's programs. Itesearch provides the bases for timely rulemaking and related beensmg and The nstem thernmbhydraulic computer codes, Ill1APS and TitAC PFI, were completed.
inspection activities that are based on NitC's longstand.
.lhe work included validation using experimen-ing philosophy of defense in depth. This philosophy pro' tal simulations by an independent international v' des a clear and lopcal structure for the safety research users group.
mission area, which consats of fne major programs: In-
+
tegrity of lteactor Cornponents, Preventing Damage to
- An evaluation of the effects of intentional iteactor Cores,iteactor Containment Performance, As-depressuritation for a specihe reactor geometry sessing Safety of Nuclear Waste Disposal, and itesolvinF was completedJihe results showed that depres-Safety issues and Developmg Regulations.
suri/ation can be very effective for reducing the ptobability of catly containment failute by di-Provided herein is a summary of findmps, results, and rect containment heating during severe accident accomplishments of the safety research mission at cas that simulations.'the work will be extended to other (1) have led to, or are being incor porat ed m,3profc Com-geometries.
minion und staf actions to ensure or enhance the level of safety in actinties or facihties being regulated; (2) demon-
- 1luman factors studies on the layout of local strate a ncca for change in regulations or reputatory reactor control stations showed that centraliza-approach; or (3)confun or support the rrgulations or rep u-tion of the control functions is an effective latory approach. A summary of the major results in the method of enhancing safety via improved hu-five safety research programs is provided below followed man reliability.
by a rnore detailed discussion. Iteputatory products emerging flom these propams are listed in the appendix
- Detailed reactivity insertion studies on U.S.
to this report, l.Wih showed that the probability of a Chernobybtype event for U.S. reactors whs very low, thereby confirming the adequacy of current e
INTI:GI(11Y Or itEACIOlt COS1PONI;NTS U.S. reactors and regulations.
- A correlation was developed to determine the
- thperirnental data from the Upper Plenum fracture toughness of r eactor components usmg Test Facility and Tit AC computer analyses con-data from simple fick tests. Reliable predic-firmed the effectiveness of emergency core tions of long term serviceability of critical com.
cooling water injection into the upper plenum, ponents in reactor systems can now be made.
thereby precludmg the need for backfitting such plants.
- Acoustic emission techniques were developed RIMCI'Olt CONTAINMl;NT PPIll'OltMANCE e
to evaluate and momtor crack seventy in reactor l
componems. The procedures are 'now being
- Metallurgical samples were successfully re-used to monitor cracks in welds at Limer;ck moved from the lower head of the TM12 vessel.
Unit 1.
't hese samples, which were in contact with mol-l ten material, will be extensively studied for use
- A computerited data tuse was devehiped on radiation embntilement using data from reactor
- Detailed reviews of six licensee probabilistic risk surveillance specimens. It is being used to up-assessments were completed.
date reputatory guides and support regulatory acti ns.
- All reviews (generidly positive) of the staff's assessment of severe accident risk at five nu-clear power plants (NURl!G-1150) were com-l
- Dynamic, simulated scismic kuding tests were pleted.
completed on typically si/ed piping. The results simwed that the effects of dynamically imposed ASSESSING SArl!!T OF NUCLEAll WASTE stinin rates on fracture toughness are ade.
DISPOS'A1' quately addressed by curt ent models used by the NitC for leabbefore-break analyses of I.WR
- Analyses of ladionuclide dispersion data using piping systems.
Fick's laws of diffusion have shown that the use ix NURf!G-1266
of such laws may not be adequate in estimating e GI-94, "AJditional I ow-Temperature dnpersion to an accuraty sallicient to ensure Overpreuore Protection f or I.Witt" See comphance with the 1:p '.10,000-year cumula-N Ulti.G-1326.
tive litmts on release to the luosphere, Gl-103, ~l) estro for Probable Maumum e
- 'the methodolorv for assessing higa-level waste Precipitation." Standard Hesicw Plans 2.4.2 performance wai completed at Sandia National and 2.4.3 werc revised.
Iaboratones and transferred to the NitC and the Center for Nucicar Waste Iteputatory Gl A-29," Nuclear Power Plant Design for e
^"" IPC5-the Reduction of Vulneralnhty toindustrial Sabotare." See NUltliG-1267.
i Gl C4
- Main Stemn Une Iwladon Vahe itLSOINING SAirlT ISSUES AND DlWLl.OP.
e e
ING lti:GUlKilONS 1.calare Control Sptem Fad ur t,,
See NUlG.G-1372.
- A draf t renerie letter and draf t NUREG-1407 on the scope, menodoiory, and reporiing re.
A, INTEGRITY OF REACTOR qunements for individual plant examinations COMI'ONENTS f or esternal events (IPlil!!!) were issued.
All major elements of the containment per-formance irrprovement program were com-a.
!!mbrittlernent of reactor pressure vessel ma-f pleted. Generie letters were issued to licensees terials is the inost important issue for pressute to begm thc plant specific backfit of the hard-vessel safety, and pressurized thermal shock ened sent for all Maik i umts.
(Pl'S)is the most serious postulated transient for these vessels. 'lhe regulation concerning
- The final report, NUlti:G-1333, concerning the permissible level of ernbrittlement to ruard the NRC policy statement on the maintenance aramst vessel Imlure due to PTS is 10 C1711 of nuclear power plants was issued in Apil 50.61.
The procedure for calculating I W u.
embrittlement m the rule, however, has not l
been in agreement with Revision 2 of Reputa j
- 'lhe fmal Policy Statement on llelow ltepula-tory Guide 1.99,"Itadiation limbrittlement of l
tory Concer n was pubhshed m the Tcdcra/ Regi3 Reitetor Vessel Materials" *lhe rule is being irr, it provide a consistent risk-based frame.
revised to incorporate the current reputatory work for determining whether a practice can be guide formulation. 'lhe revision is in the final esempted from some or all usual egulatory approval pr ocess and is expected to be r cady for controh.
publication in UY 1991.
b.
The primary system coolant pipes of many
't he fmal rule,10 Cl:R Part 72, to allow holders of nuclear power reactor operating beenses to pressurized water reactors, plus pump and s' ire spent fuelin NRC approved casks at reae-vahe bodies, m e made of east, austenitic stain-tot snes was published in the 1rdcral Regater.
less steel. This material, however, gradually loses toughness simply by exposure to normal A final rule on " Palladium 103 for interstitial operating temperatures; too much of a loss Treatment of Cancer"to amend 10 CI:R Part 35 couid eliminate our safety margins. A correla.
i was issued.
tion has now been developed that allows pre-dictions of the fracture toughness for compo-nents in service using measurement of material 1 mal technical resolution for the following characteristics from a simple field test and the genene issues was completed.
manufacturer's composition specifications.
GI-70, "PGRV and lilock Valve Reliabil-Thus, much improved judgments can be made e
ity," See NUlGiG-1316.
about the long term serviceability of the com-ponent.
GI-75, " Generic implications of ATWS Acoustic emission inspection techniques have e.
Hunts at the Salem Nuclear Power Plant."
been developed to evaluate crack severity in reactor components. The ASMli Code has Gl44 "Cli PORVs " It was determined adopted this method to monitor crack prowth in e
that no new requirements were necessary, operating reactor components. 'ihe technology NURI G-1266 x
- _ =
i is now being used to monitor <nd validate an pipmg systems of light water reactors (1.Wils).
mtergranular stress corrosion crack in a nonle PermMinn to emphiy I 1111 requires proof of weld at the 1.imerick Unit I reactor, appropriate matettal, loadmg. and environ.
mental characteristics. Dynamic, simulated d.
A comprehensive collection of radiation seismic loading tests have now been completed ernbnttlement data from reactor vessel surveil-on typical size piping in the International lance specimens, as well as other published Piping Integrity llescarch Group (IPiltG) pro-data, has been compiled in a computerized data gram. This work demonstrated that dynami-base, the Power Itcactor !!mbrittlement Data cally imposed strain rate effects on tracture liase. 'lhis data ba:e provides the information toughness of the pipe are acceptably handled needed to update regulatory guides and sUp-by the analysis methods curr ently being used by port regulatory research. The lilectric l'ower the NitC and the industry for i 1111 in 1. Wit 1(escarch Institute, reactor vendors, utilities, piping systems.
and :escarch institutes are using this data batc to help in solving embrittlement problems, b.
A draft regulatory guide was developed to ad-vise utilities planning to renew then heenses of 2.
Need for Change the necessary technicalinformation needed to support their applications. 'the document pro-a.
l'aligue i ? teactor components is governed by vides guidance criteria for the selection of the fatigue lit s ct rves in the ASMI! Code, Section key plant structures, systems, and components 111. These t 1rves were produced from polished needing aging management programs; guid-specimens t, :ted at room temperature in air.
ance on evaluation of operational and environ-liut tests th t have now been performed in mental factors that may contribute to age-typical reactor coolant environments and at related degradation of these key structures, typical operating temperatures show a shorter systems, and components; guidance on the fangue hfe. Although this has implications for identification of operative aging mechanisms license renewal applications, it is also impor-and their locations; and guidance on the mitira-tant for current regulation. Contractor testing tion and management of the resulting aging efforts are being expanded, and cooperation is degradation. The necessary information to de-under way with mdustry groups to produce ap-velop this guide is a product of the Nuct, car propriate data so that a consensas can be Plant Aging flescarch program.
reached on modifications to the curves for more accurate fatigue life predictions in the H. PitEVENTING DAMAGE TO future and to enture the continued safe opera-itEACTOlt COltES tion of today s plants.
b.
Itecent results from test reactor irradiations 1.
Specific Actions to linhance Level of Safety suggest that the current approach to shifting Completion of computer codes IlliLAPS and the fract ure toughness curves derived from sur-TilAC-PFl used to analyze hypothetical thermal-veillance specimen tests to account for irradia-hydraulic transients in reactors was accomphshed, tion damage in reactor vessel materials may not including comparisons with experimental data by an completely account for that damage, it appears independent international group of code users.
that the proecdure may be underpredicting the
'these codes are currently being used to support actual shift in the fracture toughness curves, Commission safety reviews of operating reactor is-decreasing the anticipated margin of safety in s'ues such as llWR stability, anticipated transients' some of the regulatory analyses. At the same without scram, and pressuriecd thermal shock as time, other aspects of the overall pressure ves-well as evaluations of accident management strate.
set integrity analyses are known to be extremely
- gies, conservative so that the fmal evaluations still are quite conservative. More irradiations are 2.
Need for Change required to validate the test results and to de-a.
A staff evaluation of intentional depressuriza-fine any needed corrections, tion in one type of reactor geometry was found 3.
Confirmation of llegulations to be a potentially effective accioent manage-l ment strategy for reducing the potential for a.
A modification to General Design Criterion 4 carly containment failur e due to direct contain-several years ago allowed utilities to climinate ment heating (DCH). The evaluation, which is many pipe whip restraints, through establish-continuing for other reactor types, will support ment of leak before break (Illis) in primary staff reviews of licensee accident management xi NUltliG-1266
-.. _ ~.
i plans submitted as part of the Commission's 2.
Confirroation of Regulations individual plant examination (IPE) program.
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Detailed staff reviewsol sixlicensee generated l
probabilistie risk assessments (PR As) were car-b.
Iluman factors research on the design of hical ried out in support of NRC's continuing assess-reactor control stations demonstrated that cen-ment of the safety of operating plants, tralizingontrol functions in the design stage of l
tcactor devdopment is an effective method of b.
The staff's assessment of severe accident risks enhancing safety through improved human re.
at five nuclear power plants was published as liability. Such infor mation is supporting reguta.
NUidIG-1150 (second draft for peer review) tory initiattves regardmg functional centralita.
in June 1989. Peer review of the second draft, tion of advanced or improved instrumentation performed under the provisions of the Federal and control systems. This research is also being Advisory Committec Act, was completed in applied to the development of guidance for the August 1990 with publication of the committee ieview of advanced control systems and control findings and recommendations in room designs bemp proposed for advanced NURiiG-1420. In addition. the American Nu-LW Rs.
clear Society comp'eted its review of NUREG-ll50 aad published its findings and 3.
Confirmation of Regulations recommendations. Both committee findings were generdly positive.
i In response to the Chernobyl accident, a study One of the prmciprI conclusions of the second a.
of United States 1.WRs was carried out to as-draft of NUREG-1150 was that the estirnated sess the potential for sudden, large increases in severe accident risks from internally initiated reactivity that would cause severe fuel damage.
cecidr
- %d d- - ased, relative to the NRC's a
The study determined that the probability of b75.M _
"ses (the Reactor Safety Study).
such events was quite low, thereby confirming Fun m ' m rofsevereaccidentsinitiatedby the adequacy of current regulations applicable ven,o ca. quakes in the Eastctn United to reactivity control.
bw M.J,cc artwoeastern plants)wasscen to.
.c con. able unecrtainty, principally l
b.
Data from the Upper Plenum Test Facility drivi 4.hu..a.rtaintyinihe estimates of the (UPIT) and TRAC computer analyses cow likeinood of a n earthquakes. The results of firmed the effectiveness of coergency core the peu -
as did not callinto question these cooling water injection into the upper reactor conclusions.
plenum. This information allows full credit to be given to the performance of emergency core With cempletion of the peer reviews, the staff cooling systems in those plants equioped with has proceedeo to pubbsh the final version of upper plenum injection and avoids the poten.
NUREG-ll50. Volumes 1 and 2 were pub-tial need for backfitting on those plants.
lis"ed in December 1990 and Volume 3 in Feb-ruary 1991.
C. REACTOR CONTAINMENT D. ASSESSING SAFETY OF PERFORMANCE NUCLEAR WASTE DISPOSAL C
^
"U 1.
Need for Change Management Licensing in a multinationally supported research effort, met-a.
I Research investigations completed during FY allurgical samples were successfully removed from a 1990 have shown that models of radionuclide sectionof thelowerreactorvesselheadof theTMI-2 dispersion based en an analogy with Fick's law reactor that was in direct contact with molte: core of diffusion, the conventional way to estimate material. lixamination of these samples is expected such dispersion, may be wrong and that l
to provide information regarding the nature and ex-non.Fickian estimates of dispersion may be im-l tent of any attack of melted core material on the ves-portant in evaluations of compliance with the sel head, to shed some light on the margin to failure Iinvironmental Protection Agency (EPA)high-of reactor pressure vessels under severe accident level waste (llLW) stanjard's 10,000-year conditions, and to provide some insights as to Ihe ef-cumulative limits on radionuclide releate to fectiveness of certain accident management strate-the biosphere. The correctness of dispersion gies directed toward mitigating core melt events.
models may become more important if the N URiiG-1266 xii
EPA d:cides to include concentration-based the individual plant examinations for external safety n iteria in its l ll.W standard. Research is events (IPEEE) were issued. A workshop on continuing to determine what the best alterna-these documents was conducted in Septernber tives to I:ickian dispersion models are.
1990 to sohcit comments and answer questions concerning their content.
b.
Mathernatical models using data taken from the Alligator Risers Analogue Project (AR AP) b.
The Commission has received two individual were used to produce iesults that suggest possi-plant examination (IPE) submittals. In hiay i
ble avenues of simplifying coupled hydro.
1990 a meeting was held with Yankee Rowe to chemical models of radionuclide transport in review their submittal and in July 1490 a set of unsaturated rocks.
questions on the submittal was sent to Yankee Rowe. The staff review has just started on the c.
1he high-level radioactive waste performance hiillstone Unit 3 submittal, received in late assessment methodology that Sandia National September 1990.
Iaboratories developed for NRC was trans-ferred from Sandia to the staffs of NRC and the c.
'lhe staff has completed all major elements of Center for Nuclear Waste Regulatory Analyses the containment performance improvement (CNWR A). Future.nochfications t.nd imprbve.
(CPI) program. Genene letters (Gl.s) have ments to the methodology will be made at been issued to licensees to start the plant-CNWRA.
specific backfit of the hardened vent for all llWR h1 ark I containments (Gl 89-16, dated 2.
Neca for Change September 1,1989) and requesting that other improvements be considered in the IPE (Sup-a.
The possibility of concentration-based stan-plement 1, dated August 29,1989, to GL 68-20 dards see D.l.a) for high level-waste disposal for llWR hlark I containments and Supple-s tequires accurate assessment of the effects of ment 3, dated J uly 6,1990, to GL 88-20 for the dispersion on radionuclide concentrations. On-other containment types). Seven of the gomg research on non-Fickian dispersion wdl planned 12 repons have been issued to docu-develop rnodels to provide these estimates.
ment the analyses end evaluations donc by the b.
Chemistry will affect radionuclide transport staff and its contractors in assessing the various from HLW. When NRC assesses the HLW containment types. The remamder of the re-licensee's deinonstration of compliance with ports are scheduled for issuance by April 1991.
HLW regulations that require considerations d.
The Department of Energy has submitted for of chemistry and transport, recourse to mathe-NRC review three advanced reactor concep-matical models will be needed. Current mod-tual designs. The designs comprise one modu-els, although simph and efficient, are often lar high-temperature gas-cooled reactor wrong. Ihe use 01 more correct models, a.1-( All fl G R) and two advanced liquid metal reac-i though acknowledged to be desirable, was tors (Sodium Advanced Fast Reactor [SAFR]
thought to beintractable.Thc AR AP modeling and Power Reactor inherently Safe h1odule exercise holds the promise that more valid
[PRISN1]). The reviews sought to determine modeling of chemical effects on radionuclide the licensability of these unique advanced reac-transport could become a practical reality.
tot designs. Draft safety evaluation reports (SERs) for MilTGR, NUREG-1333, and c.
NRC plans to use the IILW nerformance as.
PRISM, NUREG-1368, have been issued The sessment methodology as a major 1it.W iicens-SAFR draft SER, NUREG-1369, wdl be is-ing tool. It is important that contractual ar. sis-tance be available to maintain and improve the sued m early 1991 to document the work com-pleted to date and to close out this review since methodology.
DGE has discontinued work on the SAFR ad-vanced reactor design. The Advisory Commit-E. RESOLVING SAFETY ISSUE 5, tec on Reactor Safeguards has issued letters en AND DEVELOPING all three of these conceptual designs. Three REGULATIONS repons have also been issued: NUREG/
CR-5261, " Safety Evaluation of MH1GR Li-1.
Specific Actions to Enhance I evel of Safety censing Basis Accident Scenarios"; NUREG/
CR-53M, " Summary of Advanced IA1R a.
In July 1990, a draft generic letter and draft Evaluations-PRISM and SAFR";. and NUREG-1407 to describe the scope, accept-NUREG/CR-5514 "Mndelmg and Perform-able methods, and reporting requirements for ance of the MIflGR Reactor Cavity Cooling l
xiii NUREG-1266 m----
-r
l System." In 1990, DOli provided the NRC with sent a significant change from the methods pre-two additional amendments to the Pit!SM viously employed to assess and control prelimi_ nary safety information document that radiation doses. 'Ihe new Part 20 will result in -
address the open issues identified in reducing the annual occupational dose from a
~
NUlt!!G-1368 and describe changes to the -
possible 17 rems (3 rem / quarter external +
4 PitlSM design. The NRC waff is reviewing 5-rem annual internal) to a total effective dose these design changes and plans to issue a of $ rems per year. 'lhe dose limit for members revised SER in 1991, DOE also announced that of the general public is reduced from an im-they are reassessing the MilTGR program and plicit 0.5 rem peryear in the present rule to an performing a cost reduction study on the explicit valuc of 0.i rem peryear.The new Part M11TGR design.
20 contains appendices that give the radio-nuclide concentration limits for air, water, and c.
The Commission issued a Policy Statement on 80*""EC' the Maintenance of Nuclear Power Plants in g.
In a significant policy action, the Commission March 1988. In this policy statement, the Corn-mission indicated its mtention to pursue a published in the Federal Registcr a Final Policy Statement on llelow Regulatory Concern rulemaking on maintenance, In developing th,si proposed rule, the staff had extensive interac-(llRC) on July 3,1990. The policy statement rovides a consistent, risk based framework for tions with U.S. industry (airline and nuclear) and foreign nuclear maintenance programs and determining whether a practice can be ex-practices. In addition, a 3 day pubhc workshop empted from some or all of the regulatory con-was held in July 1988 to solicit feedback on Rols usually imposed by the Commission's rulemaking options. Infonnation gathered in Ation 'lhe policy will be applicable to these interactions and from the workshop was exemption decisions in a number of areas, in-used in formulating the proposed rule and its ciuding the disposal of very low activity waste, supportiry regulatory guide, the proposed the decontamination and decommissioning of h mi k aPpmval of consumer products, rule was published for public comment in tL liederal Rcxister in November 1988.'lhe Com-tmd the potential for recycle and reuse of mate-mission issued a revised Polig Statement on rials and equipment, Dunng late August and Maintenance of Nuclear Power Plants in Octo, September 1990, a series of pubhc meetings was Id to explam the llRC pohey and provide ber 1989. 'Ihis revised policy statement stated the Commission's expectations in maintenance an opportunity for public comments and state-and indicated the Commiwon's intention to men's. 'Ihe meetmgs were held m Chicago, hold rulemakmg in abeyance for an 18-month Illmots: King f Prussia, Pennsylvania: Atlanta, period while it monitored industry initiatives Geoygia: Arlington, Texas; and Oakland, Cah, and progress in maintenance. During this hm. An esdnoted 940 persons attended the period, the Commission indicated that it would public meetings. the makeup of the audiences continue to develop a maintenance rule. In differed somewhat from meeting to meeting, support of the Commission's rulemaking effort but the audiences were largely composed of on maintenance, a draft regulatory guide was representauves of pubhc interest or e iron-issued for public comment in August 1989. In mental groups, concerned citi/ ens, anu.
l addition, the final NUREG-1333 "Mainte-eral, State, or local elected representatives, j
nance Appmaches and Practices in Selected h.
It is clear that a major need exists for additional l
Foreign Nuclear Power Programs and Other spent fuel storage space at commercial nuclear
)
U.S. Indust nes: Review and Lessons Learned, power reactor sites.- to be available in the near j
was issued in April 1990, future. In tesponse to this need, the Nuclear 1
Waste Policy Act of 1982 directed the Secre-f.
The Gormnission approved using a complete tary, Department of linergy, to establish a dry i
_ revision to the NRC regulatlons for radiation spent fuel storage demonstration program, protection in 10 CFR Part 20. This revision.
with the objective of coming up with one or
. updates the Commission's regulations to incor-more technologies that the NRC might ap-porate recommendations made by the Interna-prove for use at civilian nuelcar power reactor i
tional Commission on Radiological Protection.
_ sites without, to the extent practicable, requir-j:
the National Council on Radiation Protection ing additional site-specific approvals. A final and Measurements, and the revised Federal rule,10 CFR Part 72, published in the Federal l
Radiation Guidanse for Occupational Expo-Register in July 1990 will allow holders of nu-j sure issued in 1987.The new standards repre-clear power reactor operating licenses to store NURl!G-1266 xiv t
i I
=
spent fuel in NRC-approved casks at reactor 1.
In October 1989, the Commission issued a final sites under a general license, rule, " Palladium-103 for Interstitial Treat-ment of Cancer," to amend 10 CFR Part 35.
i.
A proposed rule was published in the Federal
'ihis rulemaking responded to a petition Register in January 1990 that would amend the (PRht-35-7) requesting that Pa-103 be m-10 CI'R Part 35 regulations that apply to the cluded in the list of approved sources so that its rnedical use of byproduct material.The amend.
therapeutic use yvould no longer require spe-ments would require medical use licensees to cial licensmg action.
implement quality assurance (0,A) programs m.
Final technical resolution was achieved for the and would revise misadministration reportmg following generic safety issues:
requirements. Implementation of the new GI-70, "PORV and lilock Valve Reliabil-performance-based requirements would be e
supported by lauance of a regulatory guide ity." NUREG-1316 provided technical that would include specific guidance for QA findings and a regulatory analysis of this issue.
programs. The rule would enhance patient safety while allowing the flexibility necessary GI-75, "Generie implications of ATWS for proper medical care.'!he feasibility of this Events at the Salem Nuclear Power approach is being evaluated during a pilot pro-Plant " 'lhis issue was administratively gram involving 70 medical-use !icensees 'lhe resolved.
Commission expects to consider a final rule in GI-84, "CE PORVs." It was determined e
the early part of 1991.
that no new requirements were necessary, GI-94 " Additional low Temperature j.
An interim final rule was published in the e
rederal Registcr in August 1990 amending 10 Overpressure Protection for I WRs" CFR Part 35 in response to a petition for NUREG-1326 provided a regulatory rulemaking from SNN1/ACNP that requested analysis of this issue.
departures from instructions approved by the GI-103 " Design for Probable hiaximum e
I ood and Drug Admmistration (FDA) for ra-Precipitation," Standard Review Plans diopharmaceuncals so that physicians may pro-2.4.2 and 2.4.3 were revised to accommo-vide proper medical care to patients. I his rule' date issue resolution.
which was coordinated with the FDA in its de-GI A-29. " Nuclear Power Plant Design selopment, allows certain departures from 1 DA approved instructions while providing for the Reduction of Vulnerability to In-reasonable assurance of radiological safety as dustrial Sabotage." NUREG-1267 pro-well as a balance between adequate conitols vided the technical resolution for this and avoidance of undue interference in medi.
issue.
cal judgment.
GI C-8, "hiain Steam isolation Valve Leakage and Leakage Control System k.
The Commission has approved a proposed rule Failure." NUREG-1372 provided the that would amend the regulations to Part 50 to regulatory analysis for the resolution of require the licensee to implement the NRC-this issue.
approved Emergency Response Data System 2.
Need for Change (ERDS) at all nuclear power plants. The pn-mary role of the NRC during an emergency at a In J uly 1990, the Commission issued a proposed rule licensed nuclear power facility is one of moni.
(10 CFR Part 54) for nuclear power plant license toring the licensee to ensure that appropriate renewal. The following support documents were is-recommendations are made with respect to sued with the proposed rule: (1) draft necessary offsite actions to protect public NUREG-1362, a regulatoryanalysis to provide sup-health and safety. ERDS would supplement porting information for the proposed rule; (2) draft the voice transmission over the existing Emer-NUREG-1398,an environmentalassessment of the gency Notification System by transmitting possible effects of issuing nuclear power plant li-timely and accurate updates of criticalinforma-cense renewal standards by means of the proposed tion on plant conditions from the licensee's rule; (3) NUREG-1411, providing a response to onsite computer to the NRC Operations Cen-public comments fwm the November 13-14, 1989 ter. It is expected that the proposed rulemaking public workshop cr license renewal and written will be published in the Federal Register for pub-comments on the workshop received shortly there-lic comment in early FY 1991, after; and (4) draft NUREG-1412, describing the xv NUREG-1266
reguhitor) hasis for the generic finding for all nu-relicensing purposes was issued, and a notice of in-clear power plants that reasonable assurance of ade-tent to prepare a peneric environmental impact quate protection (findmp made nt the time of issuing statement on the ef fects of renewing operating li-an operatinglicense) continues to be true at the time censes of indiviJual nuclear power plants was also of renewal apphcation. In July 1990, an advance issued.
notice of proposed rulemaking to amenJ Part 51 for 1
I t
NURiiG-1266 n.i
l INTEGRITY OF REACTOR COMPONENTS
'this program is conducted to ensure that reactor plant during normal service and accidents and that sufficient systems and related components perform as designed critical experiments are conducted to validate those during both normal operations and accidents in order that procedures, their functional integrity and operability can be nain-tained over the life of the plant. Reactor safety depei.ds on maintaining the reactor system pressure boundary un-1.1.2 Program Strategy l
damaged and leaktight, Failure to maintain pressure l
boundary integrity could compromise the ability to cool l
the reactor core and could lead to a loss-of-coolant acca
.!he approach used for this element is to develop analyti-dent accompanied by release of hazardous fission pre cal procedures for predicting continuing integrity or con.
ditions for failure and to ensure that an adequate experi-mental basis exists to validate those procedures.The most critical facet of pressure vessel integrity is that of embrittlement of the pressure vessel steel as a result of 1.1 Reactor Vessel and Piping bombardment by neutrons escaping from the fuel core liitegrity during normal service, lixperiments are thus conducted to develop a base of information on all Ihe factors that will cause this embrittlement to increase during senice life.
1.1.1 Slatement of Problem Much work is done to establish correlations between small-specimen behavior and thick-section behavior to The primary system of a light water reactor (i.WR)is the ensure that the analyses performed to assess structural principal boundary enclosmg the nuclear fuel core and integrity are valid. Thus, use is made of large-scale " mod-the water co,olant used both to maintain suitably low tem-cls" that realistically reptcsent the true components.
peratures on the fuel cladding and to conduct the heat Similarly, the ability to predict integrity in piping has from the fission reaction to heat exchanger (for a pWR) required testing of full-sized sections of pipe having a where it can be converted into steam for electricity gen-variety of cracks that could develop in service to deter-cration.The primary system includes the reactor pressure mine if such cracks could cause lailure during either nor-vessel, primary coolant piping, primary pumps, and stearn mal senice or an accident. For both vessels and piping, generators. For boiling water reactors (llWRs), the knowledge of the rate at which cracks grow is very impor-primary system must include the steam line at least out to tant to ensure that a component will not fait during its t he first isolation valve.1his boundary must be kept intact forthcoming operational period. Thus, many experiments and fully seniceable at all times to ensure that water are conducted on a wide variety of pertinent materials coolant is always available to cover the fuel core so that under a very wide range of typicaland expected exposure heat, either from direct power generation or from decay conditions to determine the maximum bounding rates of following shutdown, can always be safely conducted away, crack growth. Detection and sizing of flaws and cracks in i
thus precluding a core meltdown accident. The principles all primary system components are conducted by the in-of ensuring the structuralintegrity of the primary system dustry through periodic inservice inspections at shut-components are embodied in the elements of fracture downs. To ensure that the inspections reliably detect and mechanics procedures used to predict conditions for fail-accurately size the flaws, extensive tests are conducted l
ure. These elements are (1) knowledge of the material with inspection teams drawn from the industry using typi-properties (strength, toughness, embrittlement, etc.),
cal equipment and techniques on samples whose flaw j
especially the changes in those pro erties that can occur conditions are known. From the results, it is possible to j
as a consequence of nuclear operr wns:(2) knowledge of determine which techniques are effective and the magni-the pressure and other stress loading : hat can be applied tude of the error bands for both detection and sizing.
l to the components either from normal oprations or from improvements in methods are proposed and qualification l
accidents; and (3) knowledge of the presuce and size of procedures developed that can provide better assurance cracks or other flaws in the components. h regulations, for not missing flaws in future inspections and for sizing codes, guides, etc., that pertain to the structuralintegrity flaws more accurately. Use is made of materials and com-of LWRs are focused to ensure that possible combina-ponents removed from actual sesice to measure the real tions of macrial properties, loads, and flaws will yield condition of material properties resulting from years of adequate margins against failure of primary system service, to establish the real corrosion state, and to vali-components. The goal of the Reactor Vessel and Piping date the existence of flaws that have been " called" and Integrity element is to ensure that appropriate analytical estimated in size through nondestructive examination procedures exist for assessing the safety of components procedures.
1 NUREG-1266 l
- -. - - _. - - -. ~ -
l ~l.3 Research Accotuplisliments iti IT 1990 arrest fracture toughness contrasted to the effects of low ductile teanng resistance. That work elearly demon.
1.1.3.1 Pressure Vesul Safet) strat d hat, while there were benefits to be gained by -
changmg the crack anest _ toughness curve used in the Nit ("s regulations for vessel safety have always been fracture analyses, these were negated by low ductde tear-designed to ensure an adequate level of ductshly and my resistance. This led to a reduced emphasis on the fracture toughness in the base metal and welds of reactor crack arrest iescarch but an inercased emphasis on duc-pressur e vessels. Ilowever, many of the analysis methods tile fracture analyses and on ductile tearing resistance 4
and material characteristics used in evaluatir.g vessel data.
I safety were based on engineering practice, which was I
intended to be very conservative. As plants have aged and The FY 1990 work to vahdate and expand the material j
as the industry has started to consider license renewal, it data bases involved compiling surveillance reports from
]
has become c! car that the safety analysis methods must be the commercial power reactors and testing materials irra-j accurate, not just arbitranly conservative, and that the diated in test reactors. A comprehensive collection of j
data bases used to deduce matenal charactenstics and radiation embrittlement data from surveillance reports j
properties must refleet actual material behavior, 'lhus, and other published reports of commercial power reae-l vahdauon of the analyses and the material data bases is tors has been compiled in a computerized data base, the necessary to ensure an adequate margin of safety and to Power Itcactor limbattlement Data llase (Pit l!Dil).
j preclude unnecessanly large marpms.
This data base provides the information needed to update i
Itegulatory Guide 1,99 and supports other embrittlement l
During FY 1990. research was performed adJressing both research projects. The 1.lectric Power itescarch Institute vahdation of analysis methods and validation and expan-(l! Pill), reactor vendors, utilities, and research institu-smo of material data bases. The analysis method valida-tions are using this data base to help in solving embrittle-tion work focused on developing and enhancing methods -
ment probleins. Itepresentatives from several foreign f
for evaluating flaws h>cated in materials with a low resis-countries have indicated an interest in exchanging
]
tance to n fadute mode termed ductile tearing and for embnttlemeni data information and possibly establishing shallow flaws hicated in eritical welds. This work was an international data base of this type.
I supplemented by participation in an effon by the Com.
i mittec on the Safety of Nuclearinstallations (CSNI)Prin-i cipal Working Group 3.
llecent results from test reactor irradiations suggest that the ASMii Code approach to shifting the fracture tough.
As part of the CSNI effort Oak thdpe Nationallabora, nm mnes to apunt for irradiation damage may not r
tory (OltNIj organi/ed an international workshop to P wly account for that damage. It appears that the k
Ci"&,sproc um may underpredict the actual shilt m the
(
compare pre [hetions to experimental results for the frac-ture behavior in large scale experiments conducted in fract ure toughness curves. crodmg the anticipated margin vanous laboratories around the world. The overall results of safey in many apulatoy analyses. At the same time, 1
of this effort suggested that the analyses were not alwass otha aspats of the overall pressure vessel integrity i
as accurate or conservative as believedflhis result led the anaps am knen to be extremely conservative so that NitC to increase the emphasis on refmements to the the hnM naluanons stdl are quite conservative.
fracture analyses dunng FV 1990.The work is expected to continue for i to 2 years, culminating in another analysis The embrittlement research program has provided initial -
i demonstration. The exp.mded program builds on the on-data to demonstrate the effectiveness of thermal anneal-l going work at OltNI, and the David Taylor 1(esearch ing in recovering degradation in mechanical properties j
Center and brings expertise at the University of Illinois, due to irradiation damage. The results of this research j
the University of Maryland, the University of Tennessee, have identified the optimum anneahng temperature and and the Umversity of Kansas to bear on the problem.
annealing period, at least for the relatively limited num-l Other private scetor work is expected to be included to her of materials studied to date.The results of the anneal-l provide the best technietd solution in a timely manner, ing work have been supported by industry efforts and by j
results from research performed in the USSit and ex-As the fracture analysis technolog has matured, the em.
changed under the auspices of the Joint Coordinating l
phasis in NitC's research has moved from broad spectrum Committee on Civilian Nuclear 1(cactor Safety.The com.
i scopmg research to research aimed at developing analy-bined resuits of these efforts provide reasonable assur-ses and the supporting data that can climinate some of the ance that thermal annealing is a practical rnethod for very conservative assumptions incorporated in the early mitigatmg the effects of irradiation damage. While much i
regulatory analyses. Work during FY 1990 at OltNI. used more work is needed to provide appropriate regulatory probabilistic fracture mechanies analyses to evaluate the guidance, the principle has been proved by the embrittle-benefits of further research into high values of crack ment research program.
NUl(FG-1266 2
- - - - - - - - ~ - -. - - - -
1.13.2 Piping integrity toughness, i.e., the lowest toughness that would ever be achieved for the material after long-term senice, or the The pipework of reactor primary systemsis not subject to
" service time" toughness, i.e., the fracture toughness for degradation by irradiation effects, but rather is more m' any given time and temperature of service.
fluenced by fatigue, corrosion, and thermal aging. The NRC's regulations and regulatoiy guidance concerning The NRC's pipe fracture research has involved both piping integriy are designed to limit darnage due to nor-analysis and full-scale pipe fracture experiments. During mal operations-fatigue and corrosion. for example-yet FY 1990, the major emphasis has been on conducting include leak detection and fracture analyses to ensure the pipe fracture experiments using a representative pipe-integrity of the piping under normal and accident condi-loop configuration tested at typical seismic loading rates, tions in ihe event that some form of damage has occurred.
These experiments have demonstrated the adequacy of i
These are explicit applications of the defense-in-depth the NRC's leak befor e-break analy ses, at least for cracks l
concept and have led to highly reliable piping systems in in straight pipe.The results show that the effects of scis-spite of unanticipated damage such as intergranular stress mic loading rates are adequately addressed by the existing corrosion crackmg in HWRs.
analysis meihods and material property data, This was a major step in providing the complete validation of the The NRC's piping integrity rescaich addresses all aspects NRC's fracture analysis methods for cracked piping, of piping safety analyses. Work during FY 1990 evaluated the effects of the (vater coolant on the fatigue life of 1.133 Inspection Procedures and Technology typical stainless steel piping and the effects of seismic As cracks form and grow (for example, fatigue cracks),
loading on the fracture behavior of cracked piping. '1he energy is released in the form of acoustic waves that travel results from the fatigue research clearly indicate that the through the structure. This " acoustic emission" (AE) can effects of typical reactor operating temperatures and the be detected by transducers mounted on the component coolant environments can shorten the fatigue life of pip-and can be h)cated throuch triangulation, as well as evalu-ing and piping components sW4antly-in some cases ated for quantifying crack severity.The NRC has devel-
~
the environmental effect absorin the entire margin that oped this technology to the point where the ASME com-was included in the fatigue dengn curves. C, ooperative pleted in April 1990 a code case for the use of the national and international efforts are under way to de-technology in monitoring crack growth in operating reac-velop the data necessary to evaluate the problem and to tors. The technology has now been used for several years justify changes to the fatigue design curves and overall to monitor an intergranular stress corrosion crack fatigue design procedure.
(IGSCC)in a nonle to-safe-end weld of the Philadelphia Electric Company (PECo) Limerick Unit I reactor. Since The f racture research has involved both material prop ~
the crack has been judged small enough to remain without erty evaluations and pipe fracture testing and analyses, repair, PECo and N RC decided to allow acoustic emission the material property testing has emphasized evahiation to monitor the crack to see if it grew significantly during of the effects of operation on the fracture toughness of the next fuel cycle. Results compiled this year show the cast duplex austenitic-territic stainless steels. These ma-AE-detected crack extension at the tips of the IGSC terials are used extensively m the nuclear mdustry to crack, which would be expected, and additional indica-fabricate pump casings and valve bodies for LWRs and tions at other locations otherwise not pointed out by cone primary coolant pipmg m PWRs. Recent investigations ventional ultrasonic testing. Following the current oper-conducted at Argonne National laboratory suggest that ating cycle, ultrasonic testing will again be used to verify embrittlement of the ferritic phase m these materials may any additional crack extension detected by AE. Eventu-occur after 10 to 20 years at reactor operating tempera-ally, AE may be able to provide unambiguous monitoring tures, which could influence the mechanical response and of hard-to-reach locations, providing an additional level integrity of pressure boundary components durmg high of safety for operating plants.
strain-rate loading. e.g., seismic events.The prob!cm is of -
most concern in PWRs where slightly higher tempera-In order to improve the inservicc inspection (ISI) program tures are typical and cast stainless steel piping is widely for operating power plants, NRC research has been used. Research on this subject has been ongoing since evaluating the impact of ISI unreliability on system safety 1982. During FY 1990, procedures and correlations for end is evaluating the reliability required to ensure suitably estimating fracture toughness of cast stainless steels in 1;w failure probabilities. In FY 1990, pilot application of LWR systems have been developed. Lower-bound esti-probabilistic risk analysis (PRA) for several PWRs and mates of fracture toughness can be made for cast stainless HWRs was completed, and a ranking of systems' impor-steels of unknown chemical composition or for generic tance to safety was developed for tailoring the inspection -
evaluations; more accurate estimates are also provided requirements to system importance. Recommendations and c4m be used when some information is known about on system ranking, NDE reliability, and the use of proba-the material, e.g., when the certified materials test record bilistic fracture mechanics were made to the ASME Re-is available.The correlations can predict the " saturation" search Task Force on Risk-Hased Inspection Guidelines, 3
l and a guidelines report was prepared on the use of these 1.
Identify and characteri/c aging and service wear ef-technologies for d;veloping improved insenice inspec-fects that, if unmitigated, could cause degradation tion criteria, includmg the compments to be inspec!cd, of structures, components, and systems and thereby the frequency of inspections, and the reliabihty of tech, impair plant safety.
niques required.
2.
Develop methods of inspection, surveillance, and monitoring and of evaluating residual life of struc-1.2 Agirig of Reactor Components tures, components, and systems that will permit compensatory action to counter significant aging ef-
"~
1.2.1 Slalement of Problem 3.
livaluate the effectiveness of storage, maintenance, Aging affects all reactor components, systems, and struc-repair,and replacement practices, current and pro-tures in various degrees and has the potential to increase posed, in mitigating the effects and diminishing the risk to public health and safety if its effects are not con-rate and the extent of degradation caused by aging.
trolled. In order to ensure continuous safe operation, measures must be taken to monitor key components' 1.2.3 Research Accolnplishments in FY 1990 systems, and structures and interfaces to detect aging degradation and to mitigate its effects through mainte-1.2.3.1 Aging Research nance, repair, or replacement. I or an older plant ap-proaching the end ofits design hfe and for which extended liased on the review of operating experience, including operation beyond its original license period of 40 years is the available data base, expert opinions, and interactions contemplated, aging becomes a critical concern and will with codes and standards committees, Phase I aging as-clearly be crucial to any assessment of the safety implica.
sessments were completed on the following special topics tions of license renewal, and safety-related ccmponents and systems:
Recently, the nuclear industry has initiated a significant 1.
Ileat exchangers, effort aimed at extendmg the life of existing plants beyond 2.
Compressors, their original term of 40 years. According to a Depart.
3.
Transformers, ment of Energy study, the projected net benefit to the 4.
Control rod drive system (llWR),
United States economy can be on the order of $230 billion 5.
Residual life assessment of major LWR components through the year 2030, assuming a 20 year hfe extension and structures, and for current plants. If a 40-year life extension is judged 6.
Review of data needs and recordkeeping.
feasible, the benefit is even larger. 'lhe benefit reflects both the lower fuel cost of the nuclear plants and reduced Reports were issued on the above-mentioned Phase 1 outlays for replacement of generating capacity. Utilities aging assessments to identify degradation sites within the are currently planning to apply for license renewals and component and system boundary, aging mechanisms, and have a tentative schedule for several steps in the process.
aging concerns."1 he reports, which also made recommen-The first submittal to the NRC is expected in 1991, with a dations for maintenance and aging mitigation, were re-large number of additional submittals to follow shortly viewed by industrial as well as professional society groups.
thereafter. To keep pace with these industry plans, the NRC will need to devote effort over the next several years i
to license renewal. A firm NRC policy on the term's and phase 2 aging assessments of components involve some i
conditions of license renewal applications will be com.
combination of (1) tests of naturally aged equipment or pleted by early 1991. Review of these applications at an equipment with simulated aging degradation; (2) labora-4 early stage will provide an indication to the industry of the tory or in-plant verification of methods for inspection, viability of the life extension option in sufficient time to monitoring, and surveillance; (3) development of recom-elect an alternative option if necessary.
mendations for inspection or monitoring techniques;(4) verification of methods for evaluating residual service lifetime;(5) identification of effective maintenance prac-1.2.2 Program Strategy tices; 6)in situ examination and data gathering for operat.
ing equipment; and (7) verification of failure causes, using NRC staff effort in aging is being pursued in several areas, results from in situ and post. service examinations. During -
including technical and scientific research to identify the 1990, Phase 2 aging assessments were completed on the effects of aging on the key safety-related compments of following components:
the plant and to examine methods for mitigating such effects. Specifically, the strategy is to achieve relative to 1.
Motor-operated valves, each component the followmg results:
2.
l
_= --
1 i
l 3.
Snubbers, and A draft regulatory guide was developed on the standard 4.
Service water systems.
format and content of technical information for applica-tions to renew nuclear power plant operating licenses.
The purpose of the regulatory guide is to establish a l.23.2 Residual I.ife Assessment of Major IM R Components unknn format and content acceptable to the NRC staff for structunng and presenting the technical information Intrinsic to the general exp' oration of reactor aging is the to be compiled by an applicant for a renewed nuclear residual hfe assessment (RI.A) of major components and Imwer plant operating beense and submitted by the appli-structures. The capability to predict the residual opera.
cant as part of an application for a renewed license. The tional hves of major light-water reactor (LWR)compo.
regulatory guide identifies the content of, and provides nents and structures can be indispensable to resolving technical criteria for, Ihe com piled technical information.
technical issues associated with plant aging and license renewal. As of IT 1990, the RLA of 18 components and L2,3.4 PRA.llased Prioritiration of Risk structures important to plant safety have been com.
Contributions and Maintenance pleted. The components are reactor pressure vessel sup-ports, reactor coolant pumps, PWR pressure vessels, A report (draft NUREG/CR-5587)wasissued on setting PWR containment structures, PWR coolant piping, PWR pnonnes for nsk contributions and maintenance impor-steam generators, PWR pressuriiers, PWR pressure tance of aged active components based on probabilistic surge and spray lines PWR reactor cooling system charg-risk assessments (PR As).The report (1) describes an age.
ing and safety injection nonles, PWR fecdwater lines, dependent PRA. based methodology, (2) describes the PWR control roJ drive mechanisms and reactor internals, importance of using risk analysis and PRAs to set priori, llWR containments, llWR feedwater and main steam ties for aging contributions,(3) demonstrates the nced for hnes, ilWR control rod drive mechanisms and reactor incorporating aging effects into maintenance prioritiza-inte rnals, electrical cables, and emergency diesel genera.
tions, (4) demonstrates the ways in which the methodol-tors. I n these assessments, the degradation sites, degrada.
ogy can be applied to setting priorities for components tion mechanisms, stressors, and failure modes have been and systems on the basis of their risk contributions and identified for each component and structure under study.
associated maintenance importance, (5) demonstrates The assessments also include a review of the current the ways in which effective, risk-based aging management methods for inspection and surveillance of these compo.
programs can be identified to control the dominant risk nents and structures. The results of this effort have been contributors, and (6) summarizes a procedure for priori.
documented in NURiiG/CR-4731, Volumes 1 and 2.
tizing components with regard to risk contribution and mamtenance importance.
Using the information reported in Volumes I and 2 iden-tified above, the work completed in FT 1990 focused on The demonstrations show that aging effects (1)if unmiti-developing models and procedures for estimating aging gated, cause m.ereases m plant risk,(2) cause component damage in specific 1.WR components for continued safe phitks ty change as compared to the baseline i R A, and operation. The work included the evaluation of advanced (3) result m signtficant impacts on plant risk and compo-inspection, surveillance, and monitoring methods for nent pnodtics from the interactions of multiple aged charactenzing the aging damage. These results will be 9*"*,I'he prioritization of the nsk impacts and useful for NRC licensing to establish policies and guide-m mienance unponam'c of anc components was dem-lines for making license renewal decisions. The compo-onstrated; prioritized lists of components and systems nents that were assessed or that are currently being as-were developed. Approaches to effective maintenance sessed are LWR reinforced concrete containments, PWR practices, when focused on the components having risk pressure vessels, LWR metal containments, PWR steam and maintenance importance, were demonstrated to con-generator tubes, and cast stainless steel components.The trol the impacts of aging on plant nsk.
results on the cast stainless steel components have been documented in NURiiG/CR-5314, Volume 3.
L2.3.5 Regulatory Instrument Review: Management of Aging of LWR Major Safety.Related 1.2.3.3 Technical llases for License Renewal Eight selected regulatory instruments, e.g., NRC regula-A rulemaking process to fonnulate a license renewal rule tory guides and the Code of f;deral Regulations, were is under way and is expected to lead to a technical and reviewed for safety-related information on three addi-paredurat rulemaking by mid-1991. llesides a final rule, tional major LWR components-cables, containment, more detailed regulatory guidance addressing the techni-and basemat. The focus of the review was on 25 safety-cal safety issues related to aging is needed, both to imple-related aging issues, including examination, inspection, ment the rule and to aWise licensees on license renewal and maintenance and repair; excessive / harsh testing; and application requiremenu.
irradiation and thermal embrittlement, it was concluded 5
that safety related regulatory instruments do provide im.
late component degradation to reliability and risk im-plicit guidance for aging management but that there is pacts. When the degradation indicators become severe room for improvement with regard to explicit guidance.
enough, implying significant impacts on the component -
failure rate and resulting risk, then maintenance needs to 1.2.3.6 Inspecthm Integration be performed to correct the degradation. Thus, the degra-dation indicators can provide practical and effective lhe Nuclear Plant Aging Research (NPAR) program has means of monitoring conditions. Furthermore, by moni-the potential to support the ongoing inspection effort toring the effectiveness of maintenance, degradation can conducted by the regions in accordance with the NRC be corrected before havmg significcmt implications for inspection program. One objective of the inspection ef-reliabihty and risk, fort is to ensure that safety systems and safety-related components base not degraded as a result of any cause, 1.2.3.8 Coraponents, Systems, and facilities includmg aging.
Check rake failures at nuclear power plants have been a source of concern to the NRCJihe major causes of swing The NPAll results and their potential use were presented check valve failure are degradation due to hinge pin wear to inspectors in NRC Regions 1. II, and !!!. *lhe major and fatigue of the discstud connection to the hinge arm. A segment of the presentation provided specific examples matrix of accelerated wear tests using aluminum hinge of NPAll research results and how they can be applied to pins and bushings in 3. and 6. inch typical check valves was inspection activities. llased on the feedback obtained completed. A special disc, instrumented with strain gages, from regional and resident inspectors, it was concluded was used to measure the rate and magnitut af impact that the use of NPAR results can improve the inspection forces. Using these experimental results, thes talwear process. The inspectors are closely involved with techni-and fatigue prediction models were develop assess-cal issues at the plant, and they generally agree that ing the risk of valve failure and need for m
- ace, equiptr.cnt and system aging is an issue that should be addressed by alllicensees.
The Oak Ridge National Laboratory (ORN1 ) completed the assessment of acoustic emission, ultrasonic, and mag-Supported by the information obtamed from the inspec.
netic flux monitoring systems for detecting check valve tors, NPAR reports for selected components and systems wear and malfunction using a flow loop. They were able to were reviewed, and information was extracted that could detect valve disc position, motion, wear, and seat leakage be of use to inspection activities. 'lhis information was using a combination of magnetic flux and either acoustic compiled into two documents called an " aging report emission or ultrasonic detection. It should be possible for summary" and an " aging inspection guide " The summary nuclear power plants using these techniques to perform for ca:h equipment type and system studied in the NPAR non-intrusive inser ice testing of check valves without program includes the identifwation of aging.related prob.
disassembly and examination, lems, highlights of the operating experience, solutions t Cable research on the life extension and condition moni-aging problems, and references likely to be available to toring of electric cable continued at Sandia National the mspectors lhe aging inspection guide is a more con-Iaboratories (SN! ) and the National Institute of Stan-cise summary of recommendations emphasizing visual dards and Technology (NIST). The use of Time Domain inspection techniques and activities to evaluate the licen see s programs for understanding and managmg agmg. '
Spectroscopy and Partial Discharge Techniques for con-ditica monitoring was evaluated at NIST and found to be a powerful tool for identifying and locating defects or 1.2.3.7 Degradation Modeling of Components degradation points in electric cable insulation in a labora-nment, but tp@techn$cd no1 appear to ory en Theoretical models were developed that define a degra-or use udey m ant p conh,onshas-pmc dation rate of class of components and use it to predict its umm so cetnc ad phsd pmpeMsam-failure rate.The data base used in this effort was obtained from plant maintenance records and was composed of all ples of the insulation from electric cables exposed to aging and accident qualification testing for a simulated 20, plant activities on the components, meluding inrpections, 40, and 60 year aging period were completed by SNL The testing, surveillance, monitoring, and maintenance.
insulation resistance and dielectric properties are being 1 hose events indicating some kind of degradation in the correlated with tensile strength and elongation measure-component and those requiring immediate maintenance mants to determine whether it is possible to predict po-to restore the component back to operation were bmned tential cable failure from aging or accident stressors.The toget her for this mode hng cf fort.The changes in degrada.
resuhs of the SNL cable tests are being used by the NRC -
tion rate with the age of the component were developed from degradation data on similar comp (ments.17ailures of -
n evaluating the feasibility of extending the qualified life of cables beyond 40 years for plant license renewal.
components were also analyicd to model how this rate 4
l changes with the component's age. The methodology al-Relays and circuit breakers are being studied to determine lows the development of maintenance indicators that re-the most effective testing and monitoring methods that i
NURIiG-1266 6
i 1
m
_..,,.. ~
could detect the ape-related degradation within a relay or for a senice water leakage in a diesel generator heat a circuit breaker. A number of non-intrusive test methods exchanger tube at a commercial nuclear power plant.The were considet ed for this effort. Ilaseline testing for relays conclusion from the analysis was that mlermittent flow and circuit breakers has been completed. At present, rates, above specification, through the tubes permitted a intentionally degrading of the components to assess the stepwise corrosion of the tubes. 'lhe corrective action, enantes that occur in the monitormg parameters is in after the heat exchanger was repaired, was to control the progress. Developed techniques have been tested at two senice water system flow rate to the specifications.
nuclear power plants with promising results.
Structuralcomponents are being studied under the strue.
l 1ife testing of a Westinghouse DS-416 low voltage air tural aging (SAG) program. This program generated the l
circuit breaker was completed in FY 1990. One of the following four reports in FY 1990: "Five-Year Compres-l most important uses of this class of breakers is in the sive Strength Results for Moist Cured and Scaled liigh.
reactor protection system for scram applications. In addi.
Strength Variable Fly Ash Concretes"rStructural Aging tion to the normal detenoration of the contact assembly Program Annual Technical Progress Report for FY and nre chutes, a number of subcomponents of the power'.
1989"!" Structural Component Significance Assessment operated mechanism exhibited degradation with the Methodology for Concrete Structures in Nuclear Power breaker test cycles. Maintenance and monitoring recom.
Plants", and " Structural Materials Information Center mendations were provided based on the results of this for Presentation of the Time Variation of Material Prop-researcb.
erties." An annual technical reporting system was imple-mented, and formatting of the Structural Information Simbber research results provide inservice information Center was completed. Also completed were the con pertaming to recent operating experience for both hy.
crete comp (ment classification system and assessment of draulic and mechanical snubbers, particularly in regard to nondestructive examination techniques. Initial probabil-aging-related influences. Methods were identified that istic models were developed to assess time-dependent are usefulin monitoring the service life of snubbers. 'lhe reliability and deterioration of reinforced concrete struc-pnneipal fmdmps of this research are:
tural components subjected to stochastic loads.
'lhe primary environments that contribute to aging Shippingport cast stain! css componcnts were studied in the e
degradation in snubbers are temperature, vibration, laboratory, Most laboratory studies of structural materi.
and dynamic transients.
als pertaining to degradation in nuclear reactor compo-nents involve accelerated nging and/or irradiation condi-Hased on the eight nuclear power plants investi.
tions to simulate end-of-life microstructures within a e
gated, upproximately 47 percent were mechanical short time span since the time scale for operations of a functional ten failuresand 52 percent of test failures power plant is far longer than can be generally considered wcre senice related.
for laboratory studies. Thus, an assessment of the end of-hfe condition is almost always based on an extrapolation 1Iydrauhe snubber seal life is primarily a function of of data from accelerated test conditions.The validity of operating temperature. Seal life limits originally such extrapolation becomes more questionable as operat-proposed by snubber manufacturers are generally ing times become much longer than the laboratory test conservative.
times. A program has continued during this reporting period to determine the effect of thermal aging on the Service water system aging was studied to identify and char.
embrittlement of in-reactor cast stainless steel.
i acterize the principal aging degradation mechanisms, to Microstructure and microhardness examination of the assess their impact on operational readiness, and.o pro.
cast materials from in-reactor seniec indicated that tha vide a methodology for the mitigation of aging degrada.
mechanisms of thermal aging embrittlement are the same tion. This was accomplished in FY 1990, as those of laboratory aged specimens. Consequently, data obtained from laboratory studies under accelerated The primary degradation mechanism in service water sys.
aging conditions can be extrapolated to predict insenice tems is corrosion compounded by biological accumma, degradation at reactor operating temperatures.
tion.The most effective means for mitigatingdegradation in these systems is to pursue appropriate programs to 1.3 Reactor Equipment Qualification effectively control water chemistry properties when pos-sible and to use biocidal agents where necessary.
1.3.1 Statement of Problem To satisfy the need for a formal procedure to identify the As a result of the Three Mile Island (TMI) accident, cause of age-related degradation of senice water systems, concerns and questions were raised regarding the oper-l a root-cause method of analysis was developed.This root-ability and structural integrity of components during cause analysis method was used in a survey of the cause carthquake and loss-of coolant accident (LOCA) 7 NUIEG-1266 -
l l
i
_ ~ _ _
environments. Altl.ough design criteria arid loadmg defi-ASMi! quahfication standards and O&M standards to nitions have changed over the years to improve the integ-provide the basis for ensurmg safer components.
rity of these components, the concerns and questions dealt dir ectly with the adequacy of the component quahfi-L3.3lh d Mc libre in IT 19%
i cations lherefore, those stems that were identified as i
high prionty were phen immediate research attention lixperiments were conducted in late FY 1989 and early and action. It was also intended that the results of the FY 1990 to determine whether valves in high-energy rescatch would be incorporated into standards-pipes will close as they should to prevent leakage during a pipe-break accident outside the containment.The result.
Subsequent to the TMl research activity, other safety ing high velocity flows that develop in the pipe and in the issues were iJentificJ and, wbere thcae had impact on valves must be stopped by the valves. If unchecked, with equipment qualifications, research effort was proposed to vahes that do not close, the leakage can cause serious develop the data base to aid in the resolution of these consequences, not only because of steam release outside high priority safety problems Current effort is address.
containment, but abo because other emergency equip-ing one of these generic safety issues. Another effort is ment may be exposed to the harsh water and steam envi-providing guidehnes for improving valve qualification ronment and may fail,
>tandards.
A total of six different valves were tested, thr;e cach 6-inch diameters and 10 inch diameters. The 6-inch di-1.3.2 Program Strategy ametervahesare typicalof thoseinstalledin high energy hot water pipes while the 10-inch diameter valves are One main NRC staff effort in the equipment qualification typical of those installed in high. energy steam pipes, All program is currendy invoh ed with developing the techni-valves were typical of those in actual plants and the fluid cal data base for addressing a high-pnority generic safety environments (flow vehicity, pressure, temperature) were issue (GSI-S7) related to the operability of motor-oper-selected to simulate actual conditions that would occur in ated valves (A10Vs). This same effort is also providing the event of an accident such as a pipe break at some irnportant mformntion and guidance for meeting some of nuclear plants, the requirements of Genenc I etter 89-10, which is re-lated to GSI it.En 1.This u or k is resulting in understand-I' valuation of the results of the first series of experiments ing the capabihties of diagnostic equipment to predict conducted on the 6 ir.ch diameter valves showed that two whether thrust measurements can be extrapolated from of the three valves were capable of stopping the flows in typic;d m situ test conditions to accident level conditions.
all the closing experiments. The valve that did not stop Past and current results will contmue to be incorporated t he flow entirely had experienced significant damage to its in the ASME vahe qualification standard and in the op-internal parts, due to the action of the flow during clo-erations and mamtenance (OA M) standards. These up-sure, such that the actualor could not overcome the resist-graded standards wdl clarify some of the areas the staff ing forces caused by the _ damage, believes may be contnbutmg to MOV problems.
Evaluation of the second scriesof experimentsconducted Future effort in the equipment qualification program will on the 10-inch diameter valves in the steam environment be devoted to evaluating test data to understand the be.
showed that all three valves were capable of stopping the havior of other typic;d valves in high energy piping sys.
flows in all the closing experiments Ilowever, one valve tems subjected to accident Gow environments and in experienced significant damage to its internal parts.
other piping systems w hen subjected to operational flows.
Other effort wdl address the integration of the test data It is important to understand that, prior to beginning for regulatory applications and for resolving new prob.
these experiments, the valve actuators that were installed lems und safety issues consistent with safety and licensing on the valves wcre set to deliver larper than normal i
needs. Since industry is expected to bec'ome more in.
thrusts to achieve closure so that the various vahe inter-
~
{
volved in solvmg some of the pressing valve safety prob.
nal forces could be quantified. Obviously, the one valve j
lems, some NRC effort will be devoted to following this that did not reach closure was da naged and the required j
work and evaluating the results with regard to licensing thrust was underestimated.
I applications. Another research effort, which has been delayed because of other high-priority GSI work, will be General observations and findings from all experiments i
devoted to understandmg the effecis of large earthquake ar e:
loads on the operability of an aged gate valveJIhe effects i
of the large dynamic loads on piping, supports, snuhbers, 1.
The main (netion force that m ist be overcome by l
and anchors will also be studied. The results from current the valve actuator is underpredried when the typi-l and future efforts will be incorporated in appropriate cal friction factor is used; i
l NIJR EG-1266 8
l l
l r
~,. _ _
2.
None of the valves would have stopped the flows for The publication of seismic hazard curves in 1989 by both all the experiments if the recommended valve ac-the NI(C (NURl!O!CR-5250) and itill (NP-6395) taator settings had been used; marks the end of major efforts to characterize the seismic hazard at U.S. nuclear reactor sites, Although the best 3
The equation used to size the actuator underpredicts information and procedures available were used, they the required closing thrust for all sis valves: and rescaled that large uneettainties still remain in scismie hazard estimates. Also, recent full-scope probabilistic 4.
Some damage such as wear marks and galling oc-risk assessments, performed as part of the NUl(EG-ll50 curred on the internal parts of all valves; however, effort, continue to show that seismic hazard uncertainties two of the valves (onc b-inch diameter and one contribute significantly to the overall uncertainty in nu.
10 inch diameter) experienced more - significant clear reactor risk estimates 'these large uncertainties damage.
make it difficult to place the contribution of seismic risk into its proper perspective, e.g., in the development of
'lhe results of the experiments were presented to repre-individual plant examination guidelines.
sentatives from valve, valve actuator, and diagnostic equipment manufacturers; from utilities; and from re-llecent successes in the geological, geophysical, and seis-search laboratories,ihese expert representatives found mological studies sponsored by 1(l!S and discussed below the resuhs generally acceptable and agreed that the data show that it is possible to answer the basic scientific ques-contribute to an understanding of valve behavior. Ilow-tions that underlie these seismic hazard uncertainties. It is ever, the experts believe that the reasons for the unpre-the goal of the 1(IIS carth science program to significantly dictable valve behavior must be obtained through evalu-reduce the uncertainty in seismic hazard estimation in the ating the internal valve dimensions and tolerances. The next decade through emphasizing this type of research.
N1(C accommodated this request and the measurements were performed on the valve internals. These measure-In the 197b's and before, our interest in nuclear plant ments are currently being evaluated by the experts, scismic design was mainly limited to response at design levels (e.g., OllE and SSli) and our knowledge of this was The results, including actual test data, from these latest primarily based on analytical techrJques and assump-experiments have also been made available to all inter-tiens. In the 1980's, a considerable effort has been made ested parties for their use in improving the reliability of to better predict the potential response of nuclear plants these vah es. The N1(C licensing office is using the data to to earthquakes greater than those considered in design.
assess vahe closure capabihties in operating plants. The Our understanding has been increased greatly by the test-NI(C contractors will continue to evaluate the test data ing to failure of equipment and structures, by the gather-for further information, and the results will be made ing and synthesis of earthquake experience data from available to the operating plants.
non nuclear facilities and by the large number of seismic probabilistic risk assessments that have been made.
1 A Seismic Safety This research has generally found that the seismiceapac-ity of important nuclear plant structures and equipment 1.4.1 Statement of l'roblem (when properly anchored)is high. Ilut there remain spe-cide capacity concerns that need to be resolved, such as
!!arthquakes are among the most severe of the natural how to address the potentially harmful effects of relay hazaids faced by nuclear power plants. Very large carth-chatter. The importance of plant-specific walkdown re-quakes would simultaneously challenge the abihty of all views to find nongenerie vulnerabilitihs has been noted in plant safety systems to function and, coupled with the recent seismic margins studies.
likely loss of offsite power and dependent safety systems, could pose a unique threat to public safety. As with many 1.42 l'rogram Strategy potentially severe conditions, there is much uncertamty associated with the design and evaluation of nuclear The strategy to resolve the scismic preb!cm involves re-plants for earthquakes. Seismic hazard in the Central and search to develop the methods and data that will support liastern United States remains an issue that is not likely to the necessary scismic criteria development and provide be easily resolved.These regions contain the highest per-the evaluation tools. The - research is focused.on centage of nuclear power plants in the United States.
(1) improving estimates of earthquake hazards by identi-Ilistorically, the largest earthquakes in the United States fying potential carthquake sources and determining the have occurred at New Madrid, Missouri, and at Charles-propagation of seismic energy with distance, (2) estimat-ton, South Carolina.The geology of the central and east-ing the possible range and hkelihood of seismic ground ern regions makes it difficult to establish earthquake mag-motions at nuclear plant sites, and (3) assessing the effect nitudes or seismic parameters for specific k> cations or to of these ground motions on soil, structures, equipment, ensure a proper design basis for individual power plants.
and systems of the plants. The integrated results of this l
9 NUlt!!G-1266 i
r esearch will be used to quantify the risk to nuclear plants 1.4.3 Ilesearch Accontplistinients in IT IMO from earthquakes, to assess the seismic safety margins mhetent in current or future plant design, and to help 1,,13,1 1:arth Sdences identify and set priorities for what improsements are needed in plant designs or what parts of seisruic design Seismic hazards contribute a sizable proportion of overall criteria may be related.
plant hatards and, because of inherent difficulties in de-fming them. they form an even more significant portion of the uncertainty in estimating plant hazards. Although A major focus of the NitC research programs in geology, recent N!(C (N URI!G/ Cit-5250)and Illectric Power ite-seismology, and geophysics continues to be identifying search instit ute (liPit!)(NP-6345) studies have ndvanced and defming potential carthquake sources or source the methodology for characterizing seismic hazards at zanes in the !! astern Umted States and using that mfor.
nuclear reactor sites, further seismic hatard research will mation in assessing seismw harmds with respect to nu-be needed.The goal of the lil!S carth science program is clear power plants. Many unknowns exist regarding these to reduce uncertainties in hazard estimates by continued issues including a strong basis for seismic tonation, research into the causes and distribution of seismicity, source mechanisms, characteristics of ground motions, Successes of past research programs together with appli-and site specific response. ~lhe NI(C is addressing these cations of newly developed methods prnmise to signifi.
uncertainties through research that encompasses sus-cantly reduce uncertainties in seismic hazard estimates tained seismic monitormg, geologie and tectonic studies, within the next decade, neutectonic mvestigations, esploring the earth's crust at hypocentral depths, and conducting ground motion stud.
Through a cooperatise agreement with the U.S. Geologi-i ics.
cal Survey (USGS), the NitC is funding the installation of j
seismographic stations in the llastern and Central United States that will form a part of the new National Seismo.
1 The backbone of the NitC program in the liastern United graphic Net work (NSN). 'lhe new network is scheduled to i
States has been the scismographic networks deployed become operationalin 1992, and at that time the NI(C l
throughout the !! astern and Central United States.The will discontinue its funding of regional seismographic net-NI(C is currently funding seismograplue networks in the wor ks.
I following regions: Northeastern United States; Virginia; l
Charleston, South Carohna; the Southern Appalachian The NSN will use state of theyart instrumentation and a s egion: t he New Madrid (Missouri) region; Ohio and indi, more uniform station spacing. llence, it will provide a ana; castern Kansaw and Oklahoma. An agreement was more balanced carthquake detection capability although 7
reached in 1986 hetween the United States Geological at the expense of more widely spaced stations. In the Survey (USGS)and the NitC to jointly support the estah-interim, some of the less critical networks are being i
lishment of the castern portion of a national scismo-phased out, Other regional networks, such as those in the graphic network. The national network is scheduled to be highly active New Madrid area, are being upgraded in fully in place by IT 1992. In the meantime, the currently cooperation wnh the USGS. A critical need is f a pain NI(C-funded networks in the !? astern and Central United better information on the depth of carthquakes and on l
States will be gradually phased out.
three-dtmensional aspects of ground motioa. This re-j quires replacing existing vertical motion sensors of the regional networks with three-component seismometers In recent scars, the N!(C has supported seismic testing and expanding the dynamic range of the instrumentation.
and the collection of earthquake experience data in order the purpose of upgrading is to complement the NSN with to improve and gam confidence in the use of seismic detailed information in critical areas. At the same time, j
PI(As and seismic marpm studies. These data are also the improved networks will be able to take advantage of being used to support proposed improvements to seismic the satellite communication links of the NSN, l
design criteria. The carthquake resistance of structures, equipment, and piping has been found, in general, to be N!(C support for regional seismographic networks cover-l higher than previously thought. Major efforts in this area ing the Central and IIastern United States continued
. bill be completed in 1990, and Ihe results are being suc-during IT 1990 ahhough at a slightly lower level Contin-l cessfully used in licensing actions,1(clay chatter is the one ued carthquake monitoring provides data that are essen-remaining seismic capm. issue that will require addi-tial for comparing seismicity with tectonics in this region.
tional iesting to resolve.
liccause of the moderate-to-lowlevel of seismicity, trends start to ercerpe only after many years of continued carth-quake recording, in the New Madrid area, for instance, Upcoming individual plant examinations and USI A-16 epicenter trends have become apparent that can be corre-seismic reviews will use the recent results of NRC seismic lated with subsurface faulting as determined by seismic research.
reflection investigations.
NU RiiG-1266 10 i
-7
An earthquake of interest was recorded by the Geological f lowes er, the possibility of a smaller magnitude, but dam-Survey of Ctmada, with which the NRC has an agreement agmg, canhquake should not be ovalooked, to cooperate m seismic audies. The magnitude 63 carth-quake occurred in the northern Ungava peninsula on lixpanding on the knowledge pained by this study, a new December 25,1989. Unusual for eastern North America research project has been initiated to search for evidence is the fact that a surface rupture associated with the fault for prehistoric moderate-to-large carthquakes in fluvial was identified. Reverse fault movement occurred on the deposits along rivers in the southeast such as the Savan.
I ac Turquoise fault over a length of 8.5 kilometers with a nah River. Researchers on this project will also search for maximum vertical displacement of 1.8 meters.1his event and analyze palcoseismi, evidence m the casternTennes-presents an outstanding opportunity to compare surface see seismic zone Such evidence will include scismically faulting data with aftershock distribution and waveform induced paleoliquefaction features, fault offset, land-modeling results.
slides, and geomorphic eviJence.
The New Madrid (Missouri) area experienced an carth-Palcoseismic studies in the epicentral areas of rnoderate historic earthquakes in the northeastern United States quake sequence in 1811-1812 that meluded the most severe shocks ever generated in historic times east of the and adjacent Canada have found paleoliquefaction fea-Rocky Mountams. l'oday the area is still the source of tures induced by a prehistoric carthquake and have pro, vided data on seismically induced soil deformation struc-C".nsiderable carthquake netmty. lhe source of the scis-tures that can be used throughout the northeastern
!nicity has been identified as reactivated faults within a rift
'" the crystallme basement. The presence and extent of United States to distinguish between scismically induced th.is ancient rift structure has been defmed by geological features and similar features induced by other phenom-and particularly by geophysical means. Plots of epicenters
- ena, in this area also clearly reveal the trends of the underlying rift structure.
One study has been completed in the epicentral area of the 1727 Cape Ann earthquake in northeastern Massa' A strong lineament with evidence of faulting along its chusetts. One prehistoric event has been identified using trace was identified by satchite photographic analysis.
geological evidence. A new project has just begun t Field investigations of this fault, the Hootheel fault, are expand this research to other areas of New England that imder way to define its nature and its relationship to the have relatively high seismicity' New Madrid seismic zone.
The 1866 Charleston, South Carolina, carthquake is the Falcoseismic studies along the Mccrs fault were com-largest seismic ever.t to have occurred on the Atlantic pleted in OctoLcr 19S9. The Meers fault is a part of the seaboard of the United States. This carthquake caused northwest-striking Wichita frontal fault system that ex-extensive liquefaction in its meizoscismal region. Re-tends more than 700 kilometers across south-central search in this area also revealed that there have been at Oklahoma and the Texas panhandle. It is one of the few least five prehistoric carthquakes in the Charleston faults cast of the Rocky Mountains that is known to ex-carthquake area. ily analyzing paleoliquefaction-evi' hibit evidence of late Quaternary displacement. In spite dence, it was determined that these events occurred of this, the Meers fault is relatively aseismic. Detailed j
l 600 1 100, 1200 1 100, 3200 1 200, 5150 1 500, and one trench logging and geologic mapping indicate left-lateral event more than 5150 years before the present. The oblique slip, down to the southwest, on a steeply north-carthquakes that occurred 1200 and 3200 years ago were west dipping to nearly vertical fault, Analyses of these the strongest because they caused liquefaction not only data and radiocarbon dating show that there have been at neat Charleston but also in the northern and southern least two, and possibly three, surface faulting events dur-coastal portions of South Carohna. Deposits susceptible ing the past 5,000 years that were probably associated to liquefaction are present near the coast from Florida to with large-magnitude carthquakes. The latest displace-New Jersey More than 1,000 sites were investigated ment occurred about 1.500 years ago. Analysis of faulted withm this region and no seismically induced liquefaction
- alluvial terraces along the Meers fault suggests that a was identified outside of South Carolina except for one period of qu.escence lasting many tens of thousands of feature in North Carolina just north of the State line. An years preceded the faulting events, event occurred 18001200 years ago for which evidence was found north of the Charleston meizoseismal area but After completion of the Mccrs fault investigation, a study not near Charleston Either this represents another seis-of the Criner fault, another member of the Wichita fron-mie source or the evidence at Charleston has not been tal fault system, was begun. Geological reconnaissance found as yet. 'Ihe evidence for large prehistoric carth-and studies of aerial photographs suggest that the Criner
. quakes in coastal South Carolina and the lack of it else-fault may also have cxperienced iatc Quaternary displace-where on the Atlantic coast is consistent with a unique ment. It is downdropped to the southwest and is located source in the Charleston area.The probability of a repeat about 80 kilometers southeast of the Meers fault. Mor-of the 1886 carthquake at Charleston is relatively low.
phologic evidence suggests young displacement, and 1I NUREG-1266 l
)
crosscutting sciationships between the fault, and late determine whether there materials were deposited by Quaternary terrace deposits at one location surgest that tsunamis following the carthquakes. In conjunction with the last displacement occurred hetween 10.0(X) and these st udies, an in vestigation is under way to identify and 20,0(X) years ago. Detailed investigations similar to those define seismically induced paleoliquefaction features in carried out on the Meers fault are being conducted to.
this region.
assess the scismic hazard potential of the Criner fault.
"Ihe size of the maximum or characteristic carthquake Southeastern Illinois has had seven significant events dur-that a fault can produce and the location of that carth.
ing the 200. year historical record. There has been consid-quake along the length of a lault are major means of crable debate on the geological structures responsible for estimating design ground motions. Fault rupture length is this seismic activity, but none has been identified with a key parameter for constraining the slic of future earth-confidence, l'or example, the continuity of the seismicity quakes on a fault,'this constraint can be obtained from belt, along with geophysical evidence, has led to the inter-studies of the rupture length versus magnitude or energy 4
prctation that the fault system in southernmost Illinois release of previous carthquakes. During the past decade, and Indiana is a northeastern extension of the New fault segmentation has emerged as a field of earthquake j
Madrid seismie zone. An alternative explanation is that research that has important implications and applications the earthquakes originate in a complex transition zone for evaluating seismic hatard. It is based on the common connecting two tectonic regimes. Partly because of this observation that fault zones, especially long ones, do not lack of knowledge about a causative mechanism of the ruptum over their entire length during a single carth.
earthquakes, it is not known if much larger carthquakes quake. A variety of structural and palcoscismic studies j
have occurred in the recent past (llolocene) and might and investigations of historical carthquakes clearly indi-reasonably be expected in the future.
cate that the location of rupture is not random, that there 4
are physical controls in a fault rone that define the extent j
Preliminary investigations have identified nume ous sus-of rupture and divide a fault into segments, and that j
pected palcohquefaction features along the Wabash segments can persist through many seismic cycles. Inhcr-i 1(iver Valley. These features are currently being mapped ent in the concept of segmentation is the idea of persis-and analyzed.
tent barriers that control rupture propagation. The recog.
nition and identification of rupture segments have the
'Ihe Pacific Northwest is underlain by the Cascadia sub-potential to provide new insights into characteriting scis-duction zone, in which the oceanic Jutm de Fuca plate is mie sources and understanding conttrJs of rupture initia-being subducted beneath the North American plate.This tion and termination, i
region is an enigma in that the geological and geophysical j
evidence indicates active subduction, but there have been Segmentation for selected faults is being evaluated using no historic larpe. thrust earthquakes along the plate inter-palcoscismic recurrence data and_information on slip per j
lace, a phenomenon observed in other subduction zones event and slip rate.The data base is small but there are a j
around the nm of the Pacific Ocean, number of faults that have the potential to yield informa-tion on long-term segmentation.The data collection and j
The USGS is conducimg a major study of the geology and analysis currently under way include: (1) timing of the tectonics of Ihis region.The NI(C is partially funding two most recent and prior events along the length of the fault; j
neotectonic research projects of this program, one in (2) slip distribution during the historical event and slip i
southwestern Washington and the other in central Ore-during paleocarthquakes on the same segment (repeated l.
pon. These projects are continuations of investigations similar slip would imply fixed segment lengths; variable j
that revealed geologic evidence surgesting the occur-slip would indicate variability in segmentation); (3) slip rence of several prehistoric and flolocene large carth-rates at different kications on the fault; and (4) structural i
quakes. The evidence _ lies in marsh and shallow marine geology and geophysics of the fault zone, Although some sediments, which indicate several cycles of normal data are available in the published literature, much is stratigraphic deposition abruptly terminated by cata-being obtained from unpublished files and palcoscismic i
strophic events. These events are interpreted by most studies that are in progress. Part of the data collection has researchers to indicate large subduction zone carth-involved field visits for onsite evaluation of published i.
quakes. At least five events are in evidence in southern information and, in some cases, development of new data -
Washington. The ongoing research is to better define the
_ such as slip per event for paleocarthquakes on segments ages of these events. determine their regional extent, and that have had historical ruptures. l(cconnaissance and i
estiraate their recurrence intervals using precise radiocar-field work have been done along the (1) Laguna Salada i
btm dating techniques of subsidence-killed Sitka spruce and Coyote Mountains segment of the'liisinore fault i
trees to reduce the errors inherent in the conventional (Californ ia): (2) the l layward-Itodgers Creek (California) 2 technique of dating. Another study to accomplish this is segment boundary; (3) the transition zone between the the analysis of diatom fossils and sand found on top of north end of the 1954 Dixic Valley surface rupture and i
several buried peat layers. This analysis is expected to the Stillwater (Nevada) seismic gap; (4) the north and -
NUltliG-1266 12 I
i
_,_m.
- ~
south ternarunons of the IWu l I Asnam ( Alperial sur-The stram nctwor k was measmed f or a sceond time in lT fate ruplut e; O) the bouridancs betw L cn the M A Uy.
Im Wah unprmcJ n.
m enu nt proa dm o mJ a bct-
!husand Spongs, and Wann Sprinp sermenh of the ter satelhte mnstcllanon, the an uracy of the measure.
I est kis c r fault tIdaho); and (M potential segment ments was unpros ed and maJe more consstcnt aer oss the bounJanes along the Cartvo Mops e San lierrurdmo octwoik. Accuraaes were achtesed that translate to cr-section el the han Andreas fault (Cahjonuab tors of a lcw tentimeters oser bastlines of thousands of kihunctet s. Af ter at least tht ee sets of measurement are ihe NR(' supports sescral strunp pround motion stuJics anuhib!c, these data w di be analy/cd to arrive at prelimi-relating to both the 1 astern (!mtetiStates and Cahforma.
nary conclusions on crustal stram in the UnitcJ States.
A stuJ) of sod dynanuts at Garner Vaucy incar An/a t includmp at le;nt an upper linut for sinun.1)nectional i
Cahfonna, cinploy s an anay of wide-banJ strong motion effect wdl also be mvesurated because they may inJicate scismometers placed at various depths in boreholes to non;undonn su;un, llow es er, detailed answ er s on crustal p.un inlomunon on sint dynanoes and amphfication of strain are expecttd only after a time span of about 10 carthquake monon. Tins stuJy. perfonned by the Uniser-Sears i
sits of Caldomu at 5anta liarbara, n one of the many research proer;uns th it demorntrate ef fettne cost shar-In aJJition to tlus national netwot k, the NRC also sup.
~
me betwecn'the NRC and other arencies I he study is pods a Mnahn Wat sinun network in Mame Analysis of being fundcd in cooperanon wah th'c l!SGS and the dS.
the two e of posnion measurements obtained in Mame Army Corps of Impmters and with support from the ( l A r far hw Unin n that, wlthm the error mar gins and consid-er liance.
enny that the serocal t iPS wmponent is the least accu-i rate, a consstent and clear pattern of subsiaence has not Although many theortu-;al anJ laboratory studies have
"""f II"" " # * " P
""" U"' two sets of
" *t cations of subsiJence in the castern h
j un estipated the effect of near-surface sod layers on am.
W p
on 01 une, h a o appm that suMenw phhcanon of protmd nionon, scry few dirett measure-ments are asadaNe to confirm predicuons made by the e probably m the range of 5 mdhmeters perycar or m
" P"""I""'d values ol up 10 9 nullimeters per more theoretical methods 't he Garner Valley atrav is hicated between the San Jacmto and San Andicas faults par am thus not subMantiatd in an area of lagh sennacuy. Ihe site is underlam by soil S
&M R
mh and w cm her ed pramtc os er a pramne bascinent at about a 45-meter depth. Senmometers were emplaced at the sur-In addition to t he car th science rescarch discussed above, face and m bordades at various depths rangmy to '20 the NRC seismic research program incluJes several engi-
~
meters in less than a year.125 carthquakes were re-neet mp-orieraed programs to determine the effect of i
corded w ah maputuJes ranpmg f rom 1.2 to 43. Analysis earthquakes on nuclear plant structures and safety syv of the data shows that, at low trequencies, amphhcauon t ems.
from bedroc k to surface n a lactor of sa m er a wide rance of marmtodes. The lower frequenaes ate also those th'tt Since IUSth the N R(' has sponsm ed analy tical and experi-have the highest damage potennal f or engineered struc.
mental research on low rise remf orced concrete shear tures walls, budJmps, and building segments. The research fo-cused on answenng certam structtual conecrns related to Durmg the sunnner of 1490. the 1:PRI clected tojom this (O the adequacy of cruena currently used in design and experunent hy suppottine the Unnersit3 of Cahfornia at analysis, and (2) the abihty of existing nuclear power Santa Haibar'a m deplo[mc a surface airas and an adJi.
plants to wahstand earthquakes greater than considered tional borehole mstranieni at the Gainer Vaucy site.
m the onginal design. At this ume the experimental re-
~
~
sults pencially support the foHowing conclusions:
In IT 1987, the NRC and the Nauonal Geodetic Survey Strength-of.matenals analysis that accounts for estabhshed a poruon of a National Crustal Strain Net-work tovenng the castern two thuds of the Unded States shear deformation can accurately predict the static and ds namic stiffness of shear widls having a nomi-and consisting of 45 stations whose positions are meas-n:d b$se shear stress in the rance of 150 psi. The ac-ured accurately nery 2 years ua the Global Positionmg tual salue is a function of 'the test specimen's System (GPSk The purpese of this network h to provide a different set of data to help analyze the causes of sets-concrete ternile suength, and mien). In addnion to its ongmal purpose, the network In terms of stiffness, similanty between structn es of now forms the backbone of a new GPS survey system for the same sue maJe with 3/8-inch aggregate or the nahon. Because of this, many statewiJe high-microconcrete was demonstrated.
precision networks are hemg tie-J ta n providmp a much greater density of data pomts without extra cost to the The mittal( 1982-l%hi shear wall test data indicated that N R C.
the natural f requencies of dynamic response were 13 NU Rii(i-1266 o
t sigmheantly lower (up to 50 percent) than analytical pic-view of components and tanks in conjunction wah the dictions. As a result ut this imJmp, an evaluation of the m:ugins resiew.
NURiiG-il50 Peath lhittom seismic probabihstic risk assevsment (PR A)was initiated toassess the impact of the The NRC in. house Seismic Marpm Workmg Group and a "frequena reduction" issue. Pichmmary fmJmps base peer review proup (PRG) conusung of outsiJe cimult-shown that the mean core damage freqbency increased ants reviewed the appheahon of both 1 PRI-sponsored approumatch 60 percent because of the condJeration of and NitC-sponsorcJ methoJolon anJ concluJed that i
the frequency reduction issue alone. In addition to the thew methods wm successfully applied to llatsh Umt 1 k
PI( A evaluanon, " design-hke" structural dynamic calcu.
and the destred objectnes were actueved. Iloth the NRC lations with and without the stdfness reduction were Seismic Marpin Working Gniup and the PRG weic ac-made. In many cases, floor response spectra were signifi.
tively mvolved when the analyses were being performed cantly amphfied because of the ficquency reduction, and (participated in several meetings and conducted plant spect'ral peaks were aho stufted down significantly. The widkdown4 reduced hequeng analysis resulted m inereases up to 30 percent m net wall forces and moments. lhese findings in adJition to the completion of the trial application at 1W&
M W deWoped to determine risk may represent a worst case since the latest shear wall tests j
are providmg data that show close agreement between nsights that can be developed from applying marpm tM' analysis and tests. The seismic PR As and design hke cal-culations of two additional plants will be evaluated in IT The Commission-appnwed draft guidance document for l
IWI before the project is completed. The latest shear individual plant exammations of external events has en-l wall test data will be mcluded in these evaluations.
dorsed the use of the seismic component f rapihties devel-oped by limokhaven National 1;iboratory. The compo-Seismic margins reuew procedures were mitially desel-nents included are motor control centers, switchpears oped by the NRC and the I!PRI to determme how large (low and medium voltape), panelboards, switchboards, an earthquake was needed to compromise safe plant op-power suppies, instrumentation and control panels, eration.The method used msights from scismic PR As and transmitters, mdicators, switches transformet s, batteries, test anJ earthquake expenence data to focus the review battery chargers, inverters, motors, and electrical pcne-effort. A distmet advantare of the method (over a seisme tration assemblies. The seismic frapilities are ex pressed in PR A) is that, because the direct uw of seismic hazard terms of medium and standard desiations of spectral ac-curves is not needed, the controsersies and uncertainties celeration capacities toallow the NRC staff to understand associated with them are diminished.The method retains the conser atisms associated with the estimates of seismie the phnt walkdown and mtegrated esaluation of plant fragility. Attention has shifted toward understanding how response associated with a scismic PR A.The end product relay chatter can mfluence ciremt breaker tripping. This of the marpms method is a conditional statement regard-has been iJentified as a safety concern because such a ing high confiJence, low probabdity of failure (llCLPF) scenario can cause failure to start emergency power dur-values for the plant, acciJent sequences. and compo-ing and after an earthquake,leaJing to station blackout.
nents. The plant IICI P1 value can be interpreted as an Test plan development and equipment procurement is acceleratun level at w hich there is high confidence (959) under way to address this issue.
Ihat there is only a low probabihty (59 ) that core damace wdl occur.
A inajor activity in the seismic engineering area during the past 2 years has been the development of guidance for The margms methods (NRC or liPRI) with enhance, con uW tk inNual Mant mungon for gmm y un# ment We GumnWons We A ent mn ments are consider ed to be uable options for implement-
+ ""' dedopment of gmdance has been ac@hieved ing the NRC s Ses ere AceiJent Pohev. Implementation is achiesed through an individual plant examination for ex-through the formation of the lixternal I; vents Steering
~
ternal events (IPlill) to identify and report plant spe-
" *" P. Dw S""
"E"*"
- """ P ment m turn eMa u d a ank subcom eific severe acciJent vulnerahihties initiated by external mittee (among other subcommittees), consisting of NRC; events.
stal.f members with expertise in earth sciences, structural /
i PRL Georpu Power, and the NRC hase completed the mechanical engineerine, systems analysis, and risk and reliabihty analysis.
cooperatne seismic marpms review of flatch Unit 1 (Georgia)in lY 19% While two other plants have under.
The seismic subcommittee issued a report in March 1990.
pone such reuews, llatch is the first ilWR and the first The purpose of the report was to recommend to the plant on a sod site to have a margins review.1 iquefaction lil!SG: (1) the objectives for the seismic portion of the potential at high seismic pround motions and relay chatter mdividual plant exammation for external events evaluation were also new and important parts of this (IPl!!!!!), and (2) guidance or guidelines for conducting review. Georgia Power has performed its USI A-46 re-the seismic portion of the IPlil:li Results and insights NU Rl:G-1266 14
frum various seismic research programs (e.g., seismic (SSI) experiments at lotung Taiwan. The planned SSI margins method and component fragility testing prm studies will be performed at a stiff soil site in lluatien, grams) are incorporated into this guidance. The staff will Taiwan, that historically has had more destructive carth-resohe comments on the draft guidance and will issue quakes in the past than I otung lil'RI has organi/ed the final guidance in PY 1491,(I or further accomplishments, Ilualien ISST experiment and coordinated participation see"lixternallivents"under Severe Accident implemen-with the Taiwan Power Company (Taip(m er), the NRC, tation in Chapter 5.)
the Central Research Institute of lilectric Power Industry (CRil!PI), t he Tokyo lilect ric Power Company (I'l!PCO),
The NitC's participation in international seismic test pro-the Cornmissariat a 111?nergie Atomique (CIIA),
grams is beneficial both for the sharing of research re-lilectricite de 1 rance (I!dit), and 1:ramatome.
sources and for gaining different perspectives on seismic design issues.The pooling of resources allows the devel-opment of bigger, more complex tests, an important cle-ment in the validation of methods for predicting the scis.
The duration of the Ilualien project is emisioned to be 5 mie response behavior of nuclear plant systems.
years starting January 1,1990. To date a cylindrical con-tainment model has been designed with construction and The Iarge-Scale Seismic Test (ISST) Program at instrumentation o the model scheduled for the spring of r
fluatien, Taiwan, follows the Soil Structure Interaction 1991.
15 NURiiG-1266
. ____a
2 l'ItEVENTING DAMAGE TO REACTOlt CORES
'lhis program encompasses research pertaining to the 2.1.2 l'rogram Strategy operations of the reactor as a system, includmg control-hng power level, maintaming core coolmg and heat re.
A dual analytical and experimental apptoach is used to moval, and maintaming proper coolant temperatures and achieve a firm technical understandmg of the thermal-hy-pressures ender both normal imd abnormal conditions of draulic behavior of the reactor, lhe NRC starts by simu-operation. 'Ihe program also includes consiJeration of lating the actual reactor s continuous flow of neat and operator actions as an mtegral part of the reactor system.
fluids with a computerized model consisting of many A complete understandmg of the reactor operating as a small Wscrete cells exchanging heat, fluid, vapor, kinetic systcm makes it possible to define the conditions of op, energy, and momentum at each small but finite time step.
c' ration that present core damage and also a/tions to Physical laws arc used when possible to calculate all these minimize the consequences of a core damage esent, exchanges. I!mpirically derived formulas, obtained from should one occur. This research program emphasizes se.
npenments, are used as necessary to account for such vere accident prevention and mitigation by enhanring the complex effccts as fnetion between vapor and hquid.The understanding ofleth plant and human behavior related calculatmns are made for each time step and for each cell to accidents and transients. Tins information is used to in a manner similiar to animated computer games, except ensure that regulatory requirements exist that suitably that reactor models have many more objects (cells) and ensure that plant equipment, procedures, and personnel tpse objects interact in a tightly coupled manner at every can deal with operating events and prevent serious acci-time step, requinng many more calculations.
dents or can maigate the consequences of an accident, Our reliance on the computer codes to provide predie-should one occur' tions of reactor response with acceptable uncertainties depends on three levels of experiments and comparisons of experimental results with ceJe predictions. Itirst are 2.I l'latil l'Orl,OrillallCO basic experiments to derive empirical formulas for deter-mining phenomena within each cell. Second are separate-2.1.1 Statement of l'roblem effect experiments to test the code's predictions for a single, complex component such as a steam generator.
I A wide range of reactor plant design sariations exists in Third are integral system tests that are used to evaluate l
the United States,and the safety of these plants for a wide the code predictions of a complete reactor.The results of range of normal and abnormal operations must be en-these tests provide feedback to correct the code and our I.
sured. The NRC is required to independently assess each understanding of the transients.
licensee's assertions and performance of his n'sponsibil-ity to design, construct, and operate a reactor with respect 2.1,3 Resenreh Accomplishments in FY 1990 i
to the safety of the plant for the complete spectrum of 2,13.1 Completion of Thermal.llydraulie Code credible operatirig conditions and events.
1)ncicpment NRC's task is difficult because straightforward testing of I)uring the last 15 years, thermal-hydraulic systems codes a
all transients in all plant design vanations would not be were developed to provide the staff with an mdependent i
technically and economically feasible. On the other hand, capability to realistic:dly analyte plant transients and foss-straightforward and exact theoretical analyses of a reac-of-coolant accidents. Staff analyses are used toinvestigate j
tor's thermal hydraube behavior would take too long to issues such as im R stability, amicipated transients with-compute because energy, mass, and momentum ex-om scam, pn nunied thermal shock, feed and bleed changes take place over complicated interfaces between cooling, and early phases of severe accident scenarios.
j reactor components, water, and steam, and because of the
'I he codes must model a wide range of states of the cool-moving mechanicalintertacesin pumpsand the extensive ant and its structures. pressures range from 1 to 200 baffle surfaces of steme generators in the primary loops.
atmospheres, temperatures from ambient to 2200 17, while flow rates range from stagnant to sonic, Thus, a 2
j As a result, the NIU must combine, in a complex under-wide range of experimental data is incorporated in the taking, hmited experimental data, much of which is less codes to model the physical processes of heat, energy, and than full scale, and limited calculational capability into a momentum transfers in the plant system. The resulting i
firm technical basis for evaluating design basis accidents, code complexity must be carefully tested to determine the safety implications of actual events in operating reac-that models have been implemented correctly (develop-l tors, and hypothetical transient scenarios determined to mental assessment) and to evaluate code performance be n ajor contnbutors to risk as a result of probabilistic (independent assessment by experts external to the code risk assessment studies and these operating events.
development program).
i i
17 NURIE1266
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Additionally, an International Code Assessment Program alternative electrical power sources to meet critical (ICAP) was begun as a 6-year effort to be carried out from safety needs during an accident 1986-1991.
The NitC provided the codes Defeating appropriate interlocks and overriding Illit.AP5/ MOD 2 and Til AC-l'Fl/ MODI to safety authorities of foreign countries in exchange for m, depend-component protective trips in emergeng situations, Fxamples include reopening main steam isolation ent assessment. It was envismned that usage of these codes would reveal modeling deficiencies that would re-valves and turbine bypass valves to regain the con-denser as a heat sink, extending reactor core isola-quire development of new versions.This mdced proved to be the case. A consensus was reached among the ICAP tion cooling availabihty by either raising the turbine code users on the nature and cause of the deficiencies, cxhaust pressure trip setpoint or everriding the trip and in December 1987 agreement was obtamed on a plan function, and enablmg emergency 'oypass of protec-for development of the final versions of the codes, namely live trips for diesel generators and injection pumps.
Illil AP5/ MOD 3 and Til AC-PFl/ MOD 2.That code de-2.1.3.3 lleactivity Accidents velopment was carried out during 1988-1990 with help from several ICAP participants who will also carry out an
.the accident at Chernobyl resulted m. a recommendation independent assessment.
that a review be made of potential events in U.S. light-water reactors where large reactivity increases might oc-cur in a period of time that was short enough to cause fly the end of 1991, the mission of ICAP will be com-rapid fuel damage. This study has been completed for pleted, and a successor program will take effect to main-rnany accident sequences in 17 broad categories for a tain a code user's group focused on: (1) code applications 4-loop pressurized water reactor (PWit) and a boiling for plant safety analysis, and (2) code maintenance.
water reactor (llWit/4 design)Jihe primary results are as follows:
2.1.3.2 Accident Management Strategies Transmitted M M two of the sequences studied that have the potential to cause rapid fuel damage were estimated Over the past several years, the NltC has performed and to have frequencies too low to warrant further con-reviewed numerous pr'ohabilistie nsk assessments (pil As) rideration at this time.
and severe accident studies. From this experience, it has Gf the two sequences that have esttmated frequen-become evident that it is possible to implement certain cies in the range of interest (> 1.011-6 per reactor 1
actions, or accident management strategies, that have year) for severe accidents, the more important is a significant potential for recovering from a wide variety of refueling accident in a !!WIL This is caused by the accident scenarios. These accident management strate-loading of fuel surrounding two or more positions gics typically involve the use of equipment that already where control blades have been removed and are in-exists at plants /Ihe NitC staff has compiled a list of such operative. Understanding of this event and its re-accident management strategies and sent that list to licen-lated sequences should be irnproved so that correc-sees, along with a technical assessment of the strategiesas tive actions may be taken. Changes to technical Supplement 2 to Generic i.etter 8S-20," Individual Plant specifications can reduce the frequency of this lixamination for Severe Accident Vulnerabilities." The event.
licensees can evaluate these or similar strategies for ap' plicability and effectiveness at each of their plants as part The other sequence with an estimated frequency j
of the individual plant examination (11 !!) called for in that is significant is the result of the flushing of lx>
Generic l etter 88-20.,lhe accident management strate-ron during a llWit anticipated transient without gies can be grouped into three categones:
scram (A'ITVS) with inadvertent depressurization and injection of cold unborated water from low-pressure cooling systems. Calculations of the reac.
l Conserving and/or replenishing limited resources livity and temperatures were carried out using the e
during the course of an accident. These resources ItEl.APS/ MOD 2 computer code underdifferent as-4 would include, for example, battery capacity, sumptions regarding timing and availability of low-borated water, and compressed air, pressure pumps and with different reactivity coeffi-Using plant systems and comp (ments for innovative cients.The results showed that the fuel enthalpy rise e
applications during an accident.This usage would in-was insufficient to cause catastrophic fucl damage clude enabling crosnies of service water and compo-hhough less severe fuel damage might still be possi-nent cooling water or the use of fire systems, control e because of the overheating of the fuel cladding.
rod drive pumps in the case of a boiling water reactor Several sequences under shutdown conditions were (llWit) for decay heat removal, and use of conden-found not to lead to rapid fuel damage but to lead to sate or startup pumps for feedwater injection. In ad-core melt; however, they have low estirnated fre-dition, this category includes procedures to connect quencies of occurrence.
NUltliG-12M 18 i
2.1.3.4 Emergency Core Coolant Tests and Analyses ditional steam was generated by interaction of the falling, Under the 2D/3D International less of. Coolant Acci' spreadmg water with the hot core, and some downflowing water was entrained in the upflowing two-phase fluid.
dent (LOCA)llescarch Program, a series of emergency The downflow area decreased with the size of facility scale core coolmg (l!CC) tests were conducted in the German and increased with the 17 oude number of the emergency t
full scale, separate-effect test facihty called the Upper Plenum Test Facility (UI'!T) and in the Japanese 1/21 core cooling (ECC) jet, For PWR conditions, the volume-scale, integrated test facility called the Cylindn-downflow area is estimated to be about 30-40 percent. In cal Core Test Facility (CCIF). Data from these tests were the two-phase upflow region, the core cooling was compa-used to assess the TRAC-PFl/h10D1 and h10D2 com-table to cold leg injection tests. In the downflow region, poter codes so that the codes can be confidently used to the core cooling was somewhat better (typically about analyze commercial reactors. The UI'IF is a model of a 30'K lower cladding temperatures) than the two-phase four-hxip,1300 h1We PWR with simulated core, steam upflow region, especially in the upper half of the core, generators, and pumps. The CCrl'is a rnodel of a four-Throughout the tests the net flow out of the bottom of the loop 1100 h1We PWR with a full height pressure vessel, core was downwards. This was a positive result since the steam generators, and simulated pumps. Iloth facilities effect was to refill and refhoj the reactor vessel rather are designed to simulate the end of blowdown, refill, and than to entrain water to the steam generator. There was refh>od phases of a large break I.OCA.
more than enough water in the downflow region tosupply the upflow region.This was true even for conditions simu-1.
liffectiveness of Upper Plenum injection lating the loss of one liCC train.The excess downflowing k
Six plants of Westinghouse design are equipped with liquid simply flowed out the break via the lower plenum d
I!CC systems that inject water into the upper plenum as and downcomer, y
opposed to the cold leg. For such configurations, the possibility was suggested that upflowing steam might pre.
The TRAC computer code (TRAC-PFl/h10D1) pre-sent the injected water from fiowing downwards into the dicted major trends in the data reasonably well although core, thus preventing effective core cooling. This phe.
certain phenomena such as the downDow rate (over-nomenon is known as a countercurrent flow limitation predicted) and the core void distribution were in less (CCFl.)or flooding. In this event, water would pool above agreement (unpredicted at the top of the core). The loca-the core in the upper plenum. Furthermore, the upflow.
tion of major downDow varied from test to test and this ing steam might entrain the pooled water from the upper variation was not always predicted by TR AC. TR AC pre-plenum and carry it through the hot legs to the steam dicted peak clad temperatures (PCfs) within 80*K for a generators where it would evaporate to create backpres, single pump failure case and within 20 K for a best-esti-sure, known as steam binding, mate case. In all cases, TR AC correctly predicted that the net flow at the core inlet was downward.
The 2D/3D data and associated analyses showed that upper plenum injection (UPI)is effective in cooling the 2.
Concern About !!CC Hypass core over a range of decay heat levels that bound PWR When fiCC is injected into the cold Icys in response to a conditions (less than 1.21 KW/ht). Steam binding was 1.OCA, a significant portion of this ECC may bypass the concluded not to be significant, as most of the steam was core by Dowing around the downcomer to the break.This condensed by the UPI water in the upper plenum.
bypass is expected to occur until the vessel is depres-surized to the containment pressure. If this refilling proc-CCFl did not occurat the interface between the core and ess starts late in the I.OCA transient and takes a long time the upper plenum. Rather, the CCIF and UI'IF data to complete, the core can be reheated significantly after showed that water Dowed from the upper plenum down-the blowdown cooling ends and before the liquid fills the wards into the core in certain preferential hvations.The lower plenum and starts refkioding the core.
core flow divided itself into two regions and a type of circulation cell was established, with one region being A series of experiments in the UI'IF showed that the predominantly liquid downflow and the other region pri-ECC started penetrating into the lower plenum well be-marily two-phase steam / water upflow. The downflow re-fore the blowdown ended. Under simulated PWR condi.
gion did not always form at the same initial h) cation within tions, the ECC started penetrating rapidly at a rate the core / upper plenum interface; rather, there was a de-equivalent to about 90 percent of the injection rate start-gree of randomness in its k> cation, indicating that the ing at 8 to 11 bar while the blowdown was still taking place facility did not have a designed.in bias. Once the and refilled the lower plenum in about 16 seconds; or downflow location was established, however, it stayed about 5 seconds after the blowdown ended. 'lhis refilling put.
rate included liquid accumulation due to condensation of steam, which was estimated to be about 80 percent As the liquid descended further through the core, it efficient.The favorable delivery rate of ECC is achieved tended to spread as it encountered upflowing steam. Ad-by the two-dimensional behavior of the downcomer; most 19 NUREG-1266 l
)
- of the !!CC injected into the two cold legs farthest away depressurization of the RCS were considered. One strat-from the broken hop entered the lower plenum while egy, called early depreswurization, assumed that the reac-l most of the liCC injected near the broken loop bypassed tor head vent and pressuri/cr power-operated relief l
the core.
valves (PORVs) were latched open at steam generator dryout. 't he second strategy, called late depressurization,
'Itc TRAC-PFl/ MODI code predicted much more by.
assumed that the head vent and PORVs were latched pass of liCC than the UI'lFdata. For instance, the UI'lF open a' a core exit temperature of -922*K (1200*F).
Test 6 Run 133, a countercurrent flow test, showed that Depressurization of the RCS to a low value that may 47 percent of the total ECCinjecteJ was delivered to the mitigate DCll was predicted prior to reactor pressure lower pienam whereas TRAC.PFl/ MODI predicted vessel breach for both early and late depressurization.
only 24 percent delivery. In comparison, the improved
'lhe strategy of late depressurization is preferred over modcling in TRAC-PFl/ MOD 2 predicted about 47 per.
carly depressurization because there are greater opportu-cent delivery. Overall, TR AC-PF1/ MOD 2 predicted the nities to recover plant functions prior to core damage and Ul'IF countercurrent flow and lower plenum refilling because failure uncertainties are lessened.
data within an acceptable error band of 20 percent.
The above behavior is much dtfferect from the smaller scale CCTF results, which showed no discernible differ-1he potential for recriticality in llWRs during certain ences between loops and no initiation of refilling process low-probability severe acciJents was studied.
until the vessel was completely cepressurized. Ilowever, the 1/50 scale LOFF test results tended to support the llased on a conservative bounding analysis, this study UI'lF test tesults.
concluded that there is a potential for recriticality in HWRs if core reflood occurs after control blade melting The ECC penetration data from the UI'lF showed much has begun but prior to significant fuel rod melting. Ilow-higher delivery rates of ECC than the values based on ever, a recriticality event will most likely not generate a previous small. scale test results.
pressure pulse significant enough to fail the vessel. In-stead, a quasisteady power level would result and the The UI'IF and CC'IF data showed that the liquid levelin containment pressure and temperature would increase the downcomeris somew hat lower than the bottom of the until the containment failure pressure is reached unless cold leg during most of the reflood period because of actions are taken to terminate the event.
entrainment of liquid by steam entering the downcomer from the intact hiops, then going out the break. For a.
Two strategies w cre identified that would aid in regaining typical PWR condition the liquid levelin the downcomer control of the reactorand terminate the recriticality event during most of the reflood period is analytically estimated before containment failure pressures are reached. The to be about 0.7m below the bottom of the cold legs.This first strategy involves initiating boration injection at or reduction in hquid level corresponds to 20 percent of the before the time of core refkod if the potential for control downcomer height measured from the bottom of the core blade melting exists. The second strategy involves initiat-l to the bottom of the cold legs. This reduction is estimated ing residual heat removal suppression pool cooling _to l
to cause a peak clad temperature increase of about 13 *C, remove the heat load generated by tne recriticality event and thus extend the time available for boration.
2.1.3.5 Depressuritation As Arcident Management Stratep,y To Minimize Consequences of Direct 2.1.3.7 Workshop Eliciting Uncertainties in Accident Contamment IIcating Management Strategies l
i Probabilistic risk assessments (PR As) have identified se.
Severe accident management can b e defined as the use of l
vere accidents for nuclear power plants that have the existing and/or alternative n esources, syst ems, and actions i
potential to cause failure of the containment through to prevent or _ mitigate a core melt accident. For each l
direct containment heating (DCll). Prevention of DCII
- accident sequence and each combination of strat?gies, l
or mitigation of its effects may be possible using accident there may oc several options available to the operwor.
l management strategies that intentionally depressurize and each involves phenomenological and operational l
the reactor coolant system (RCS). The effectiveness of considerations regarding uncertainty. During the period intentional RCS depressurization during a station black.
May 15-17,1990, a workshop wr.s held at the University =
out sequence was evaluated considering the of California at Irs Angeles to address these uncertain-phenomenological behavior, hardware performance, and ties for PWRs.
operational performance. Phenomenological behavior was calculated using the SCDAP/ REl.APS severe acci.
The obj, ctives of the workshop were to answer the fol.
dent analysis code. Two strategies to mitigate DCil by lowing set of questions:
NUREG-1266 20 2
1.
What are the necessary conditions for u coolable the plant and systern behavior, regardhss of what com-state? What is the uncestninty regarding such a puter codes and analyses tell them state? Ilow will the state be defined?
2.
What are the posible management strategies that 2.2 litiman Performance willyield a coolable state or a new transition state?
d Wm 3.
What are the possible negative consequences of these strategies od what are the uncertainties?
A large fraction of all safety related events reported at nuclear pow er plants continue to involve human perform-As a result of the workshop, these questions can '
ance. M ethods and data are needed to identifysystemati-swered as follows:
cally set priorities for, and suggest solutions to human performance issues in the operation and maintenance of During the in-vessel stage of severe accident p rogression, nuclear power plants during normal, transient, and emer-it was generally agreed that water should be added to the gency situations.
vessel as soon as it is made available to the operators 'lhe major uncertamty during this stage of severe accident Program Strategy management concerns depressuri/ation. Depressuriza-
'the human factors and reliabihty assessment research tion reduces the threat of direct containment heating and program has three objectives: (1) to broaden NRC's increases the threat of steam explosion. Depressurization also increases the number of systems available to the understanding of human performance and to identify operators for the injection of water.
causes of human error;(2) to accurately measure human performance for enhancing safer operations and preclud-During t he ex-vessel stage of severe accident progression, ing critical errors; and (3) to develop the technical basis the major uncertainty is determining how much molten for requirements, recommendations, and guidance re.
core and debris is in the vessel and how much is in the lated to human performance, containment (cavity) as a function of time. Knowledge
'the human factors regulatory research program is di-permits the operators to determine whether water should be injected into the vessel or recirculated in the contain-vided into five interrelated program elements: (1) per-ment via the spray system. It was generally agreed that if sonnel performance measurement, (2) personnel subsys-water were limited, it should be injected into the vessel tem, (3) humanaystem interfaces. (4) organizational first since it would end up in the cavity eventually.
factors, and (5)teliability assessment.The purpose of the personnel performance measurement element is to de-
'the major uncertainty, ex vessel, involves the potential velop enhanced methods for collecting and managing for a hydrogen burn / detonation if the containment spray personnel performance on the safety cf nuclear opera-tions and maintenance. personnel subsystem research acts to de7mert a hydrogen-rich containment. This aspect will broaden the understanding of such factors as staffing, requires further study.
qualifications, and training thi influence human per-It was recommended that fkioding the cavity prior to formance in the nuclear system and will developinforma-tion necessary to reduce the negative impact of these s essel melt through neight be a viable strategy. 'this severe influences on nuclear safety. Research in the human-accident management scheme requires further studyas to its effimy and any potential adverse effects, should they system interfaces element will provide the measures for evaluating the interface between the system and the hu-occur.
man user from the perspective of safe operations and 17 om a fission product retention viewpoint, the addition maintenance. Organizational factors research will result t
of water was generally seen as being beneficial at all-in the developruent of tools for evaluating organizational stages of a severe accident. The major negative aspect of ssues within the nuclear industry. And, lastly, the reli-adding water is increasing the potential for structural ability assessment element includes multidisciplinary re-failure when adding cold water to very hot surfaces, i.e.,
search that will integrate human and hardware considera-the vessel, the steam generator tubes, etc.
tions for evaluatmg reliability and risk in NRC licensing, nspection, and regulatory decisions.
lastly, the decisions needed (1) to shut down the reactor and cool the core and (2) to maintain suberiticality and a 2.2.3 Research Accomplishments in FY 1990 heat sink may not bc affected by uncertainty. The poten-2.2.3.1 Iluman Factors Research tial negative effects of accident management schemes, especially during the late stages of an accident, as well as Two ongoing projects in the personnel performance the large uncertainties, may force the operators and measurement element are (1) an investigation process Technical Support Center staff to use their knowledge of that could provide a standardized method to identify the 21 NUREG-1266 i
I i
causes of human errors, and (2) a project designed to relative to their adequacy as applied to computer. based determine if there is a need for a human performance safety sytems. Research continues on developing guide-j information system. Three new projects related to human lines for verification and validation of expert systems.
factors evaluation of processes conducted by materials which is being jointly funded by liPRI and NitC.
l bcensees (induurial radiography, brachytherapy using re-mote afterkiaders, and teletherapy) were initiated to Organizational factors research focuses on group per-identify factors contributing to human error in these proc-formance and factors influencing that pers'ormance. Dur-esses.
ing the past year, research was continued on Nuclear Organization and Management Analysts Concept (NOM AC) for modeling, gathering information, and in-Research in the personnel subsystem element eontinued I
with an evaluation of the impact of overtime and shift dexing plant organizational performance reliability and scheduling on operator performance bord on nuclear on the development of leadmg mdicators of the safety i
power plant data and to pmvide a quantitative data base perfomance of plant pmgmmm AccompMmentsennq to evaluate the safety implications of 12-hour shift sched, the past year meluded: (1) field testmg parts of NOM AL I
at a foss 1 fuel plant and at a nuclear plant;(2)domg m,tial I
ules. Work continued on the first step in the development i
i of a method to evaluate the effectiveness of training pro-correlation studies for leadmg mdicators of the safety l
grams at nuclear power plants,i.e., a workshop of training PC'.IU " "CC "I IP ""I E*8***
E" ma atenance and trammg); and (3) completmg a study of the feasibility of i
esaluation experts was held during the first quarter of FY IW1. A research effort -addressing what factors are con-transferring leadmg mdicators of safety performance sidered when making staffing decisions and how these fmm the chemical processing industry to a nuclear set-8 l
factors relate to safe startup, shutdown, and operations of ting.
nuclear power plants continued. A new study involves l
conducting a review to establish the applicability of avail, 2.2.3.2 Reliability Assessment able infermation to the understanding of the impact of This continuing RilS program provides the tools and data environmental influences on human performance, necessary for assessing human performance in ways i
adaptable to plant probabilistic risk assessment (PRA)
Iluman systems interface research continued with NRC studies and for systematically applying the results 6f those j
participation in the OliCD llalden Reactor Project. The studies to the resolution of genericissues and consequent i
results of a survey of the commercial nuclear power in-regulatory decisionmaking. Accomplishments during the i
dustry's planned use of advanced instrumentation and past year included: (1) cxpandirg the component failure controls technologies were published in NUREOl rate and human error likelihood data in the Nuclear Com-1 CR-5439, "lluman Factors Issues Associated with Ad4 puterized Library for Assessing Reactor Reliability F
vanced Instrum;ntation and Controls Technologies in (NUCLARR), which is a probabilistie data base: (2) con-i Nuclear Plants." The results of an assessment of the costs verting the Maintenance Personnel Performance Simula-and benefits of expanded regulatory guidance on normal tion (MAPPS) computer code to a personal computer to and abnormal operating procedures were published in enhance its user friendliness; (3) initial ficid testing of NURiiO/CR-5448, "Value-impact Assessment for a Task Analysis Linked Evaluation Technique (TALENT)
Candidate Operating Procedure Upgrade Program." An as a method for fully integrating human factors expertise analysis was performed to assess the effects of upgrades in the PRA process: (4) applying artificial intelligence-l to hical control stations in mKlear power plants on both based computer code, Cognitive Environment Simula-human performance and plant risk. The results of this tion, to assess an operating team's diagnosis of an Inter-analysis were published in NUREG/CR-5572, "An facing Systems loss of Coolant (ISLOCA) sequence; (5) -
1 Evaluation of the Effects of Local Control Station Design contimiing development of criteria for the transfer of l
Configurations on Human Performance and Nuclear non-nuclear human error probability data (civil trancpor-f Power Plant Risk." Work on developing a guideline for tation, military) to be used as bounding (anchor) values in use in performing human factors reviews of advanced analyses of nucIcar power plants: (6) evaluating a risk-control and display technology continued. Work also con-based safety-system function trend indicator, including i
tinued on the identification of the frequency, severity, and the use of Nuclear Plant Reliability Data System nature of procedures violations in U.S. nuclear power (NPRDS) data: and (7) continuing development of a risk-i plants. Research continued on computer classification, based configuration management concept for enhancing 4
which involves review and evaluation of existing regula-safety system test, surveillance. and maintenance techni-tory guidance documents and quality assurance methods cat specifications.
i NUREG-1266 22 i
(
-,w c
= -
3 REACTOR CONTAINMENT PERFORMANCE The basic criteria for licensing nuclear power plants for failure of a pipe or steam generator, (5) energetic fuel-construction and operation are judged to have provided a coolant interactions that occur as molten debris falls into considerable safety margin, affordmg the pubhc protec-the water filled lower head or as water is added to molten tion from radiation even unJer severe accident conditions debris, (6) the composition, morphology, and tempera-such as those that occurred in 1979 at Three Mile Island, ture of debris at the time of vessel (or reactor coolant The physical possibility of even more severe accidents system) failure, and (7) the mode of vessel failure. The than that at TMI is, however, recognized. Considerable core melt and reactor coolant system failure program is progress has been maJe in recent years in understanding divided into three main activities: (1) cote melt progres-Ihe underlymg physical and chemical phenomena that can sion and hydrogen generation; (2) f ucl coolant interac-occu r in a sever e accident. Such information is essential as tion; and (3) the chemistry and behavior of fission prod-a basis for assessing potential safety improvements and ucts released during core melt. He in vessel core meli for making decisions on whether or not particular im.
progression and hydrogen generation work includes in-provements are warranted. As pointed out in the Com reactor experiments, out of-reactor experiments, exami-mission's Severe Accident Policy Statement, such deci-ration of specimens from TMI-2, and analytical model sions should be based on a combination of engineering development. The fuellcoolant interaction work is judgment (i.e., a deterministic method of setting and as-focused on the development and validation of a model for sessmg safety margins)and the application of probabihs-fuel / coolant interactions for use in accident analy sis. The tic risk assessment techniques based on up-to-date ex-research on the behavior and chemical form of fission perimental mformation to evaluate the likelihood of the products released from the fuel in the course of a severe occurrence of rare events.
accident is being conducted at high temperatures when core geometry is changing and fission product chemistry in similar fashion, the same underlying science and deci-and its effect on retention of fission products within the sion process can be apphed to reevaluations of existing reactor coolant system is significant.
safety sy stems and regulatory requirements to determine if particular conservative assumptions have been war-3.1.3 Research Accomplishments in IT 1990 ranted in terms of risk reduction.
3.1.3.1 Core Melt Progression and ll drogen 3
Generation 3.1 Cere Melt and Reactor Coolant System l<,m, lure In December of 1988, a draft program plan was issued i.or the nu 2 vesselInvestigation Project (vlP), a multina.
tion program under the aegis of the Organization for 3.1.1 Statement of Problem Economic Cooperation and Development (OECD) being conducted to determine (1) how close to failure was the
(
Major uncertamtics in estimating the probability of early bottom head of the TMI 2 reactor during the accident,(2)
I contamrnent failure, and the associated radioacave re" what damage occurred in the bottom head, and (3) what lease, appear to be significantly related to uncertain ties in the m vessel progression of the acci&nt whde ti,e fuel can be applied from the TMI-2 accident to understanding challenges to vessel integrity from core melt accidents. A material remams in the reactor pressure vessel Until a wcond revision of the plan was supplied to the OECD hetter understanding of core melt,includmg fission prod' TMI-2 VIP management board for their June 1989 meet-uct release, hydrogen generation,5md response of the ing. The plan details the procedures and schedules for reactor coolant system to fuel meltmg and relocation, is obtaining and examining the samples from the lower head reaheed, containment failure probabdities and related of the TMI-2 reactor vessel.nc vessellower nead sam-source terms will continue to be consenatively based to pies were successfully removed during FY 1990, and ex-ensure an adequate margin of safety.
amination of the samples is under way.
3.1.2 Program Strategy Pieparations were completed for the Fl.irf-6 BWR test in the Canadian National Reactor Universal (NRU) at In order to better understand just what happens Ourmg a Chalk River early in 1991. FLIIT-6 is a full-length test in core melt accident, and thereby reduce the uncertaintyin simulated IlWR geometry that includes a stainless-clad both the nature of the accident and the potential release H C control blade, a Zircaloy channel box, two high bur-4 of radioactivity, the NRC is pursuing a program of re-nup fuel rods, and 12 fresh fuel rods. FLi(T-6 will provide search addressing (1) the heatup and meltdown of the data on length effects to supplement current results in core, (2) hycsogen generation, (3) fission product release BWR core geometry with one meter and one-half meter and transport withm the reactor coolant system, (4) the test fuel bundles that were obtained in the DF-4 test in natural cuculation of hot gases that might cause early the Annular Core Research Reactor (ACRR) and in 23 NUREG-1266
1/
h BW R teos m oc (icn un( ( >R A cerem tor fut I damage R1 A code was completed in July 1490. The accompany mg tem b h. I h I i 111' h test will etuume f uc I dam y -
uwr manual u n pubbshed as NL'Ri'G CR--5M5 in i
pro;ic'u m osi Ac mch itlocain n. hy-Jrogen pencra-Octohcr 1490 l'or the containment, the TRlWDS code j
tm mJ h' on prolxt u lcase. Ito will be the fust calculates the partaion of iodme betucen the aqueous m Tdc tcq i ! a lutt-lcngth sectism ol It% R luct anJ udi phase and the gas phase m the containmenh the produe-prwoJe a tw s Wa k f or data generated by in-pdc anJ tion of otrame iodide species, contairanent water pool out of d
t ' tos w ah ?orter test fuel hundles.The fust m chenustry, and the extent of ioJme revaporization and a 'encs td I tc-phne melt propewan nperiments was resuspension from contamment surfaces and sumps.The pcdorn 'J m t!u M 'HR test reactor on the mch dy nam-code was used to calculate the revolatilvation of iodme u of a nu talhc uust suppor ted t cranuc p.uticulate de.
hom the containment water pool and the proJuction of hns bcd mJ tht tcia:me pool nu hout processes anJ organic ioJme in the containment. The calculations were thren M nm n the "bhw ked on e" conturation that completed in December 1940 and the lesults will be used oa un cd m the'l hil 'acciJent and appears to occur in ;di to support the revision of TID 14S44 (dated March 23, PWR xcdcnhimJ a cunently unknown fractum of IlWR 1%2), which outlines a procedural method to calculate coJcnt s Poq tuaJianon cumnunon tPli ) was per-the off sne radation dose from ioJme esposure.
ior med on u.c oper m,cnt c.pu:e. and the resuhs were anaMed. A ccenJ cermmc t rust was found to form A report on the ST-l test at the Annular Core Research arounJ the m mg <cr.umc meh pool that mteracted.
Reactor was received in June 14S9 end has been pub-hshed a, NURIGCIM34i'this test was one of a few
'lcW cal pl mmac was perlonned anJ a propam plan m pile hssion product release tests conducted to serve as osu J f m < w ncs of cu cactor espenments on metallic benc h mar ks for the more num erous and much less expen-
- rmh i ch e om.mJ hM km, e Im mauon under ItWR dn,-
sne Vi senes of out-of-pile fessior nroJoct release tests.
j cmc conJmorn _I he putpose of these espenments is to detairaue the conJmons. if any, m w hich the metalhe 3.1.3.3 I ucl! Coolant Interactions j
mt h Jf uns iro,, the ItWH core and core plate as tornied
,o m od ik i,L s kt,J core anJ lar; e ecr anne melt pool ol-In a severe accident Mtuation, there is little doubt that the t
t I
the l MIO aJent.
prnnary efl. orts of the operators will be directed toward rnakmg water available to the reactor vessel. An impor-Arcpon Nt RI (l ( 'RJM. "In-Veuel Zircalov ( hida-tant qumam that needs to be considered in uew of such l
uoiih J:o,< a hnuanon lichauor Durme ScEtc Acen hkely eff orts is u hat are the likely consequences of these de:
' w n pu %hcJ to sununar ve eunent openmental dorh Along wnh the potential benefit of ach>vmg a
{
resah and mtt rpretauon el hydrogen generation along stable, coolable configuration, restoring water to c core
}
wnh a pm
,uus!c that sumrnarveJ the teport.
that has been sescrely damaged can have eff ects of which t
the operator should be aware. lhe operator also should Instrun cnt tuhe.uptun and gecuon anJ reactor sessel
- 'unnon cognizance of possible symptoms and response hmer heaJ hu!ure weie anak/cd fo: tspical PWR and of the plant to addmg water in such circumstances (e.g, HWR desiens in lY IM lDulure maps relaung system inhn< ore / concrete interaction, increased hydrogen presstu e, ti clor sessel temperature, and fadure mode pen tion, increased containment pressure). Inperi-I wcre produced t he ladure inaps can be used to cstunate mmMw conducted at the Umversity of Cahfornia at
- he rnost hktly f adute mode for a gnea despn under Santa liarbara and the CliC-JRC facihty m Italy to ad-cuJent conj uons. Informanon trom the TP11-2 VIP dress those issues.
sample cumnano:a is bemg used t o ex.unine consistency
}
of the piedk nans wah the known IMI-2 asciJent md 3.2 Reactor Containment Safety cnJ-state-i 3.2.1 Statement of Problem i
3.1.3.2 Ikh.nior and Chemical I orm of l'ission pmintg
( ore melt accidents have the potential to produce high pressures and temperatures that nught cause contain-Computer codes are bemp deseloped to prethet fission ment failure. It is known from previous ri6 studies, and product release and transport m the reactor coolant sys.
from the experiences at Chernobyl and Three Mile tem iR( N anl the contamment. 1:or the RCS. the Island that containment suruval or even delayed failure VICI ORI A toJe a deseloped to esumate the quantities has an all-important effect on minimi/mg the release of of frion products anJ aerosols released from the reactor radioactivity to the environment in the esent of a core core, the nient of their transport through the reactor melt accident If reahstic assessments of the conse-cooiant system. th; msentory of raJionuchdes mailabic quences of core melt accidents, which so strongly depend for release alter core debus is espelicJ from the reactor on w hetheror when containments might failin thecourse s essel, and the nient of insion product tevaporizauon of the accident, are to be maJe, then an understandmg of f rom the tc. ctm coohmt spitm A \\ crston of the VR'l 0-the phenomena that occur in containment in the latter i
4
- - - ~ ~ _. - - _ - - - - - - -
h b
stares of the accidcnl that could lead to contamrnent and refine testing procedures for dispersmg incit with failure is imperative, steam as the dnving ras. These teste, designated as the TDS tet,t series, resulted in the demonstration of a reli-able, reproducible, cost-effecthe testmg rnethodology, 3.2.2 Program Strategy I ollowing development of cificient testmp through the TDS test series. Sandia inniated the limited fhght path NitC's scscarth efforts m this program clernent focus gey,3, lhese tests were iritended to confirin prehminary dtrcctly on the phenomena believed to be most hkely to scaMg analysis, which surgested that debris pas heat produce high pressures and temperatures that rnight fail transfer and hydrogen production were a linear function the contamment: high pressurc ejection from the rcactor of the interaction length.1hc hmited flight path teMs sessel of hnely dmded particles of rnolten core debrts; continued into early FY 1991. I inally, DCil activities in rencration of noncondensible and flammable cases from FY 1990 included documentation of the llrookhaven low l
the decomposition of concrete by hot core debris; the temperature simulant tests investipating rnett dispersal direct thermal and chemical attack by molten core debns characteristics of the Watts har cavity (Technical 1(eport of structures and engineered safety features; and the A-3024, April 1990).
burning or detonation of hydrogen and other gases pro-duced in the course of the veciccat.
3.2.3.2 Il drogen Transport and Combustion 3
NitC's research prograrn dealing with reactor contain.
W i
'd W den i
ment salcty conusts of three atcas of phenomenological tion experirnents were published in NUlt!!G/ Cit-4298 in research plus the dewloment and use of computer
'7 39g.lhe CONTAIN code was used both as an codes that combme the e' cts of all the separate phe.
experimental design tool and to perform blind post test nomena. lhese four resemh activities adJtess: (1) the calculations of hydroren distnbution experiments in the mteraction of molten core debris with struental concrete, Rdend 1(cpubhc of Germany's full scale HDit contam-incluthop the ablation of concrete structures heat trans-ment,,the aussment of the experimental data was re.
fer to structures in the contamment, the generation of ported as an llDit standard problem m Nosembtr 1990.
flammable and 1.oncondenMble gases, and fission prod-Agreement between the cakulathms and the observed Octs contaming acrosols; (2) direct containment heating by molten debns partides ejected frorn the vessel at high hydrogen distribution was generally good.
pressnre and hydrogen production resulting from steam oxidatton of the metalhc component of that debris (3)the 3.2.3,3 Integrated Codes and Applications transport, miung, and combustion of hydrogen m the ec tamment,inJuding the potential for detonation; and the SCDAp/ittil.APS/ MOD 2 code manual, which in-(4) the development, validation, maintena.u.. 4 appli-cludes code structure, models, and correlations, and us-cation of vanous coraputo codes that are capable J de-cr's guide (NUlti;G/ Cit-5273) was published in Febru-senbing the multiple phenomena that occur m rvere ary 1990.
accident sequences of interest.
'!he COMMILIC code manual, which includs code 3.2.3 Research Accomplisliments in iT 1990 uructure, theory, models, and numerics, and user's guide (NUlWO/ Cit-5649) was published in November 1990.
3.2.3.1 Direct Containment lleating 3.3 COntalHH! CHI StruClural IHlegrity 1)unny IT 1990. NI(C's research under this acsity em-phasi/c J development of an integrated scaling methodol-ogy at well as development of a testmg procedure in-3,3.1 Statement of Problem tended to more closely simulate nuc! car pow *:r plant conditions. Dunng this report period, the technical pro-1he major source of risk to the pubbe frorn the operation i
gram proup (1PG) developed a severe accident scaling of nuclear power phmts sems from accidents that lead to i
methodology (SASM) and used this methodology in its a containment failure.1he regulatory concern is that the first application, direct containment heating (DCll).
failure modes and associated hiad levels for containmt ni I acility-specific scaling analysis was performed by Sandia struetures cannot be predicled with any seal confidence National 1,aboratories in etder to develop a preliminary by the methods used for designJihis is especially so if the test plan for integral effects tests, i.e., tests with melt, contemplated failure mode is localized leakage. Both as-mcluding unoxiJi/cd metallic components which is steam sessments of the risk posed by h> ads outside the design dnven into a pre-pressurized volume, with and without a basis and estimates of the effectiveness of proposed pre existing water pool. In conjunction with scaling analy-mitigative Steps require an ability to predict the way in sis, a number of scoping tens were per formed to develop which a containment will fa L 25 NURiiG-1266
~
E i
3.3.2 l'rograin Stralogy 1%9. Testing etJminated in an overpressure test to fad-ure that was complekd on Jul) 31, In.w. 'the wntain-liesearth on containment failure modes is based on the rnent moJel had oser 500 cmbedJed or attached sensors I
observation that neessn e leakape can oscur, basicall)i that morntored its behavior dunny testing.
Inun four sourtes:
A post test workshop was held m the United Kingdorn in 1
1.
Gulure of the shell, either the containment shell Nosernber 1990 to evaluate the resultsof thc esperiment.
i itself m the rase of steel eontainments or the hner
'lhe nnici faded at pressurc oIO.7? MPa(about 2.2 times m the case of concrete containments:
its design prewure). ~lhe failure wasin the basemat of the j
model, a part that was not depicted accuratel). Conse-
{
2.
l.calage at 1.upe ps nctranons as a result of the in-quently, w hde the test can be used to benchmark calcula-l clostic ddormahons and/or degradation of seals and tions of contamrnent response up to the pomt of failure,it j
fasketN did not yield results that could give any insights about failure modes on pf olotypes.
A lxakare at electrical penetrations due to degrada.
hon of moa rials under the hyh temperatures asso-2.
proposed Cooperatis e liflott in Japan dated vath acadent scenanos and
'lhe NRC research program on containment structural 1.
I rai. ape throuph vahes due le prewure and tem.
integrity has included tests to failure of models of steel perature elletts.
and reinforced concrete pWit containments. A research l
proptarn on containment perlot mance under severe acci-Itoscart h related to shcIl fadur e or deformations of pene.
dent condiaons has recently begun, under the sponsor-tratmns rests on ;malpes of and esperiments on model ship of the Ministry of International Trade and industry tosts of actual contamment deugns. These tests mvolve (Mil 1) of Japan, prehrninary discuwions have led to the prewuriwion up to failure lesels under ambient tem.
conclusion that it would be mutually advantarcous to pelatures. Since seal and rasket materiats are adversely engare m a jointly sponsored research progratn.
allected b) the temperatures awociated w ah severe aceb dents, separate tests focusing on Ihe development of!cak.
,fwo areas of potential cooperation has e been identified, are ate performed under pressure and temperature wn-One deak uith sicel containments, w hich are used m both dstions. utually at f ull scale. I Aamining the posubihty of the UX and Japan for llWit dedyns. A test to fadure of a developmp leakare throuph electricid penetration assem-model of a Japanese flWil containtnent vessel has been bhes and sahes aho requucs espenments under tem.
Propowd, Also proposed is a test to Imhne of a perature and pressure canditions at full scale.
prestressed concrete contaimnent model typical of de-signs used for pWits in both the UX and Japan. These l
4 3.3.3 l(esearch Accomplisliments iti IT 1990 (SP F P ""' # " # **"""
- al methods previously developed.
3J.Al Stimtural f ests 33J., I.,ympment flatch.t ests 1.
Cooperathe lif! orts m the Umted Kingdom
'!hese tests were part of the cifort to develop a more complete understandmr of research results after a The NRC participated in a test of a rnodel of the Si/cwcll U6 scale model of a reinforced concrete contamment was "II" contamment performed by the Central lilectricity tested to failurc in July 1%7, A 40-mch (1.0-m) diameter Generating iloard in the United Kingdom. Si/ewell "l!"is a pressurized water reactor (pWit) that is housed in a equipment hatch, typical of equipment hatches found in UA contamment bmldings was incorporated into the ptestrewed mnerete containment. 'ihe design pressure model.
used in the containtnent and the rnodel is 0J45 M pa.The contaminent structure is based on a llechtel design, mak-A series of tests are unJer way to provide engineering mp it s cry simdar to some of the prestressed contamments data that can be used to validate analytical approaches to in the UMed States.The contamment model was tested predict leakage f rom pressure-unscating equipment to sitocural failure to demonstrate its pressure resene hatches and drywell heads. Specifically these tests ate to and proside data to benchmark computer analpes. One provide mformation on the effects of bolt preload, bolt major difference between the containment nnkl and the stiffness gasket material, gasket aging, and temperature full si/cd prototype is the lack of a steel liner in the on the containment pressure at which significant icakage model, which used a rubber bladder as the pressure initiates and on the leak rates that anse.
boundary during the hydrostatic tests.
lypically, the total preload is specified such that separa-the U10 scale model was prewuriicd sescrat times to tion of the scahng surfaces wdl occur in the rangc of 1.1 to 1.15 times its design prewure durmy the week of July 24, 1.5 times the containment design pressure, llowever, the NURI G-1266 26
.~.
rae t un stdl nutntun a seal f or some posnive separa-the enutoninent donty'socie tcactor nudenN (h de-tion. the nurrutuJe of whh h depends on the mmpresuon u lepN, u nty mp, J; m wu almr. at.J nantanung
% t rt towon mJ wntamment picuure anu temperaturt methoJs for anal nny the wn%quenen of in plant and 3
ellule sncre ;nudent ph 9 sal proccurs for ine m tal 3
An analy t h al method lus N en deu loped to ntunate the aucurncnt.md den lopmp and denmnttratmp mcthmh prtourc arul terupciatute at whhh urruhcant leakape for quantifymp the unct rtamt) la ink otonato anJ the hnt murs and the rate of Icalare for prcuure anJ tem.
tclatne conthbutions of spcuhe issue uncert unty to the tienaut e wndens abtwe thn loc! Aho, the amount of userall uthettamty; th tc.nscump periodical > the fic-leak y e espetted n ulcubitcJ uung fluid rnet hatues queneirs, consequences ynJ rnk of tncte acuJents m eqtunie lor t hokcJ ! tow throm h a du.t of known arca, nuclear powcr ;4 ants and perfornnnp tw r reuew of where the arca n equal to the pp between the rasket anJ tacthoJs used and inolts oblamed; and W dnchipmp scahnr 'urlac tunts the pcrimeter ief the pasket.
lok AiscLiinanarernent tuoh capable of dctermitunp the incremerital risk reduction aM utCd woh ploposed At the end of IT 19no, four anhent temperature tests plant deurn anJ oix ranonal rooddnations and auntiny lute bctn tompt.ted wah an aJJiuonal(nen tests to be m the puotiaration of ciforts in Ikenuny and tescatt h conJuard ct eloated temperaturn. Tbc ruaumum test actinuet h mpetature udl be 700") 'lhe diccis (! ternperature are prmunty twofo!J l ast, cloated temperatures tt nJ 3,4,3 [{esearch Accolnldislunents in lY 199()
to dc!IaJe the ra'Let nulerul,whhh may lead toleakape c:a her tiun at andm nt temperature. Also,if the lu dts are 3 A 3,1 l(okw of PRk urmhcantly wolcr than the lutsh s!ce e, the tileitwe bolt prchud n m.tc o.ed, thus adJ.tional tont.unment Prolubihsuc rnk atu!)sn (PR A)is uw J by the NRP staff
!}
preuulc would be itqmred to produse Icalare.
to support the inoluuon of a w eJe 'pect um of reputatory mues. I or heensed plants. PR As ar e sornetone volun.
the rouhs to date mJicate that, ahhourh the actual tanly subnutted by hccinces to suppori then specihc pro-l hatc h ht h.o ier n trot urnf or m, the.o cr ape,t o ponse of pm d means for t esoh mg sush w ues. l'or adsanted I
the hats b l.nound the cactuulerence t miiudmp the picy plants of the luhae. npphcants are ttynted to perform sure at a hn h pichud n os etcome, compares lawrably anJ subnut PR As as part of their ou:rall bcense apphea-
)
w n h cakuhuions I he ons t of Icalare was piedwtcJ w ah nont Reurws perlor rm d m F i loud imluded the hilloe
(
- c. unable accur aa. IIou ner, the pr ethcted leakare was mp:
mus h ytcater than measured, lhe soutte of thn daticp-(
ana n 1,eme imnurated.
& ce E le ldan1 U mt I ipa ) i hn PR A was subimtted
}
~
voluntanly by the beensec f or NRC ttall rcuew. ILc l
t euew has uncos rt ;d sn er al weakotun m inc in enst e's
)
M lleaClor Acti(lolit l{isk Allitlysis anal)us whh h has c ha n remhed. I he iturw was com-plcted m 1 Y 14n0.
j 3.4.1 Stiiternent of l'roldent jg,un n 4 W (') thn PR A m aho a udunte subnut-Probabdnue rok analyus d,R M has been shown to be a tal by the inttate, who pl ms to the the document as a j
s)Metnau ' and nim p! Chensn e InC! hod for idc ritil) lop and iclerente m lutme Inhmcal dmuworn on reyulatory g
g evaluatmp the dlatncnc6 of salcty unpron ments pro-Ivsed to aeJuce the hkehbood arvJ wnsequence of no-3M /
W) lh PR A A
mlun u i.uba b dear pow er plant acuJentt PR A n used by, NR( stall 1
in a number of ways mdudmp for evalum np the tcu t of tal by the hcensee, who plans to use the docurnent as a h
m futm h'hui dNw mn on m'd%
safet) at nelectcJ operatmp planN for awesung the mar.
W Th ree wiu be wglend m lY luut.
pms of micty m eut rent regmrcmen% in hrht of the ( om-inmiorMd ety roal pohey; f or momtor my plant per form-fgqqg g,pn Cul ) In order to comply wah a hccine ance; and 101 iJe uL my potentul unprovemems in r
condaion, tht hcensee for thablo Canyon has developed eqmpment or operator iclubdity.
, ;g ggg
.n prop a k ut of tb propm, We f
lis ensee n perfornung a PR A. llecatae the senmie por-3,4,2 l' rope ain Strategy thin of the work unohes the t'erclopment of some new PR A methods the staff review n proseedmp as the vari-
'lhe icactor acudent tok analysn rescatch cllort n ap-ous stages of the PR A are hemp performed. ~the renew phed m four war These mduJe (1) proudmp opert was neatly completed at the dose el the report penoJ.
reuew of sesete acudent PRAs to assew for nample, the rok unpli,atuu of atodent nunagement strategies Ul1L!wm nl/dlR A PR A has been submated as part of m order to muunu/c the tt lease el raJnuctne matef ul1o the hcenung apphcauan for thn aJvant cd llWR. Rent w 27 N U R I-( i - 1266
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ui tidl im At u A at it-c t h" f 1. r ( p4 01 i, < si m, totd u t H, a
b-hsst:saf) tie A s t lop imJ use 1
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tt ite-of-In hoolop int thoJs for pctfotmmy anst tesica-3 4.32 (%niph th:n and Ih,ica of Ib.u tot IthL ul PR A and to duchip, tauntaim and preside quahty Ittit n ote thu nmt ut esmance fot s us h nu thoJs Ih l t!4 uals l% b dd N R(' o Rol Rt int ns e 16 [J the dlalt s t f wu o! thV ion 1 of A e a compe r cok that Rt n tof
- u. nt" ( N l:RI (i il5DL o wol as a win s of sgportm cimit,ntor reports foi cstunates the let-atuJent telcase of raJuuctive inate-rul to die ensuunna nt anJ heaWi and etontunic (one puNu wtoment. l he dndt r t pert at ', eJ the t @ hem 3
ournces to the pubhc~-um comt IcicJ and in nie avail-lti d t e (. if e i, s'
.M,'tt Ut ja c l
..ra.
p atl to dw puk Hendnu bnp d. the (o h w10:
a pl eas Nuity m L /t.m ilH.L Nc-jugah (lenn L Pcmh lionom (Pa. L and (it.mJ ( suH iMm) 'I he rcpoit un unu nahon il(tand od problems v. in propreu at the close t U - t d tht linj,h ittom ad the In t analy n s im rero!ati*ry of the icport penoJ.
n acs tu.h as unplemcuit n m of 1he ('omunsuon S ilc t) t it'd.ifid hes tf c Acclk f.1 l\\i!!S '.lat( flW!jls. '] a gi bl(('
! be b) Ate m AIhd) D idid 15nk b u es\\nllnI h b N A) sys*
f unJt J r n ecu s of tht Jndt rt pot t w( re At.um J anJ tem atal the IntegratcJ Rehabihty and libk Aucuruent pnWhed n NI R1 (i ( R 50nh;mJ Nt'RI (H R- $ 113.
53 stem (IRit AM were conctned to addren the need in a lj,th en t!c A!'ictican Nm Icar Nyh h $ piqwitcj anJ dC'Cil!'UdAlbthC AUd U!M'IO pILMdClbe b b('Mllb ICb' puNnhed a ruinc of the dialt repon abihty d,ita that are eutrintly available only on large rn.unt r a me wmpulcis. 'the development of luph.
I ht N R(.' Fla!I anJ sup;s rtmp (< intr.as o s up.litcJ the perforrn iiice innrocomputers has prosiJed picater ca-hsc t o k analun 'l he upJ itn, whn h w eir ymte cuen=
pacdies to mtcratt with estenuve data bases for a large j
- n c. w t :e intt rnicJ to rt Det t comnwnts in en cd, to ic.
number of tistis.1)urmy i Y 194U, f ordback from a hrn.
m,t the pin cni plant des 70 anJ opetatmp ( h oactern iled trial of the tuo wJes was mcorpotated and "produc-th. to nupr m e the tut thods utcJ, anJ Io nn or porate new tion" Vot*nms made hnal ('ourses have been condutted i
nn innental J i a on u scic au iJcnts tc ohmp f rom the to train (tall personnelin the tne of the codes.
n ' cart b profunn of NRU and otht n liv nicans of these wJes and othet inethods, risk analysis g
I he,omplctcJ ra w s cf von of NlHI (i l1$U we dtin' dipporI was pronJed bi the elf to http m the resolution o ed to Ihe ('omnuswn m Apr d PM and pubinhed as a s
danuMUdM m n&dy M nu@
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'nonJ dr dt foi pect icuew m Junt ph9. A peer stucw c
pu o, mmmd edm m umd u_,m_om Act, hniomplett J ns fornulicuca el the d$woment anJ A set of scnuovay stuJin on a l'WR tn response to e
l
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p%ia d rencraF pontn c hnJmps. 'l he hoal s etwn el f
3 i
the r epor t (N t :Ri ti IINh was pubh hrd m lb c.nber Amdm W die elfcct of residual heat removal l
lWn i Vols I and A anJ l chru.ny 1WI (Vol 3I pumh opmbddy on cote damare frequency at Sequoyah.
3.U.3 IthL MoAI lit utopment amt \\pplication A repot t on msights into plant safety emerymg from e
PmbAhsuc nL analp he bcsome an unportant toolin probabihstic analyses, presented to the NR("s Sen-i t ! u N P. ( 's aw unenn of N d !V hsues m thc dcQn and ior Management Mceting.
. t I
i 4
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Nt'iun lx 3
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- _ ____U m~.%
4 ASSESSING SAFETY OF NUCLEAR WASTE DISI'OSAl,
'the NltCi waste management research seeks to (1)de-4.1.2 l'rograin Stralegy velop and s erify inethods for predictmp and assessing the pctformance of waste disposal facihties;(2) evaluate the
,lhe research propram has been runded by the need to data bases used in such performance assessments; (3) provide the technical foundation for NitC development prosiJe technical support to the licensmg staff in ' heir of a set of regulations and a licensing process for the interactions with the Department of 1.nergy (DOli) and review and licensing of the Ill.W repoutory. 'lhis frame-the States; and (4) develop regulatory standards to sup, work for N1(C review will allow the formal heensing ac-port the liendrig of facihties and methodL for the dis.
rivities and the turporting research ta bc fwuxd o.,114 posal and manaptment of high level and h>w-level radio, significant tecanical issues.
acuve wastes.
At present, the NI(C has active research programs in hydrohiry. peolory, materials science, peochermstry, and Durmg IT IWO, research program plans for both high-
" U'"' " " d'N) lines related to the management of leul ra.hoactive waste and low level raJioactive waste high lesel waste..the research combines theoretical were developed to help ensure the responsiveness of Wyw la mtory and field experiments to identify NltC's waste management research program to the needs of NitC and State radioactise waste licensing staffs, and quanufy the physical processes and phenomena irn-por tant to wasM isolation so that the N1(C can determine repository performance and quant fy the uncertainties 4.1 Illgll+1,0 Vel Waste associated with characterization and measurernent of these processes. All this work is integrated into an inde.
4.1.1 Slalement of l'roblein p ndent ilLW performance assessment methodology, liffort is also requir ed to validate many of the snodels that De high level waste (llt.W) thsposal pohey for the und@c We methodology. %e ulbruate poal of the Umted States is defmed by the Atomic linerry Act, the way manapetnent r esead prormin is hi prowde linerry lleorganization A'ct, the Nuclear Waste Policy the techmcal basis to support the licensing staffs inde.
Act, and the Nuclear Waste Policy Amendments Ae't pendent review of the appropriateness and adequacy of (NWPA AL The last. signed into law in 1%7. provides for DOI s demonstration of impliance with 10 CI It l' art 60 i
and the lipA's lil.W standard, in addition, NitC's waste the development of a reologic repository fcr the perma.
nent deposal of high level radioactive waste in the State
- """ECment research seeks to provide technical support of Nevada at Yucca Mountain and assigns responsiinlity h) the licensmg staf f in their interactions with DOli, the for repoutorY desclopment to the doi!. According to the te of Nevada, and other participants and interested l'ederal Goiernment's lleorganization Plan No. 3 of panics and to develop regulatory standards to support the 1970, III.W enuronmental standards development is the
'nung of the dnposal and inanagement of h$h level mdioactive wastes.
responsibihty of the linvironmental Protection Agency (l.I'At and the linergy 1(corpamration Act assigns the regulation of 1II.W diposal to protect pubhc health i.nd 4.1.3 Research Mcomplishments in IT 1990 safety and the ennronment to the NitC.
4.1.3.1 Ilydrogeology An i11 W repoutory poses problems invohing regulatory Since transport by ground water is the most hkely path by conuderations and uncertainties related to waste em-which most radionuclides from disposed waste rnight placement, momtoring, and performance assessment that reach the environment, the N1(C is actively studying the are unique in the history of the NI(C. Much of this movement of pround water in the unsaturated fractured uniqueness stems from the type of f acility, hrst-of its kind media currently under consideration by DOli. An experi-peologie disposal installation, its very long performance mental site has been hicated in unsaturated fractured tuff ume hpecified as 10 000 years oy the liPA), and the fact in Arizona, where field and laboratory testing is being t hat it will be placed in low permeability / low flow peologic conducted by the University of Arizona.The objective of systems that have not been investigated previously be-the field study is to determine what types of measure-eause of their low economic value. The NI(C must have ments are needed to characteriec the hydrology of frac-an independent calubility to evaluate the DOli safety tured media and how measurement data should be ana-analyses and decide w hether long term releases predicted ly/ed to model ground-water flow. This work currently by DOli will be within established litnits. The NI(C re-includes (1) assessment of techniques and methods for search program objectis e is to provide the technical capa-fracture characteritation,(2) mhbration and percolation bility necessary to evaluate DOIPs site characterization studies, and (3) rock matrix permeability testing. The activities as required by the NWPAA and to assess Doll's project is using numerical simulations of vapor and lig-license apphcation w hen it is submitted.
uid-phase flow and transport in partially saturated media 29 NUltiiG-1266
to assess the impirlance of larre, natural, anomalous curred through the accumulation of joint shear deforma-hydrogeoh ic features. the appropnateness of contin-non resulting fiorn repetitive hudmp. A field site in a t
uum versus discrete f racture models. measurernents of inins ubject to rock slabbing is being instrumented to effectne parameters theories of spatially varymg dnper-study ts offeet of seismic loading on rock dnplaccment, sion measurements, and distmetions hetween and among rock stress, and pore water pressure response, matnx ditfusion, dispersmn, and sorption.
4.1.3.3 Staling of Itor(hnles and Shafts I s Tuff 1(esearch iwestigations completed during IT 1990 have shown that modek of radmnuchJe dispersion bued on an
.Ihe isolation of nuclear waste in deep peologie reposito-analory with I ick's hw of Alfnen t he convent vm! ww nes may require that penetrations into the host rock for.
I to estiinate such dispersion, may be wrong and that non; mation-such as shafts, drifts, ramps, and boreholes m
}
Fakian estimates of dispersion may be important in the Mniy of the repository-be scaled to prevent the evaluations of eompliance with the lipA lli.W standard's creation of potential pathways for the migration of i
10,000qear cumulative limits on radionuclide release to radionuclides to the accessible environment.
)
the biosphere. The correctness of dispetuon models may j
becorne more important if the I PA decides to include To evaluate the performance of seals in the unsaturated I
concent ation-based safety criteria in its film, standard' llLW tuff environment, the NRC is supporting research j
Research is continuing to determine what the best alter-studies at the University of Arizona. Iloth laboratory and natives to Fkksan dispetuon models are.
fictd performances of seals are being conducted. Charae.
terization testing confirms that tuffis an extremely het-(
)
InvestiEalors at the Center for Nuclear Waste ReIwlato'I cropencous rock with highly variable properties and ex-l Analyses (CN%,R A) m San Antomo, l'exas, are examin-tremely low hydrauhc conductivity. Mixtures of bentonite mg metho-Js to perf orm stochastic hy drologic analyses for and crushed tuff show that sampics containing 25 or 35 repository scale systems.
percent bentonite have permeability of the sarne orderof magnitude as similarly prepared and emplaced samples The validity and applicabihty of the models used to de-containing bentonite only.
scribe ground-water flow and nahonuclide transport is I
being appraised in an international project called IN-4'h3 A ""d' U"Ck"E' E*'I"""""C' j
TRAVAlJihe NRC staff and rescarch contractors from The CNWR A completed program plans for its integrated i
the CNWRA, the University of Ari/ona. Sandia National iILW package testing in IT 1990. Some preliminary ex-j Laboratories, Mawach u set ts instit'it c of Technology. a nd periments were done to examine general corrosion in 11atttile Pacific Northw est I aboratories are participating poirs proposed container materials 304L and 316L l
l in the 13-country validation effort-stainless steel; Incoloy 825; the copper alkiys CDA 102, l_
613, and 715; and the " reference" material, llastelk>y Cooperatise expenments and data analyses being done C-22. A literature assessment of pitting corrosion, stress j
under a cooperative agreement betw een NAGRA (Swit*
corrosion cracking, and metal stability of the DOE-pro-i tentand)and the NRC, negotiated during IT 1987, con-posed container materials was begun, j
tinue to augment the field testing program cited above.
{
4.1.3.5 C. chemistry 4.1,3.2 Stability of Underground Openings Knowledge and application of geochemistry is important When specifying suitable site conditions for a repository, to an understandmg of many aspects of repository 10 CFR Part 60 specifically regnires consideration of performance, including problems related to waste pack.
natural phenomena and site conditions that could ad-age corrosion, radionuclide release and transport, and l
j versely af fect achievement of the presenbed performance alteration of ground water flow paths as a result of min-j objectives. An important phenomenon that could affect cral dissolution or precipitation following waste emplace-both the short-and long-term performance of a reposi-ment. The NRC has an active research program in l
tory is ground motion resulting from seismic activity.
geochemistry as it affects the management of IILW. In Similarly, ground motion caused by underground nuclear 1990 the thermodynamic data base used to predict explosions at the Nevada Test Site needs to be evaluated, chemical reactions in tuff and ground water in the
[
Ground motion from seismic activity could cause rock thermally affected arca of an llLW repository was exam-displacement, nse in water tables, etc, which could vio-ined, and research into the modeling of the evolution of l
late the established repository performance objectives.
water as it moves toward the waste packages was started at j
the CNWRA. The NRC is participating in an interna-To investigate the effects of seismicity on the under-tional field study at the Koongarra on e body in northern
[
ground openings for an til W repository, the NRC is Australia, observing the actual movement of radio-sponsoring research at the CNWR A. Initial results from nuclides. This study is providing a basis for validating the study mdicate that structural damage at depth is in-performance assessment models to be used in llLW i
NURiiG-1266 30 u-
- a..
repoutor) hcensmp. the second ycar of the study has Second, the States and sompacts of States have chosen to seen the complet on of the hydrologic anJ prothemistry conuJu alternatise drposal mtthods to sh.dhm land inoJehnp scenanos. A rooJ deal of the held data on burial. Certain of these alternatives must be critically transport properties of the site h.nc already been col-examined by tightly focused research to determme their letted. At Oak itiJre National Iaboratory, laboratory acceptability and to rive guidance to the States and com-studies surycsted that there inay be a practical approach pacts.
for simphlymg coupled hydropeochemical modch of radionuclide trantort at Yucca Mountam, Nevada. A
'the direction of the ILW research program has re-study was begun at Johns llopkins Unnetuty to des elop sponded to legislative action (P.I 99-240h the changing counted hsdrocca hemmal transnort nWeh and to test polin of States now tcymnsible for disposal, and the ther'n apauht dina from natural s) items such as the Koon.
lessons learned frorn the history of shallow land burial of g.u ra or e boJy.
wastes at a number of sites for ses eral decadesNague and differmg criteria as to Mte suitability, waste package de-4.1.34 Pu formance Asement sign, etc, hau b en emphyed and may charactente fu-ture efforts.
In IT 1940, the NRC transf errcJ the task of mamtaming
.md improung its Ill.W performance awessment meth.
Ihsposal criteria for 11W have evolved as experience, oJohiry f rom Sandu National I ahoratones to the NRc knowledge, public awareness, and political controversy stall and CNWRA.
hm e grown. In particular, through the i owlevel Radio-active Waste Policy Amendments Act of 1985, the Con-4.1.3.7 Rulemaking gress has required Ihe NRC to provide guidance for regu, latory decisionmaking regardmp engineered IL%
A proposed g uide, t empleted in September 1940 for pub-ditos:d methods. 'lhis change has broadened the scope lication for pubhc comment in Nm ember 1900, gives the of NRC 11W research.
infonnanon needed by the NRC to reuew DOlPs license apphcation for the lil.W repository.'lhe NRC continued 4.2.2 l'rogram Strategy to closely momtor l PNs development of a revised high.
levcl radioactise waste stanJaid. The NRC stafi com.
NRC research in support of licensing activities for Lt.W m;cnted on i PA woikiny draf ts of the standard. In April disposal facilities is examining enhancements and alter-1000, the NRC receiveJ a petition far rulemaking from natn es to shallow land burial, l.l.W waste forms, infiltra-IX)l: to amend Part 60 to mclude specific accidem dose tion of water, ILW source term modeling, hydrologie critena for the design of engineer ed safety features at the flow and contammant transport, and perf ormance assess-lil.W geologic repoutory. In July 1490, a petition was ment. 'the NRC's 11W research staff also prepares recened for rulemaking from the States of Washmgton rulemakings that affect ILW disposal. The NRC 11W and Oteron. 'lhe petition asks the NRC to undertake research pregram is desenbcd in NURl!G-13SO, which rulemakmg terardmg the clauification of some radioac-was published in November 1989. NURl!G-1380 identi-tive wastes at 1 M11ilacihties at ilanford, Washmpton, fics issues, regulatory needs, a strategy, and a schedule for resolvmg them.11 was widely circulated for comment to potentially interested parties, including State regulatory 4.2 I AW-l,CVol W3 SIC agencies. ' RC-funded I LW research is useful not only to N
the NRC licensing stalf but also to the States regulating 4.2.1 Statement of l'roblern ILW disposal. In order to make their research results available to the States, NRC research contractors, be-Ihsposal of low-lesel waste (1.1.W) invohes issues con-udes publishing their work, gave presentations at meet-cer nmg waste for m and waste package mtegrity, transport ings well attended by State representathes-such as of radionuclides through the dnposal facihty erniron-
" Waste htanagement '90," the Annual DOlilLW hian-ment, anJ evaluation of long term doses from releases of agement Conference, and the first annual 1.1.W research radionuclides beyond the dsposd facdity environment.
review meeting organi/ed by Rl!S.
Research is required to estabbsh regulatory criteria and
= license appiication awessment information to permit The diverse I LW regulatory user community makes the sound evaluation of propowds for dnposal facilities and to coordination and defmition of ILW research and the ensure that all reputatory requirements, panicularly dissemination of associated proJucts a much more corn-those on radionuchJe release limits, will be met. Per-plicated undertaking than similar activities for the lit.W forming the needed research in a timely manner is made program. liceause many States are licensors of ILW dis-more urgent and complex by two factors. l'irst, the 1.ow-posal and at e looking to the N RC for technical support in Level Radioactn e Waste Pohey Amendments Act of 10S5 their iLW licensing and egulatory programs, NRC's (P.l.90-240) sets a sery tight schedule for establishing i LW research has to be more prescriptive and develop-facdities within individual States or compacts of States.
mental than the lil.W research program.
31 NURiE1266
4.2.3 Researcli Accomplisitments in IT 1990 llowever, its long term performance needs to be as-sessed.
4.2.3.1 1:ngince:Ing 1:nhancements and Alternathes to Shallow I.and llus ial 4.23.4 Performance Assessment
'lhete is great interest on the part of States and State Research was begun on a performance assessment meth-compacts in attematises to shallow land burial for the odology that will mtegrate results from other NRC llW dispwd of low.lesel nuc! car waste. Cones ete is expected rescatch projects, limphasis is being given to engineered to play an important role in these engineered enhancements to shallow land burial. 'lhe Sandia alternatives. In 1990. the National Institute of Standards National Iahoratories are assessing the validity of per.
and Technology (NIST) continued investigating, for the formance assessment models. 'lhe Pactfic Northwest NRC, the durability of concrete for 11W applications, laboratory (PNI.) is exploring mathematical models for j
and the Idaho National lingineering laboratory contin.
radionuclide transport through concrete.'lhe Massachu-ued to develop a mathematical model describing concrete set ts i n st it ut e of Technology (hi rl') has been investigating 4
l performance. In a NUREGICit report, NIST has made the use of stochastic methods for dealing with large scale recommendations on the seicetion of apperates for con.
heterogeneity of site hydrologic flow parameters. ~lhe i
crete.
University of Ari/ona and New hicxico State University are working cooperatively with hitT by providmg a field 4.23.2 11W Waste l'orms test at Ias Cruces, New hierico, of MIT's theoretical wor k.
i 1 aw. level radioactive waste collected from operating nu-clear power stations and solidtfied in cement is being 4.23.5 I I.W Source Term Modeling tested at the Idaho National lingineering laboratory.
'the studies are aimed at ensuring that radionuclide and Ihelopment of the MW soun'c tenn coJe,111l1 con-chemical leaching characteristics. and the compressive u.nues. Oc Hnmkhaven NationalIahoratory has refined strength of the solidified waste, ate consistent with NRC and espanded the leach submodel to incorporate four I
technical pmtions and requirements of 10 CFR Part 61 new ddfusion controlled release models and a coupled i
for waste for m stabihty. Under examination is the stability dissolupon-diffusion release rnodel. To provide confi-i of decontamination waste obtained frora the Ilrunswick dence m Ihe model predictions, the lilll' code continues t be benchmarked against 1ysimeter experiments of 1
(North Carolina) and Fit / Patrick (New York) nuclear reactors using commercial decontamination processes sahstone waste forms at the h,avannah River laboratory and solidtfied m cement. Samples of solidified decontami-and cement, bitumen, and polymer waste forms at PNI.
nated ILW were collected from the Peach llottorn llesults of sensitivity analyses continue to be used to l
(' Pennsylvania) nuclear station, and the samples are un.
"S"s radionuchde scleases as a function of key parame.
J dergoing curing before testing is statted. Field studies ters..Ihis work rcpresents a first attempt at quantification using lysimeters are being conducted at the Chk Ridge of soum tenns for use m performance assessment, and Argonne f lational laboratories to determine ion exchange resms are 4.23.6 Hydrohmy and Contaminant Transport whether radionu:lides on released from soddified waste forms under environ-The NRC continues to sponsor field tests of flow and mental conditions avolving natural precipitation. the transport in unsaturated soils at a New Mexico State effects of radiat on on the stability of ion-exchange resins University field site near 1as Cruces, New Mexico. The contammg radioactive material were completed by the program, which includes NRC sp msored research by l
Idaho National lingineering I almratory.
PN1, the University of Arizona, and MIT,is intended to i
confirm the reliability of unsaturated flow and transport 4.2JJ Infiltration of Water models of 11W dispos:d facilities. Some of the data from this work is being used in INTRAVAl, an international
'lhe University of California at flerkeley, in cooperation study that deals with model validation of ground water
+
with thc University of Maryland,is continuing to field test flow and transport models.
a variety of covers for 11W disposal units at the Maryland l
Agricultural lixperiment Station in lleltssille, Maryland.
4.2.3.7 Rulemaking i
Results are reported in NURl!G/CR-4918, Volume 3.
Two designs are proving to be particularly effective. One, Final amendments to 10 CFR Part 40 that provide licer's.
called bioengineering water management, not only re-ing for the custody and long-term care of uranium and duced water infiltration to a negligible amount but also thorium mill tailings disposal sites were prepared for dewatered two experimental cells. A second cover con.
publication. A draft revision of Regulatory Guide 4.18 I
sists of a resistive layer barrier (compacted clay) over a that provides the standard format and ccmtent for i-conductise layer barrier. 'lhis second system has func.
environmental reports for near-surface dispsd of radio-tioned perfectly since its completion in January 1990.
active waste was developed.
i
. NURl!G-1266 32 4
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5 ItESOINING SAFETY ISSUES AND DEVEl.Ol'ING llEGUIATIONS
'Dns program is directed toward the development of the issues.1he results of the NRC's contmuing effort to techrucal hvis and relateJ regulatory requirements identify sigmficant GSis will be meluded m future supple-neeJed to protect the health and safety of the pubhc from ments to NURI'.G-0933. Durmg 1990 the NRC identihed radiation and from the risk resulting from the generation three new GSis, established pooritics for 16 iwues (see of electricity and the rnanufacture, use, transport, and Table 5.1), and resolved seven GSis (see Table 5.2). In storage of nuclear fuel and other radioactive materials.
addition, one GSi scheduled for tesolution wasintegrated
'!his program also supports efforts to ensure that pro-into the action plans for resolution as part of other GSis.
oosed Commission regulations are ady.ne and that ihn are docloped m an ethetent and timely manner.
5.2 Standardized and Advanced D"d"f 5 5.1 Generic Safety issues 5.1.1 Statement of l'roblent 5.2.1 Statement of I'roblem lhe Commission directed the NRC staff to prepare a The Commission has issued a policy statement on the prionty list of all genene safety issues, includmg TMI-te, regulation of advanced nuclear power plants (51 17R lated issues, based on the pote' tial safety significance and 24643), which states that the NRC will reuew and com-n cost of implementation of each issue. In December 1983, ment on new design concepts. It also encourages early the listing was approved by the Comminion. 'lhis guid.
interaction with apphcants. As part of this program, the ance is renceted m the NRC Policy and Planning GuiJ-N RC will develop, review, and implement advanced reac-ance, the NRC Strategic Plan, and the NRC l'ive-Year tor safety and policy issues in the ongoing NRC review of pian.
advanced reactor concepts (NUREG-1226). In-depth in-dependent analysis will he performed as necessary to o.l.2 l'rogram Strategy verify that advanced reactor designs have the potential for enhanced margins of safety and that appropriate means A genene safety issue is one that involves a safety concern wiH be rsed to accomplish their safety function.
that may affect the design, construction, or operation of all, ses cral. or a class of reactors or facihties and rnay have lhe Cournission has issued a policy statement on stan-a potential for safety improvements and issuance of new dardization (521 R 34884). The purpose of the policy or reused requirementsor guidance.Timelv resolution of statement is to encourage standardization and to provide these issues is a major NRC concern. A prioritization and for certification of plant designs that are essentially com-management process has been established for identifying plete in both scope and level of detail. The Commission important issues for immediate action, for climinating also added a new part to its regulations,10 CI R Part 52 non safety-related or non-cost effective and duplicate is.
(54 FR 15372), which provides for issuance of early site sues from further consideration, and for keeping the permits, standard design certifications, and combined Commission and the publicinformed of the resolution of construction permits and operating licenses, with condi-these issues. Currently, a backlog of approumately 30 tions, for nuc! car power reactors.
proposed genene issues is awaiting priontization. Strate-gies for this program are to proviJe timely prioritization 5.2.2 l'rogram Strategy of proposed new genent issues, eliminate the backlog of proposed issues (as resources permit), and issue periodic The Department of linergy (DGli) has submitted three updates on the status and progress toward resolution of advanced design concepts for NRC review. These are the generic safety issues.
Sodium Advanced Fast Reactor (SAFR), Power Reactor Inherently Safe Module (PRISM), and Modular liigh-5.1.3 Research Accomplishments in FY 1990 Temperature Gas-Cooled Reactor (M11TGR), Ann Combustion lingineering submitted a preliminary safety The NRC continued to use the methodology set out in the information document for the Process inherent Ultimate 1982 NRC Annual Report for determining the priority of Safety (PIUS) reactor design. 'lhe strategies for this pro-generic safety issues (GSist in December 19X3, a com-gram are: (1) conceptual design review of these plants,(2) prehensise hst of the issues subjected to this method was identification of major issues that need to be addressed published in "A Prioriti/ation of Gencrie Safety issues" pnor to hcensing,(3) identification of design features that (N! !R l!G-0933), w hich is updated semiannually (supp!c-should be verified and research that needs to be per-ments in June and December). The list of issues includes formed, and (4) issuance of safety evaluation reports
'1311 Action Plan (NUREG-0660) items and new generic (SliRs).
33 NURiiG-1256
Table 5.1 issues prioritired in IT 1990.
Numhtr
'l ille l's ior it) 63 Cse of !~quipment Not Classified as 1:ssential to Safety Di(OP in 11Wlt 'l ranuent Analysis 71 I ailme of itesin 1)eminerah/cr Systems and 'lheir 1:llects on I ()W l
Nuclear l'ower Plant Safety 61 Impact of I aked Doors and liariiers on Plant and Personnel Safety Dit( )l' 45 1 oss of 1.lfective Volume for Containment iteeirculation Spray Iti:SOI.VliD oi 11111t Suetion Vahe Testing Cosered in lu.ue 10$
r 107 Generie Impheations of hiain 't ransformer I:adures 1 ( )W i
109 iteactor Vessel Closure 1:ailure DitOl' l
116 Accident hianagement Cosered in Severe Accident l'iogram i
l17 Allowable Time for Diverse Simultaneous liquipment Outages Di(OP 129 Valve Interlocks to Prevent Vessel Drainare During Shutdown Cooling DitOP 137 itef uchng Cavity Serd l'adure DilOP i
l 140 lisuon ProJuet itemovat Systems DitOP l
141 1111 OCA With Consequential SGTil DitOP 142 1.eakage 'thr ough 1.lectrical isolator s h11:DIUh1 11-2 9 lif fectiveness of Ultimate lleat Sinks 1.icensing issue 11 -3 2 Ice 1:fIcets on Safety ltelated Water Supplies Cosered in issue 153 I
i Table 5.2 Generic safety issues resobed in lY 1990.
Number Title 70 POltV and lilock Valve Iteliabihiv i
75 Generie implications of ATWS livents at the Salem Nuc! car Pov.er Plaat S4 Cli POltVs k
94 Additional 1 aw Temperature Overpressure Protection for 1. Wits 103 Design for Probable hiaximum Precipitation A-29 Nuclear Power Plant Design for the lleduction of Vulnerabihty to Industrial Sabotage C-8 hiain Steam lane isolation Valve i.eakage Control Systems Parallel with the policy development elfort. Nillt is re-house SP 40, and the Cl! system 80+ ) and design re-viewing three standard designs (AllWif, the Westing-quirements prepared for standard plants by liPiti.
NUlti!O-1266 34
Genene issues pot;unmg to stand.udution requamt arrheation for deopo cenkation for the CANDU 3 further attention are expcitcJ to emerge from these re-dcVpo M I40?
uews. I he stratery for ttus prograni is to resohe penetw iwucs pertaming to standardvation arismp f rom the on-rmnr st mJard plant renews.
U l,tiel L,yele,,I,r;uisportation, im(I Slifeguarcis 5.2.3 1(eseareli Accomplishments in IT 1990 5.3.1 Statenient of l'roblem 5.2.3.1 Adiarned l{entor Conu pts and Idfectise regulation of f uel gi,le, trantlvrtation, and Standarditation safeguards actanties mvohes the tatk of planning, devel-
'lhe NRC stall has wmpleted its detaded technical oping. and issumg appropna t e r epulahiry poutions. lkmg renew of three advanced reactor concepts that were sub-information renerated mtertially or thrbuch nattowly di-nutted by the DO!Elle rcuew s sought to deterrnine the rected researth, new posituins be develo' ped or existing acceptabihty and htensability of these unique aJvanced posihons ate modified. These positions can take the form i
reacter deurns.
of regulatory requacments, pobey staternents. puidance, or criteria for actinties willon this element. Specific ac-The DO!!conceptualdesynsincluJe two advancedliquid tinties melude fuel cycle of heensed nuclear facihties, metal reactors (PRISN1 and sal R) and one advanced transporting radnuetive matenah, and safeguarding. Ia-eMhes and special nuclear materials. Setting of prionties inoJular high temperature pas-cooled reactor (hillIGR). Draf t safety evaluation reports (SI:lts) for for regulatory needs or dehciencu s are undertaken to hilll GR, NURI.G-1335, and PRISN1, NURI'G-136S, ensure that the problems of prcatest significance to the public health and safety or the common defense and base been issued. 'lhe draft S1:11 for SAlR (NURl:G-136W wdl be iuued in early IT 1991 to docu~
security are addr esseJ in an espeditious manner through ment the wor k completed todalc and elosc out this teuew propeily defmed reputatory and supporting iescanh pro-
{ since Doli has stopped work on the SAI R advanced
- prams, reactor design. 'the Adusory Committee on Reactor Safeguards has issued letters on all three of these concep-5.3.2 l'rogntm Strategy tual designsflhrec ropportinp N URI-G reports have als" been issued: NURI G/CR-5261, " Safety 1 valuation of The program strategy for emh of the activities in this element is: the NRC needs to des etop or moJify regula.
hilU GR 1.icensmp liasis Accident Scenanos"; NURl!G/
CR-53M, " Summary of Advanced 1.h1R livaluations~
tory requnements and guidance to protect workers and PRISh1 and sal'R", anJ NURiiG/ Cit-5514 "hloJelinE the public from radiation risks associated with the fuel and Per f ormance of the h11l'1 GR Reactor Casity Coohng cycle of heensed nuclear lacihties.
System."
In the area of transportation of nuclear material, the U.S.
in 1990, dol:. prouJed the NRC with two adJitional Trade Apreements Act of 1979 directs 1 ederal agencies
- amendments to the PRISh1 prehmmary safety informa-to develop standards that are internationally consistent,
! tion document that adJress the open issues identif ed in whenever appropnate. A proposed rule has been issued
' NURl;G-136N and desenbe thanges to the l'RISh1 de-to revisc the transportation regulations for low-specific.
. sign.The NRC staffis renewmg these design changes and activity material and ensure that the properties of the
! plans to issue a reused St.R in 199L DOli also an-radioactive matertals being shipped and the packapes nounced that they are reassessing the hill'lGR program used m shipment aJequately protect the pubbe and the and performing a cost reduction study on the hilllGR occupational health and safetyJlhe final rule will achies e design. 'the NRC staff assessed the research needs for maumum compatibility betw'een NRC regulations and the transportation repulations of the international these designs and deter mmed w hat confirmatory tesearch Atomic linergy Agency (I Ali A).
should be per formed in order to support future hcensmg reviews.
The purpose of the safegt ards program is to issue changes or additions to safeguards pohty, regulations, or The staff :s also deseloping research plans for Westing-puidance that meet the needs of the Offices of Nuclear house's APbOO and General lilectric's SilWR advanced Reactor Regulanon and Nuclear hiaterial Safety and hght water reactordesigns.The NRC staff was notified by Safeguards. The grategies are to (D determine that AllB Combustion I!ngmeenng that they plan to submit physical sceuiity and accountabihty of strategic special appheations for l'inal Design Approval and Design Certi-nuclear matenal (SSNhti remain adequate; (2) ensure fication for the PlUS reactor in IT 1992 and the Safe that the s alues of secuoty, physical protection, and mate-Integral iteactor (SIR) in IT 1993. 'lhe staff was also rial control and accountabihty at e balanced aramst imple-notihed by Al:Cl. Technologies that they plan to file an mentation costs; and (3) develop or mosfy safeguards 35 NURI G-r260
regulatory requuements and ruiJance to be internally accompanymy reputatory ruide base been appuned by womit nt.
the Commmm The rtilemakmg o f o owmp an uccler.
n ated scIvdule sms e a hcense appheation f or t he constr uc.
5..U Hesearcli Accoinplislirnents ist lY 1990 U"" ""d "P"d'"" UI " f as cennifure phmt that w ou1J produce low cr nched uratuum for the commerual mar-6.3.3.1 l'uti C p le ket is anticipated in the near f uture. it is cspected that the proposed rulemaking will be pubhsheJ m the Tedcral
'lwo aports relatmg to decontammation have been is.
Rcgntcr for pubhe comment in early lY 1991.
sucJ. I he first, osm d m Noumber 1959, was NURl:G/
CR-3444 on the unpact of IER decontaminations on sohdihcstion, w aste dnpos.d. and auouateJ occupational gj g)g.pggp g g,gg g ggp,.gg, g
uposure,'this :eport prmides the fmdings of an msesti.
kll!UbllS ranon of therrnal escutsions durmg dnmg of contam!-
nated resm wastes and of the compatibihty of a range of 5.4.1 Sintenten' of I'roblet:1 wotamer matenab. with decontammation rcsins. Tbc sewnd, iuued m December 1989, is NURI.G/C R-54h3 lil!S has the primarv responsibihty to manare.coordmate reviews of, and cont'rol all NRC rtileming activities and on raJionuchde buddup in llWR reactor coolant recir-culation pipmp. lhis report proviJes the f mdings of stud-to monitor scheduling of such rulemakmps to ensure that ies conducted to determine recontamination and dose rules are developed in a timelv manner. In addition, Rl'.S hum >tt for pr eparation of the regulatory impact j
(hanres followmg 11% R t hemical decontamination proc-analyses (RI As) that accompany all rulernaking through l
the development of generic rnethodology and guidance.
I Technical reviews of all RIAs are performed upon re.
t 6.3.3.2 Transportatinn and Sahguards quest. The NRC Regulatory Arenda Report and other A hnal rule,10 Cl~R Part 71, on moJdying NRC's trans.
Inanagement informahon systems awociated with i
pottation regulations was delayed until the Department rukmaking activities are mamtained.
of 'l ranspoitanon tuued a comparuon rule for pubhc comrnent. Pubhe comments on the proposed rulemaking Needed regulatory products, e.g., regulations and r egula-base been esaluated, and the final rule is being devel.
tory guides, are developed. Rulemaking is proposed or eped. I his rule proposes hmitations on the shipment of initiated, as appropriate, and camples rulemakings that span the techmcal or organitational responsibilities of low.speubc.acuvyty matenals and maumizes companhil-ity between NRC anJ I Al:A regulations.
several groups or that involve novel or complex questions of
'ulatoy pohey are managed. Peutions for rulemak-A f mal rule,10 CFR Part 73, on accew authoruation at
'"E *
- rmclear power plants and an accompanying regulatory rtnJe hase been developed for the Commission's ap'.
5.4.2 l'rogram Strategy proval.'the rule requires a nuclear pow er reactor hcensee The purpose of the NRC nuclear regulatory program is to to have an access authorvation program in its site physical ensure that nuclear facilities are designed, constructed, security plan. Tins woulJ proviJe increased assurance and operated in a safe manner. 'lherefore, there exists a that persons granted unescorted access to protected and continmng need to revise rules and guides and to develop j
vital areas are trustworthy and do not pose a threat to new ones. 'lhe strategies of this program are: (1) review mmrno n.*olodcal uhopu:e. I's mpected that the final
,i,p gregyg,e ef IMR =,cgulatory tequirements and rule will be pubbshed in calendar year 1991.
guidance and make recommendations for revisions; (2)
A fmal rule,10 CI:R Parts 70 and 74, on the centrahration develop screening methodology to systematically review requircments and guidance; (3) coordinate and review of matenalcontroland accounting (Mrk A)ticensmg and proposed changes to the I Al!A safety standards; (4) de-mspection activities for nonreactor facilities was pub-velop or assist the development of rules and regulatory lished in the federal Register in February 1990. 'lhis guides; and (5) continue to develop and maintain manage-rulemaking completed the phased centralization of re' ment information systems for rulemaking.
gional MC& A actisines in lleadquarters for nonreactor facilities 'lhis action was necessary because, given the small number of facihties in the respective regions, the 5.4.3 Research Accomplistiments in IT 1990 regional offices could not support and maintain the full M3d Dmlop w MWify Regulations spectrum of knowledge, skdis, and disciplines needed to conduct MCA A licensing and inspection.
In a program initiated in 1985 and continued through 1989, the NRC staff undertook to evaluate custing regu-A proposed rulernaking,10 CFR Part 74, on the MC& A latory requirements in terms of their risk effectiveness requirements for uranium enrichment facilities and an and to eliminate or modify requirements with only a NURI G-1266 36 I
marpm.d ufety importance. A threc.udume rescars h re.
mJutt ry unt utw..uiJ prortcu m nuinte nant e imprm e.
part (Nill (! (i ( }( 4330) prouJcJ det.oled tec hmcal.u ments.md recsalu :tt the nct J f or mumr a bn d rutcM.
M wntnts of requncnu nts aum ulcJ with a numh r el my t
topws In a follow on t flor t. a < t i of r c; itory rcquu c-ments wn iJt nuhrd n canJulates tor pa lle i honna 11 n t h at t b a a mor nced emb h o,W tu d spcnt f ocl tion or mn!dwauon Wer k w n untutcJ m 1900 to culu-
'h""f C 'P 'a at tenunt n ul nu&s pow a rem 1or ' des ate the safety signihcance of these candidite rcrulations to be.o;nWe ni the near future in mponw to thn nni to identify those of marrinal ufets sn'nif nan,e and for Ow Nudcar Waste lyhcy Act of 1%2 direttcJ the Sccre-w hit h moJtficatmn or ( hmirution can be pnipo' t d. Itec.
tar), licpartment of I nerry, to estaNish a dry (pent fuel ommendations wdi be forwatJcd to t ic ( ornomsnm m M"'JEC *""'n<uanon inognun, w d tQ ohn tm of pgg l y jty;7 cornlMf up with orlC or Irlor e IcchmdoPC$ tbal Ihe bM nurht apprme for use at cnihan nuclear poact reactor in May 1990, the t 'omnnumn puhh6 a propo<cd rule
$dC' "HI"'ut, to the estent prm ticable, regumng aNi-that would amend the crulanor. m 10('l it Parts N. 3(L nonal utcaprofic app oA A hn:d rule, 'O Cl l< 1.at 40, and 'O to res ne lht nsee ro porting ret nrt ments re.
E, puyicd in tk IolauWma in My PhD w d aHow gardmp notifications of mcidents r t lalcLI to raJution holdm of nudcat power reador operating ken'a to i
Lif tt). 'lhis u dl ensure that syrnh ant ott uncnns at stot e spent fuel m Ni(U-apprm cJ ca'Ls at rea tor utes fauh' tics operateJ by nutetul incrot es me pnunptly rc.
unJer a renes.d hccrac.
ported to the NRC. lhe ('omanwon udl be ahic to l
determme whetht: a hccrme tus taken athons nt tewary A pm oseJ rule w.n pubbshed m the Ib!aul Recorrr m hum 1000 ht wold mend tk 10 El R Pd 35 to pmtett pubhc hcahh and salcty and whetta r rencrie salcty conecrns are iJenuhcJ ttut may requhc prompt ude At apply to the mcdwal use of byproduct Nlt( acuen' netui. t he amendrncras would requhe medical ute l
I m mm@matpluvwumce m m m i
l
.lhe (.(iMirmuh $n has appnntd a pnyned rtdc that anJ would revise nusadmmbtration reportme require.
woulJ amenJ the terulations to Part.'O to rtquuc the ments. Implementation of the new perf onna'nce. based r equa rments would be suplW tcJ by P^uance lif a refula-hcensee to unplemcnt the NRs,.
ton ymde that woulJ mclude 'peahe rmd mce for OA l(esponse 11ata System (1 RI)h. -appuncd 1 merrenes
) at aH nude.o power plants.I he pomary nde of the N Rt.dunng an emerpency prornons lhe rule would enhance panent safety wlnic al a htenscJ nuclear power laahty n one of momtonny allowmp the lleubihty necessay for proper methcal care.
the hcensee to ensure that appropnate ccomrnenJations I.he leaMbillt) o[ this appioXh n beinf esaluatedilunng a are made with respect to onesso) offute acnons to pro.
[ dot prof rain insolvinf i0 rUcdlCal-usC licensees.l he l
teet pubbe health and safety. ILHI)5 would supplernent
(,omuusuon espccis to tonuder a fmal rule m March pg,'
l the soice transnnssion m et the existmp 1 merpera) Noti-I fication Systern by traranuttmp timely and accurate up.
An mtet un fmal r ule was pubhshed m the Inlen;/ /he!3ra dates of cnucal infonnanon on plant wnshuons from the hcensee's onute computer to the NR(' ()perations l'en-m Auguq F 00amendme 10 El R Patt 35 in icsponse to a pention for iulemaking l'som SNM/Al NP that requested ter. lhe proposcd rule was pab!b bed in the lh!cnillWM-departures f rom mstrucuons approscJ by the 1 ood and j in for pubhe comment on ()ctober u, 1900-
- 1) rug AJministranon (l~l)A) for raJiopharmaceuucals so i
that physicuns may rrmidt pmper medical cart to pa-g in Man h 14SN the Commiwon nsued a Pohey Statement uents. This rule, umco was coorthnated with the 1 !)A m on the Mamtenance of Nudcar Power Piants. In the statemcob the t 'ommiwon maicated us imenuoa to pu" us de,elopmen'. M.u.ug am depo turcs from F11A ap-
+
proved mstructwns whilt prmiJ ng icasonable awurance sue a rulemaking on mamtenance. In descloping ttus of raJiological safety as wcH as a balante between ade-i proposed rulemaking, the stall haJ cstensis e mteracuons quate controls and amidance of undue interIcrence m wah U.S. mJustry (airime and nuJear) and stuched f or-mcJ teal judgment.
eign nuclear maintenance programs anJ practices. A 3-day public workshop was helJ m July 14M to sohcit A proposed rule,10 (T R Parts 31 and 32, on requne.
comments on rulemalmp optiont 'the mfor:nanon rath-ments for the powenion of industrial deuces contammg cred was used m formulatmg the pmpmed rule and us bypioduct matetial has been deseloped. This rule re-supprting repulatory rmJe. The Commimon nsued the quires pencral hcensees to provide NRC with specific proposed rule for pubhc conunent m November 143S.
information about radioactne material used under the
'lhe supporting draf t regulatory guide was pubhshed m proviuons that estabhsh general domestic licenses for August 1959. In 1)ecember 1934, the Commission iwued byptoduct matenal. 'l he pmposed action would impmve a reused pobey statement to restate us news with respect the pubhc health and safety by reducmp the potential for to maintenarce and to mdicate its intention to hold unnecewan public esposure to rachation that coulJ be rulemakmp in abe)ance for a penod of 18 months,1)unng caused by improper transfer, mamtenance, or dnposal of the IS-month ume mienal, the Comnussion wdi momtor the radioactise mdustrial devices. It is espected that the 37 NUR1h12hh g
prepuscJ tulcinakmg wdl be puhhshed m the /n/nal offices and malmp tecommendations foi awpnment of Rosto f or pubhc comment m early i Y 1991, requested rulemalmrs withm RlX l >urmg 1940,77 t ulemaking actions were procewed. ()!
A hnal rule was pabhshed m the lo!rrd /hnn t m (kto-these,17 final rules were pubhshed,5 were terimnated/
ber Im9 amenJug the 10 Cl R Part 35 regulations that wahdmn, and $$ are onpoing. The detaded suus of apply to the me jical use of byproduct matena!. 'the these reviews, as of September 30.19n0, is ptovided in amendment aJJs 9altadium-103 to $35.400 as a scaled Table 53.
sourte m seeds to he list of brachytherapy sources pt r-irotted for use m thi ticatment of un(cr.
In addition to the 55 ongomp rulemakmg actions, it is estimated that in lY 1991 ther e will be apptoumately 25 to 30 new rulemakings that will reqmre Rl:S teview and 5A3.2 Ilegulatory Anal sis FI)O approval for mitiation.
3 "the NRC lus, as one of its prime goals, Ihe responsibdity t
for the osersight of regulatory impact analyses (RIAs)iri 5A3A I30C"$C Nt"C""I support of regulatory actions fe g., rulemakmps, backfits, renene safety iwues or regulatory yuides). To emphasiec The NRC has been considering what requacments this real, the NRC has published operatmp procedures should be placed on nuclear power plants in the event for apenn use for the support and/or review of regulatory that licenses to operate beyond the 40-year term of the i
unpact analyses af fecting all reyulatory actions. In March origmal license should be granted. Pubhe wmments on 1940, the Commiwion established plans to revisc 1hese license renewal requirements have been solicited three i
operatmg procedures. These revisions wdl expand the times through the / rderal Repuct-the first time in con-guidance and structure of the existing operating proce-nection with seven major license renewal issues (pub-dures to better mtegrate backfit analysis requirements lished November 6,19h6) and the second as part of an and safety roal pohey consiJerations and to update the advance notice of proposed rulemaking (pubbshed methoJs anJ information bases for performinp regulatory August 29, 1988). The advance notice requested com-unpact analyses to refleet espenence gained oser the past ments on NURI'.G-1317 " Regulatory Options for several ) cars. Also, to aid NRC analysts in preparmg Ni ele, i Plant 1.icense Renewal," issued in August 19SS.
RI As improvements are bemg made to 1 ORECAST, a C unents were summarued and analyzed in NURl:G/
PC based cost evaluation model. This :.oftware package
- f. R 5332," Survey and Analysis of Public Comments on allow s for the quantification of cost inputs associatcJ with
- Rl!G-1317: Regulatory Options for Nuclear Plant a wide ranpc of new or revised regulatory requests.These e Renewal," issued m March 19S9. The third time renerie costing methods can be useful m quantifying im J when the NRC published the proposed rule foi pacts to both mJustry and the NRC. Desclopment and power plant license renewal on July 17, 1990 (55 m
s modification of these procedures coupled with existmg "43). 'lhe final rule (10 CI R Part 54)is expected to regulatory impact analysis methods will continue to be
, published in mid-1991. The following supporting rehned to facditate NRC decisionmaking in evaluating documents were issued with the proposed rule:
the need for and the effectiseness of a variety of regula-tory actions, includmg rulemalmg, standards develop-ment, and backfitting safety improvements on nuclear NUREG-1362, " Regulatory Analysis for Proposed power plants. During this report period, approximately 16 Rule on Nuclear Power Plant License Renewal,"
safety-relatcJ reputatory impact analyses (both initiated draft for comment July 1990.'this regulatory analy-and completed) have been processed.
sis provides the supporting mformation for the pro-posed rule that will define the NRC's requirements for renewing the operatmg licenses of commercial nuclear power plants.
5A33 Independent Resiew and Control of Rulemaking NUREG-1398, " Environmental Assessment for Proposed Rule on Nuclear Power Plant I.icense Re-Rl!S has the lead responsibility in the NRC for rulemak-newal," draft for comment, July 1990. This docu-ing. The control of rulemaking responsibilities includes ment provides an assessment of the possible envi-coordmating rulemakmg activities with the requesting ronmental effects of promulgating nuclear power N U R EG -1266 38
l
'luble 5.3 Ituh making mtions timeurd during IT 1990, llulemaking Atthitits Numba l~ mal 1(u!ctnalmrs Pubinhed 17 l< ule mak mps 'l e r nunat ed/ Wit hJra w n 5
()npomy 1(ulemaking Actions 55 Proposed Rulcrualmgs (26) i I mal Rulemakinys (19) kulemalmrs on ilold (10)
Tot d Rulemakings 77 plant beense renewal standards by the proposed r ule Department of Attorney General, State of hiame;andan rather than applymg iequirements in an ud hoc man-individual, Kenneth G. Sexton, PkD. The Citizens Task ner in indmdual licenung actions.
l'orce petition (PRhi-50-31) teques,ed that (1) the emer-ger ey planning tone (liPZ) radms around noticar power NUlilli 1411 "Remonse to Public romrnents Re-plants be extended f rom 10 mik e to 20 rniles (2)inJc-e sung trorn the Pubbe Wor Lshop on Nuclear Power penJent radiolorical inomtoring systems operated by lo-Plant 1 icense Penewal," July 1940. This document cal communities be established, and (3) mandatory utihty repor ts the NRt"s re'ponse to the public comments funding of the emerrency preparedness c(forts of local from the November 13-14, 1 % 9 pubhe workshop comrutmities be required. The petition submitted by Dr.
and wntten comments on the workshop received Sexton (PRht-5(MS) requested that the si/c of the shortly thereafter.
plume esposure pathway i PZ be determined on a ute.
specific basis, usmg the most up to date inethodalogies, NURI G-M12, "I oundation for the Adequacy of and that the si/c of the liPZ be reevaluated at least every e
the l acenung liasis," draft for comment, July 1990.
5 years. The petition submitted by the State of hiaine This anahs:s describes the reculatory basis 'for the (PRht-50-46)(1) requested expanMon of the emerpency
~
reatne finding for all nuclear power plants that the plannmg zone for both plume esposure pathway and for imdmps of reasonable awurant e of adequate pr otee.
the inpestion pathway;(2) r equired that emergency plan-ton for issuance of an operatmg beense continue to ning be done before any construction of a nuclear f aedity be true at the tune of the renewal application anJ is permitted and that the povernor or governors of any accorthnph need not be made anew at the time of affected State appiove the emerpency plan as a preconds-heense renewal.
tion to construction; and (3) required that offsite erner-gency preparedness Imdmgs be made before any fuel As part of a separate rulemaking. the NRC is unJettaking loadmg or low power operations are permitted, a renene environmental study with the purpose of defin-tne the scope and focus of environmental ef fects that
.Dw (.omnuwon conMdered that these three petitions nYd to be considered in individual relicensing actions.
hm a (onunon Wnne, thus wananung Unmhancous e
a
- n. A idonally, the State of hiame fermally re.
An advance notice of proposed rulemaking (10 Cl R Part
- 51) was issued on July 23.1990 (551 R 29964k Also, a qu'ued that ]..the hiaine Petition be consohdated with notice of intent to prepare a generie environm' ental im-t os &d, Won Wudon, In &n9ng the peudons, U*
on concluded that its present terulations on pact statement (GlilS) on the effects of renewing the operating license of individual nuclear power plants was emnreng pwpan nq aw a&qu w to protect public issued (551 R 29967).'fhe draft and final Part 51 rules and heahh and saf ety. AdditionaUy, on h1 arch 20,1990, the Gl~lS are espected to be published in mid-1991 and in Man forwarded to the Comuussion for review a proposed mid-1992, respectn elv.
annymg paparednm ruktnaking package aladng to Part 52 licensing of nuclear power plants. A draft regula-tory guide that amphfies the emergeng plannmp require-5.-l.3.5 I:nwrgency Preparedness ments for fuel cycle and material licensees was pubhshed On l'ebruary 16,1940, a federal Regnter notice was pub.
lished by the Commission denymg three petitions for 3,,g,3 3 gj. kd lmplanude rulemaking concerning emergeng preparedness at nu-clear power plants. These petinons were submitted by the In 1936, the Commission published its Safet) Goal Policy Citizens Task l'oree of Chapel I hit, North Carolina: the Statement. On June 15, 1990, the Comnus%on directed 39 N U Rl!G-1266
k h.
s 4
5 k
k and Js s t hpc
- rq 5thm ain] repulatory prat tis ts 'le opetatof s ate pM &tl w ab, opct rmi r c an.]
minty
'y tin c. plins tus e hech ( stabhthed to to take ht ndis tal allons a ht n nt t Jol enJ. ne,.mpor-
. mph S ti s.
1 A s t L'p a h o m u na s h anun to cmur e that tutnic u rula.
tantly,ich.un lium A tions tlut can h.o c aJs cite (lin tt Itis I!ld ut ts ' s alc cs Mllatal It y (imhy mity with the the (onwquenccs i'i a scs ct e,R ciM 111 t an lk'll at t,ilh he
lt ty, mi W a rt ytt mt nts Whde the ( onumwon of tuhcantly f cduccd. Sune in uty an iJs ul inanaremcht h
it mm/o IL tt mh tatMi of the saf et) roal m awess-strateries do not mwlu urmhtant plant dcurn(hanycs, e rt ; ' bhin MW w dl mitulb Luy, the ('ommtwen substantial saf t ty benchts can bc quu kly at hu u d by co-so ld s c' as &ticJ rmJ mt e n do chipeJ and espe-turmg proper opt rator ;ations t hus the uutution of tie m e ram J. thn to uten w di he natunut auident maturrown! propiams at opt ratmp plants a a lorial result of the IPl prottu.
5.5 MCIO awl (lellt Illipleilleillillioll
'lhn prortam cicment prouJes for the unpkun ntation of the Commisuon\\ Sescre Accident Pohey Statement b.b.! blH!llllt'lti O! NrO!!!c!!!
UI d
I IC #
ID scatth directly to the tcgulatoly ptoccw hhheation of A 'n t ie, m sh nt m a nutlur po6er plant is an cvent m the ( onmuwon's ;ules or pohcies t cratdmr utiny, emet-h dany J and there n a potential for pericy phinmny, and wntainment deurn are cumples el whh h IId i of t tclc r e of I oic au monts of hwon proJutis. Sigmhcant arcas m which the esults of sn cr e an uh nt recan h may 10 't af t h lun lu. (li pi, t hu mni on the hLehhooJ, propics.
allect f uture thanyts-wm anJ w w 9t racs of a u;u re acudent as dncuued (af h r. Mm h (d (b. wi:f k has WoicntiatcJ on the pct-5.5.2 PrOgnitti Sintlegy Im m a t iltheti ntam:Pt ni dunnp a (ncte accident, l% lushfiC lb d( msd simlamnwnt Lulurc rncchannfus, and
" d C
!d (
b the.h hh d ib umt m ment to nutirate the conw grain sy steinatwally cununn inuphb rained irom sn et c qat ru cs H a < u n m a ni.
accident rescatch to identify contamment vulncrabihties and to identify potential unpunements to conect vul-in thc (ic on p *hg statement on sacre acckents UCl"hd'IIC' m nu :c.a p',t r pb'.h m MJ on Anrust 5, l%5 ($t)11(
llecause of concerns about Malk I wnt.unmchh, the (,Pl 2
3., I M.L tht tmm ' on winluJed that emtmp plants s
l piorram trutully stuthcd these contamment% llowesci, pmc no u a hm t o L to tbc puMN health and salcty and that studies of all types of mot.unments ate also m portest lf thetc n m m t Jute occJ hir p nene rulemakmr re-potentkal linpttactnents are identl[ led lhal plosiJe sir-(
lat(.J to ww e u W t.h f lown ct, based on Niu? anJ tubcant enhancements to saf ety and ate shown to be cost j
mdutti) esp !!t m e w tlh pkmt-Speci[ic probabihstlc tisk pmsuant to W t,i R R M tlus prorram wdl e
I aucunwms IPR Au the ( omnuwon n conunted of the j
necJ for a u urmanc cununation of e.a h emtmg plant to reconunend specib, e terulaton requucments.
l identih any l! mlynihc sulocrabihtin to sn ere acet-The UPi ptorran n tlowly r elatcJ and complement.uy to drnh l he puha statement mJwated the intent of the the ind:uJual plant cununations llPls) and accident
('onmuwn to take all reavnable steps to reduce the management programt The CPI ptortain cununes wn-s prohAhh of a sn cre aaiJent anJ should a severe acci-tainments for sulnetaNhties on a renene bau.s so that dent (E ur,12) millrate tis conwquetites to the extent gg};gjg ggggg g ggggg ggg g9 powhic. As part of the unplcmentation of the Commis-utxettam severe accident phenomena on an mdiuJual sioni Severe Aaident Pohn Statement, the staff has baus Ihe IPlh on the other hand, deals with plant spe.
reqmted mJaidaal pl ant cununations(IPl s)of all cust-cific contamment sulacraNhhes umque to a particular mg plants to iJemd) any plantapccihe s ulocrainhties t" plant that are not tt eated un.Jer the rencric cpi ptorram.
}
An ere acchh nts Moth of the work performed to implement the Severe liliS has been given the responsihihty for the iniplemen-Accident Poh0 htatement has locu ed on research mto tation of the IPl: I his nnplementation has mwhed de-phenomena that woulJ occur durmy snere accidents and velopment of ruidance for perfonnance of the IPli pic-methods to mtemancauy dnwver vulneraNhties for se.
paring a genene letter to plant operators requotmg the sete acctJenn ihn work has shown that the causes and IPli, anJ developing review plans and esentuaHy revww-wnsequensn of snere ace:Jents can be greatly mflu-mg the tesults of the IPl! subnuttah m cooperation with enecJ by nutlear power plant operators and that many NRl(. ~lhe requirement to correct any iJentihed plant-sulnciaNhties to sncre acc dents can potentully be spectfic vulneraNhties not voluntardy conected wdl be chnunated by proper operator actions. The TMI-2 acci-detertumed by the backfit rule. Accident management is dent and other abnonnal occurrentes in nuclear power not required as part of the IPli process but was lugh-pl mis h-ne shown that operators do not stand idle but hghted an the IPli renene letter as a future requirement NlHi ti IM 40
that wdl make use of the results of the IPli process.
issuance of NUlt! G-1335 formally started the lPliproc-Severe accident vulnerabihties due to external hazards ess. Utilities will have 3 years (until September 1,1992)to (e.g., carthquakes, fhvJs, fires) are being consider ed un-complete and submit their IPI:s to the NitC.
der the IPli for 14ternal !! vents (IPlittii) program.
Major e fforts on the IPl:s in i Y 1990 have involved des el.
On May 25.1988, the staff presented to die Commission oping detailed review plans and sofming estimates of re-an
- Integration Plan for Closure of Seveie Accident 1s-sources required for the review. 'ihe staf f has fortaed a sues." St CWSS-147. 'this plan discusses the relation-team to review early IPl! submittals and initiated the ships among the major elements of the plan, which in-procurement process to obtain contractual assistance for clude (1) CPI propram, (2) IPlis, (3) cxternal hazards, (4) the !Pli reviews, acciJent management (5) improved plant operations, and (6) setete accident research program (see Sections
'ihe Commission.ts curr ently developing plans to evaluate 3.1 and 3.2'L The commission paper aho discusses the the IPli results to assess the mdustry status relative to the relationship with related elements such as safety goals, Comminm's safety goal poh Mans for capture, re-severe accident poho for future plants, and generic safety trieval, and use of plant specific data from the IPlis have issues as wv!! as requirements for closure of severe acci.
been developed.
dent issues.'llus pr ogram element includes work to exam.
ine the areas of siting, emergency planning, and generic Initial plans for information to be collected and main-tained will be tested using the Yankee llowe (Massachu-safety issues for potential tesolution of issues or changes to exo. ting regulations as a result of severe accident re-IPI and updated as required. liasicaHy, the infor.
mation will be depostled into two data bases that are searchft Jmps.
ntly being developed by llrookhaven National 1 abo-ratoiy.'the first will contain information related to plant 5,5 3 llesearch Accomplishments in IT 1990 systems and their relationship to cach other. it is expected that this data base will be very t sefulin obtaining generic 5.5.3.1 Indhidual Plant ih aminations insights and a better understanding of plant characteris-On November 23,19hS, the NitC issued Generic Letter tics under various initiating events.
8840, "Indnidual Plant lixamination for Severe Acci.
.the second data base will contain IPi! results and find.
dent Vulnerabihtics-10 (l it 50.54(f), to all licensees q
of nuc! car power reactor facilities. ihis letter requested ings, including items such as core damage frequency, dorninant sentributors, and the licensee's identified vul-that all heensees perform a plant exammation that looks for vulnerabihties to severe accidents and cost effective nerabihtles and improvements. In addition, analytic tech-niques used in the IPli analysis to evaluate important safety impros ements that reduce or climinate the impor!
areas such as human factors and common.cause failure
- tant s ulnerabthhes. the specific objectives for these indt-will abo be included. It is expected that both data bases vidual plant examinations (ll lis)are for each utihty to(l) des clop an m er all appreciation of severe accident behav-w 11 be used together to gain a maximum understandingof ior: (2) undentand the most likely severe accident se-plant specific behavior.
quences that could occur at its plant; (3) gain a more l'o date, twc IPl! submittals have been received and are quantitatise understandmg of the overall probability of tmder revie v by the staff. 'lhese are the Yankee llowe core damage and radioactive material releases; and (4)if (Massachusetts) and the Millstone Unit 3 (Connecticut) necessary, reduce the overall probabi!ity of core damage submittals. the stalf reviewof the Yankee Itowe submit-and radioactive material release by appropnate modifica*
tid included a meeting with the licensee in May 1990 with tions to ptocedures and hardware that would help prevent a set of qu tstions transmitted to Yankee Itowe in July or mitigate severe accidents. Upon completion of the 1990. Curn ntly, the staff is awaiting a tesponse to the examination, t he utthty will be required to submit a report questions.T ic Millstone Unit 3 submittal was received in to the NitC describing the results and conclusions of the late Septeml et 1990 and staff review has just begun.
exammation. ~lhis submittal will be reviewed and evalu-ated by the NitC.
5.5.3.2 Nunal P.wnts
'the NitC also issued NUlt!!G-1335. " Individual Plant in December 1987, the NitC established an lixternal lixammation: Submittal GuiJance," as a draft for com-livent Steering Group (liliSG) to make recommenda-ment in January 198u to provide guidance on the conduct tions concerning the individual plant examinations for of the IPlis. A workshop was held on 1 chruary 28 and vulnerabilities to severe accidents initiated by external March 1 and 2,1989, in 1 t. Worth, Texas, for utilities and events (e.g., earthquakes, floods, fires). Itecommenda-interested members of the public to address comments tions were needed relative to: (1) w hat external events and questions on the IPl! process and the guidance doeui need consideration in the IPil, (2) w hat methods can be ment. NUlt!!G-1335 was revised to reflect comments used in the examination, and (3) coordination of IPl! for received and issued in fmal form in August 1989. The lixternal livents (IPlilili) with other ongoing regulatory 41 NUltliG-1266
'suisuup iiiusiisiii
ils tl\\ itles Irl\\( n lIl[ esti r llel es ents. llilk [U sir An k le dke, M cII (luk i ! It l a Inc
- kbpiNkb die Teb nx.a ea med. with the tem onJer of the reiwn n beduled hir iMtLince 11) April {Wl.
11 n e tubuonnttees were t ttahinheJ m Aprd IN to ide reumonenJ osons m the alcas of (1) sentmc, C) g g g g g gg g gipg g gjg g ;gg g g p;gjgjg hrt s nd W larh wmJs flooJs, and others (e p., man-t1tje h t/,lrds sut h as nearby tt;tilq1o!tation. rnilltary, arid I
mJustri d f aihucu l)unny 19s9, the three subuimrmt-I tecs temph t.J ihnr stuJics and maJe resornmendabons 5.6.1 Statement ofl'roblein f or the ll'l ! l 10 the i I.M L the NI(C must prosiJe radiation protection standards in May 19%. the st.dl ivinpit ted wor k on a draf t renerie anJ ruidance that ernures that workers and members of
!rtici and draf t ruidarne da una nt (NUlti (el407) to the reneral pubbe are aJequately protected from the be sent to hcenset s, wla< h devnbes the scope, accept-adscise consequences of esposure to iom/mg radiation ald e ine t h t hj s anJ ttportLy> requircinents for the hm beensed activities. I(! S actinties needed to support g
g g g. -
rddum protection ll'1 l l 'lhe drait dskuinents w ere 1%aeJ for pubhccom' rut nt on July 2'.1000. In September 1990 the Mall con-mWmN hhpy pWhm M unHemermy thesc j
dus teJ a wot kshop on the draf t rencric letter and on standards; and plannmp, devdopmp, and du cetmp safety N t 'l<l t i IMO to sohut wnunents and answ er questions
- tm e & dou m m-m for liceW uncernmr thcu contt nt. Approsunately 210 representa-ing dccisions, inspection and enforcement ac'twities and k
nde dM*pm pu Cl his incluJes analyn j
un s hiun mJuttn. State agencies and the pubhc at.
dMe memk evi&m N evaluMe the relatm-l ten &d the workthop ()) nyst concern to heemees was I
n @ m WMdW udiahon and l
the co.t (d domy the Ipl l 1: and the requef ted.b) car g
g g gg,
W Mh w htdule for wmp!ctum. lhe staff a currently revamp d ch mdmmic kM dfects, intluding the the rener w it uc t and N t !!(1 (r IM)7 to meor porate clan-M M to Mn M k pbbn d nuned lwat on.md t hant s a sultmg hom itedbac k received at tk pba%.l het.c an dvses are used to provide bees f or of inaemed incidence of cancer and re-the wor kshop and espn ts to hac proposed hnal docu-g
.g ments to the ( onummon for reuew m i ebruary In91.
ihm emdexe m@is Abdiuk rid 1
Alter wmplenon of the hn.d 1pl'l F renene letter and assessment (Pil At the desclopment of safety roab. and Nt'lil.(i-lW the staf f wdl dnelop a renew plan for emerrency plans the identdieahon of radiation protec-the Ipi 1 1 subnuttah. lt n espet ted that the approat h for tion problems, the a! location of prionues for regulatory esico of the Ipil I w diloihm dosely that des cloped for action, and enuronmental impact awessments, llecom-reuew of the mternabewnt IpH subnuttals-mendatiom of such organizations as the International Conunisuon on 1(adiolopea! Protection (IUllP) and the National Council on 1(aJiation 190tection and Measure.
5.5.3.3 Containment Pc formance imprmements ments (NCitP) presidential ruidance to l'ederal apen.
cies. consensus standards, licensee performance indiea-All majot tiemems of the Cont unment Performance tors, cost and feasibility data, and available techmeal Impmvement (Uph piorram hme been completed. Ge-information also proviJe bases for developing regulatory nene letter s (Gl M h<n e been iuued to licensees startmg and technical documents related to radiation protretion the plant-specihc batkht of the hardcned vent for all for wor kers and the pubhe.
HWit Mark I cont,unments (GlM 16. dated September 1, luxu) and requestmp that mht t impros ements be con-lif fective r crulation associated with environmental policy siJered m the li'l (Supplement 1, dated August 29,19S9, and decomnuwioning activities invohes the tok of plan-to Gl. n -20 for hwR Mark I containments and Supple-ning. des cloping, and iwuing appropriate regulatory posi-ment 3. dated J ul) 0,19W. to GI M20 lor other contain-tions. Using infoimation rencrated internally or thiourh ment ty pen 'Ihe on)) remaining actmty under thk pro-narrowly directed rescatch, new positions are developed fram a to complete and iuue for mformation a senes of or existmg poutions are modified. These poutions can Nt!RI (i/('ll technwal reports to document the analyses take 'he form of reptdatory requirement % policy sta'c-and evaluatiom done by the staff and its contractors in menti puidance, or criteria for activities pertaining to awessmg the sarious containment types. These reports decont iminating and decomtninioning licensed nuclear aJJress the potential vulnerabihties identified (charac-facditi :s, cxc mption of mateiials or products from r egula-ten /atWn reportst the potential fnes esaluated (en-tory coatrol, and disposing of low-level radioactive waste hancement reports), and analyses of the ef fects of uncer-streams Setting priorities for regulatory needs or tampes (parametries reportu it n expected that these deficien;ies are undertaken to ensure that the problems reports wdt proviJe beensees with information they may of greatest ugnificance to the pubhc health and safety or find usef ul m awessmp their plants as part of the IpH To the co'nmon defeme and security are addressed in an N t 'It F( i-1266 42
expeditious manner through properly defmed regulatory quires NI(C to establish standards and procedures for and supporting research programt expedited action on below regulatory concern (Hl(C) waste disposal petitisms, l'ederal agencies, includmg 5,6 3 I,rogram Strategy NI(C. are currently in the process of establishing and tmplementing intC or esemption levels for radioactive
'ihe Comtnission's r eputatory process requires that safety waste disposal as well as other areas.'the strategy of this enhancements to rules and guidance be systematicall'y pnigram is to carry out ( ommission directives to develop sercened to ensuie that there is substantia 1 increase in and implement ConunMon UK policy, lhts policy
~
public protection and that based on analysis the costs are w ould serve as the framework for specific exernplion deci.
justified. Itealistic values of the dollar per person tern sions invohing disposal of low level waste str cams, as well cnterion are needed for analysis to justify changes, but as other acuuucs concenung We release or use of radio-ase matenal, technology paps in knowledge associated with radiation health cif ects cause uncertainties in these analyses. 'the strategies of this program are to idenufy and compensate 5.6.3 Itescarcli Accomplisliments in IT 1990 for uncertainties in radiation rnk cocificients used for health effect estimates in pl( As and regulatory decisions, y,.3.1 Itadidion Pmintion issues (A feasibihty study im whether cellular and molecular cffccts data can reduce the range of uncer tainty in health AlslRA Center. 'the llrookhaven National latmratory nsk effects is nearly completed.)
(llNI.) Al.Al(A Center, funded by the N1(C. continued its wor k on surveillance of dol! and industry dose redue-When the Commission approved the whole body tion and AI Al( A rcsearch. llNI,has published a scrics of dosimetry accreditation rule, they directed the NI(C stalf eports (NUlt!!G/Cil-3469) that abstracts 2$2 national to estond the rulemakmg to include extremity dosimetry, and international publications discussing dose reduction
'therefore, the strategies of this program are to (1)im.
in arcas such as plant chemistry. stress corrosion cracking.
prove regulatory performance for radiation protection by steam generator repair and replacernent, robotics, and estabhshmg measurement performance criteria and ac-decontamination. In 1990 llN1, focused on high dose crnhtation programs in the areas of extremity dosirnetry, worker groups and developing an international dose re.
bioassay, and air sampling;(2) investigate effective new duction data base.
measurement techniques f or these areas:(3) establish the data base required for regulations; and (4) monitor spe-
'lhe center is recognited by the nuclear industry and cific indicators to detect improving and declining licensee others as a major source of information on new and cffec-performance, tive dose reduction techniques, and its publications are standard references for AI Al(A planning.'the llNI staff l'ederal guidance was approved by the President on occu-is available through the center to the entire N1(C organi-pational radiation protection. As a result of this newguid-ration and its licensees for information and advice on all ance, NI(C regulations and regulatory guides will have to aspects of radiation protection and dose reduction.This be revised. *Ihe strategies of this program are to (1) mod-effort becomes even more important with the implemen-ify radution protection guidance and standards to be con-tation of the new Part 20, which makes AL Al(A a require-sistent with Presidential puidance on radiation protection
- ment, requirements, and (2) cor"inue to monitor licensee per-formance indicators by using the 1(adiation IIxposure in-In 1990, the llNI. Al Al(A Center published NUl(liG/
formation 1(eporting System (Illill(S) program.
CP41100, a report on an international workshop held at ilNI. on new developments in occupational dose control
'the NI(C needs to develop a regulatory appmach to on AI All A implementation at nuclear pow er plants and evaluate futute requests involving decommissionings and similar facilities, license terminations. This regulatory approach should define acceptable alternatives, requirements, and criteria -
Low trerl liquid Effuent Recirculation Study. The pur-for decommissioning before such a request is received.
pose of this study is to determine the potential for nuclear The strategy has two parts: (1) to develop or modify power plant recycling of low level liquid effluents re-regulatory requirements and guidance to prot ;ct workers leased to a coohng water source and the feasibility of and t he public from radiation risks associated with opera-modeling such regeling. Modeling of potential regeling tions involving the decommissioning of liccased nuclear would be used in site evaluation studies and in offsite facilities and (2) to establish radiological cri* cria for r esid-radiation dose calculations. Such calculations are impor-ual radioactivity.
tant during the plant design stage when there are neither actual plant effluents to measure nor environmental sam-In the area of below regu!aiorv.oncern the Inw-1 evel plcs to collect that may have been impacted by recycled llad oactive Waste pohey Amendments Act of 1985 re-effluents.
43 NUl(l!G-1266
. h or daunun um/ lolog ell'a wa l t uvavy ha e s wr3.
measurements. I he r esulh of this work indicated that the An onromp propam that requocs accreditation of use of such a phantom rmrht be more meful foi internal pe sonnel w hole bod,i dodno,try pioceswis became el-dosimetr) calculations. ()ther priorit es precluded fur-letIne in i t bruary 19 W Acsrcthtation is acquir ed Iher f unding of worL m this arca at Ihis time.
through the Nationai Voluntary 1.aboratory Au redita-tion Proriam (NVI AP) operated bs the National lasti.
Sr/bPonard Garnin+ Ray Drtn tor. Research under an tute of Standards and 'leshnoHry (NISTh and reae.
SillR contract to desclop a self powered ramma ray de-ued:tation of proccwon is requue'd cury 2 scars. 'the tector (SP(iRl". continuing, using a concept simdar to l
pregram roal is to unprose and m.untain quahty awor.
that for selfy
_ red neutron detectors (first developed afne anJ quahn control over all aspests of personn(I m the Sosiet Union in 1961 and improved upon and pat-doumttry pro.chmg by requirmy all raiscssors to meet ented m Canada m 1968). Two prototype models have the performante icymr'ernents of the national contensus been designed and cons.tructed, one for isotropic re.
standard for pnicessmg. ANSI N13.11-19N3.
sponse and the other ior duettional response. liield test-ing of the prototype detectors is expected to start in IT As of March 1000, 62 laboratones were accredited for 1991. If successful, the device is expected to be used by processmp whole body doumeters. These mclude com-radiation workers to rninimi/c exposures to radioacuve mercial doumctry pusessors, inihtary establishments, paruma ray sources.
wmmetaal sinpbudders. nuclear power wmpames, and f
other commcicial estabbshments that use radiation
.hu,luo aknt hutmolumirmmnt DmimrIns. Re.
measurement techmques. A draft reputatorY pmJe that wanh unda an SillR contract to develop a gamma. ray I
wdl discuu methals of meeung the NVI.Al procedures spectnunetn/dounwter hai. begun. Dw purpose is to i
for proccuor accreditauon will be pubhshed for wmruent demonstrate the feaubihty of derchiping a differential
~
m carly FY Pol.
enetry absorption spectrometer coupled to a small rniero-computer that would have essentially the same response in the extremitt dosimettwaea, a seconJ set of perform, to radiation as that of human tissue over the energy range ance tests waicomt lcted anJ the results published in of 0.5-10 MeV. Current dosimeters are (ssentially flat i
Apnl 1090(NURI:G/CR 4540). Preparations ate under over this tante w hile tissue response varies by a factor of way to tmtiate a third set of tests against the reviscJ around eight.
standard OlPS l' N 133h which is espected to reach j
pubhcation arounJ December 19no. Should the tes;s mdi-
.. Dot n 5 on 1% Detutot ne rapid detection, cate that the revised stanJard is a suitable criterion for neutunent, and location of small particulate radioac-l testum, appropnate rulemalmp wdl be imtiated to re-maknal on laun mMan )pmkcudoWng h I
quac 'esticmity doumeters to be processed by procewors wo of won un an IR mntract. Under the I haw H of s mntract, a protoype of a pm for surveying clo@ thing wdl be deseloped and demonstrat cerufied under the NVI.AP procedures m u'se at NISI.
New %n DoAr Compurrr Code. A new computer coJe for m pm has poknual tot reducing radiation exposure calculaung dose to the skm from radioactive materials on of personnel who may wear " clean" protectisc clothing the sLm is being descloped to replace the VARSKIN unaware that the clothing bears particulate radioactive i
code in use since 14% The new code wdl be a prcat deal
- " k "" h mor e Ocuble than VARSKIN, allowmg for self-absorption of rathanon within radioactwe particles on the M
Heahh meets Rescanh j
skm and backscattenng of radiation, and will pernut the Embryo //rtal Do3e from Afalernal Infalc. A study to im-calculanon of dose frorn different shapes of particles and prove our ur.Jerstanding of the contribution of maternal particles s"parated from the skin by clothing. The code radionuchde burdens to prenatal radiation exposure was wdl also calculate the dose from both ramma and beta continued in IT 1990 with significant progress made. A radutions.
preliminary report on methodology for calculating
/]Jrcts of IVor4cr/Do3/mrter Geometry on Daar Afrasure-embryo / fetus dose due to maternal radionuelide burden ments. The feaubihty phase of a Small lluiness Innma-has been completed.This information is needed to assess tion llesearch (SBill) contract whose objecove was to consequences of acciJental releases of radionuclides and determine the quantitatise clicets of worker / dosimeter also to ensure compliance with the proposed 10 CI R Part 20.
geometry on dose measurements by simulation and ex.
periment was completedflhe firm apphed fractal geome-Improcement of Health f/Jects Afodels. In IT 1990, a re-try as part of its innovative approach to the investipation scarch project was initiated to develop modification in of tius penene problem.The contractor des cloped a com.
heahh effects models used by the NRC that are being puter simulation phantom to model a worker in a radia-required in view of recently pubbshed reports on the tion environment in an attempt to quantify the effects of eff ects of low-level ioniting radiation.The possibility that vanous worker! dosimeter peometnes on external dose internal exposure to alpha emitting radionuclides might 1
l NURIG 1266 44 i
oa ut alor r with n posure to hm l.1 I raJution na>Jthea.
(enscJ facihty. Such mformation n rma m.unt.uned for l
hon lot unluwn el alpha raJution wiu be deseloped.
some 575.000 persont mmt of whom wod ut at nudcar power plants. the 6omputett/auun of these data enables P.u l 1 of Re m on I to Nt'RI G ( lt 4214. "licalth the N1(C staf f to respond qunkly to requests for mJaiJ-1 !!nis \\1 ode i s for Nuslear Powet plant Accident ual exposure histones and to analy/c the data for tit ndt
('oineque in e Analy sis" we pubbshed m IT 1940 af ter the data also assnt m the eununation of the doses in.
Pait 11 was pulhhtJ m lY 14W to determme whether curred by transient workers as they mme from plant to the niemne data hec on u llular anJ motecular ef fects plant. I or eurnple, further analysn of the data repotted un be used to reduce the uncert. unties m health risk for b6300 persons terminatmp employmer:t durmy 1%7
( tunates f or loa dm.e and dme rates.
rescaled that N,7(H) of thern had worked at two or rnore nuelcar power fanhtics arid that none of them haJ re-( 7a rm of loui th of l'ramam Huduondr ('om;'arn/ to ceived doses in cu ess of the repulatory hmits as a r esult el Ru! tarn m luo f \\l RI 6-1.N1) Thn drall stall report their inultiple emphiy ment.
uunp.ned the thenmal touaty of uranium heufluoriJe with t!n acute dietts of a raJuhon dose of 25 cim to the Rrumin yl' art M Radwnon Na n!ards lhe Uemonssion w tade body (the ulue used m Part 100 deahng with reac-appuned usmg a complete reusion to the NRC teruhi-tor siung mter ut lhe work was done in support of a mun for udution mitechon m 10 C1 R Part 20 't hn hcenunr.nuon lot a unmnercul uramum entnhment revnion updates the Commiwon\\ reputations to incor-plant l he draf t repor t we pubbshed hu pubhe conunent porate recommendations made by the Interfuhonal m Aprd PN A hnal rt port a schedulcJ for pubheation Conunnsion on Radiolorical Protection, the Nanonal
'" U
- Couned on Radution Protection and Measurements,and the reused l'ederalltidiatam Guidance for Occupational SW lh ub,pmint of Itulo and linulatory Guido I Apmute issued m 1%7. 'ihe new standards represent a (b u;u;7n il ; m /h;ta krr"n in 1%4, the Atomic sigruheant (hange f rom the methods pr cuously crnphud t
m nhmuol uJubon doses.'l he new Pari 20'w di I netry ( onumv mh ht can requamp tert.un licensees to meWdm kmd cuptud dose f rom a subnut nj or ts on oa upanonal raJunon desc recen ed by wo!Left lk e data are collected anJ eomputen/cd m an posuble 17 tems(Fretn! quarter esternal + 5. rem annual g
NH( ystent called the Radutlen lispii ure Itilotmation internal) to a total effectne dose of 5 rems per year. The gg ggg g
,gg Relwtmp Sysu m tRI IRM ihe sptem prouJes a per-j manent ru oid of the data anJ peinuts expeddious analy-from an unpheit 0.5 rem per scar in the present rule to an
~
ses el ihe t a o L nJs of ieports requaeJ tannual statntical hm vh of 0.1 tem per year. The new Part 2tl wn.
j sunan an s anJ mJ:uJual ternananon reports) l.spo-mdices that pn e the :adionuthJe concentiation um j
sures retened as a re uh of mcJteal proceJutes are not lurnts for air, water, and sewerare.
reqmred to be tcported Proposed Rule un I arge Irradiators Pubheation of a pro-s A pu hnunan wmpdahon of summanes of the annual posed rulc on tarre u raJutors was apprm ed by the Com.
stahsucal reports lor luo revealed that about 230.000 rmion. large u radutors are defmed a; thote capable of j
pe r son s weie momtorcJ of whom about 50 percent re, dein ering a dosc of 500 raJs an hour to a person standmp tened measutaNe doset The workets received a collee-1 meter f rom the sources. Pubhcation of the proposed tn e dme el appmsunately 44.000 per son-rems or an aver.
rule for publiccomment n npcetedin early IT 1941.The are annual dose of about 0.4 r em per uor ker among those final rule is scheduled f or pubheation m late IT 1991.
I recen mp a memm able dose. These figures are about the same as thme lounJ for 1%7. Of the persom monitored, Im;4toungBmancy
- a3nrements As patt of a program to i 40 percent worked m nuclear power p' ants, anJ they imprme health physics measurements, sescral c! forts m meurteJ about 40 percent of the total annual collectne the area of bioassi) were imtiated or contmueJ. An inter-dose. Af ter det hmog f or ses eral y ears, the annual collee-ageng apreement wuh the Nahonal Instaute of Stan-tne dme meutred k nuclear power plant workers ap-dards and Technology (NIS I') was established to investi-pears ta h.n e les eled olf. I rehminary compdatiom of the gate the standardi/auon of phantoms used in Ihe exposur e data r eportcJ by nuclear power plants for calen-eahSration of m neo countmp eqtapment. An interna-dat scar 19W mJicate that the collectac dose decreased tional workshop held at NISI in the spring of 1440 pro-to about 37.100 person-remy esen though six new plants vided guidance for f uture actions. A contract was contin-reported. this n aluut 10 percent less than the 40,S00 ued with Pacific Northwest laboratories to evaluate the penon rems reported for inn trial use of a performance standard for bioassay and to determme if the standard needs to be reused or if aJdi-A second kmd of nimsure report requircJ of ecrtain tional testing of the standard is oceded.1 fforts to coordo NRC heensecs piouJes idenufwahon and dme data each nate actinties in these areas with the llepartment of tune a momtored mJniJual ternunates work at the b-I nergy are also bemp contmued.
45 NURI G 1266
Mty la p:ronmnfm /C e / /Li.!:: No;J z 1 4!:. T " ent Union of Amerca (l'WU Ai nuJeat conf erence.1*resen.
'l he im d r ule on tim wN < t u puhh' V d m imu ay utions wu c nuJe on the tot partu ' poNt m. hunun DN W l 11 W i lia r ule m,orp.mh s by it lerens e the f actors the preposed new taJution protec tion regula-
!cquirenletlls (d the /TNsl staqfird ici fjdhier aph) de-thins lht' **l5cloW t Cf ula tort (('ncei n" inue. Iow-Icte!
m s plus a riurdu r el o% r A ty h oures, m !uJmp waste dnposal, the NI(U:Occupanonal Safety md licalth auumune source !M, r on n u. tu. orpnn ed sourt c.
AJmmnttation (Ohil A) workmp relationdups, the NI(U to-able u macetm s,md aJJiumul 1 die and u ar rnngs enf orcernent progrant and eur t ent htnew for duty nours lhe rule.d a reqmu s alJiaonal it porung of speufwd mudents anJ reqmics raJhyraphen and radiographer Particular mterest was exprewed with respect to NIIC/
auntants to wear alarmmr dos nu tris.
umon meetmps the avadabihty of NltC reports workmg hours; drup testing; the impact of regulatory puides, pol-Con %tm elIm!a3tna! h!n vjp A hnal rule that icy staternents and generic letters on workers; tranwent would temnvc a tturJ [w ty wr tihcanon propr.un of the worker dom wmkn dose and du huoN hnutanons on Amencan soach f or NonJeuru;uve I niinr( ASN lihas OSII A mspector access to nuclear power plants; OSilA been dt:dicJ ard is expecicJ to be puhinhed m the In!-
and NW requirements related to worker safety; and the cra!!Lmm m l b emht r PMt !!m rule would gn c lwen.
effects of enforcernent on mdiviJual workers. In adJiuon sees the ophon o! uo the A$N i proyram m heu of to DWU A workers. repruentaines of the Od,( hemical deunb:nr their tra:mer prerram to N1W lhe ret uhea-and Anunic Workus Umon and the Intemahonal H oth-non propiam n opreted to improu both traimnp and edniod of 1:lectrical Workers were present.
8 sdt ty per!omunce m the umLpac.
4 l
M.14 I:rnironmental Policy atal liceornmiuloning l
//or IWtu le /G:./ zen / /% 9 hrk conhnue'. on dehamp 7 m by scry small radm" jgf, g.gfyn. (3,,y (fgfg pgfyy gfy,ygg in 3 l
the c!!cets of irraWauon of 11 k
j ine parudes i, hot p otidt s t 1 he tecommendation of gn heh pohd action, the ('otnmnsion pubbshed in the l
the Nanonal( otmed on ILJunon l'rotunon and Meau RhM' -
fM PolMtm ement on Below ilegul+
GmEern on Julv 3,19h0. The pokes statement pio-o i
tnen;ents tM Rilil t P oinontanu J m its lu pott No lub."i mmt dn a consistent, risk lused framework' for deteunining f,or I alw m e io o
. sL o;the SLm. lonned the h.nn hy a nea ( omnwon erJorcement puhey on espo-whether a practice can be exempted honi som( or all of A
ubton controls usually imposed by the Commis-sures hem hot pattklet thJet an arreement with the N( RP, another repmt dealme a ah the elfu ts of uradu-son's r egulationCthe poliev wdl be apphjable to esemp-uon of the sLm l y parthles near. lut not on, the skm(e p.'
hon & mions in a number oIareas meludmg the dnposal of v low achvity waste, the accontaminatian anJ de-on hair or clothuno and parudes m the eye anJ on the eatdrum udl be descloped Revanh into ways of num.
comtmssionmg of facihties, the approval of consumer minny the producuon anJ drpend o! partides minpat-proJaets and the potential for recycle and reuse of mate-my sus h produ;uon and deposal and interpreung and d anJ cpipmen apply my nongos einment r ewarch tesu!b Io the not parti ^
l tie problem o unJer way at hrooLLo en Nanonall abora-1)urum late August anJ September UtntL a series of pub-I tory.
lie meetings was held to esphun the HRC policy and proviJe an opportunity for public comment and state-ments. 'the meetings were held m Chicago, Illinois; Kmg
<br.%;m;! mein rhe ll'od;las e. Wt k ber.m on a reputatory of Prussia, Pennsylvania; Atlanta, Georgia; Arhngton, guide on air samphny m the workp! ace to meet the te' quirements of the new Part A the putde wdl deal with lesas; and Oakland, Cahfornia. An estimated 440 per-nsues such as a hat should the hcensee do to demonstrate sons attended the public meetings. The makeup of the i
that samples are representatne of the,or mbalcJ by audiences ddfered somewhat from meeting to meetmp t
workers and w hat measurements are necesstry to be able but was largely composed of representatives of pubbe toadjust dern cJ air conientrations toaccount for particle interest or environmental groups. concerned citi/ ens and sue. 't he ymde w di be accompamed by a technical manual Federal, State, or local elected repiesentatives desenhing how the recommendations m the ymde can be in the area of low-level waste dis}wd and exemptions for met. lhe draf t guide and technical manual are scheduled to be published for pubhe comment in the summer of eettain types of wastes, the staff evaluated the public 1441.
conunents from a proposed rulernaking that would allow onsite inemeration of waste oil generated at nuclear power plants. 'the final rulemakmp package will be con-Coon!manon wai WorAcr Goups 'lhe rootdmation and sidered by the Commission in early IT 1991. Also unJer t
mformanon exchange rifort willi unions tepresenting consideration were petitions from Rockefeller Unisenaly unhty workers was wntmucJ. Sescral members of the for exemption of certain biomedical research wastes con-NRC staf f f rom three NRC of f acs conducted a lecture taining small quantities of tntium. carbon-14, and certain vnd discauton program at the I Y 1940 Utihty Workers other isotopes 'lhe NRC staff is eurtently detenmmng if NURIElM6 46
thete petioons shoulJ be treated as a single practice un-Decomminioning Actieitics Development of information det the new ll11C pohey statement.
on the safety, costs, and wastes related to the decommis-stoning of 1. Wits and other nuclear facdities has contin-ued. Data are being developed for reports of the decom-Rcm!ual Contamination Criteria in l'ebr uary 1940.
missioning of 1he I a Crosse (Wisconsin) and the llancho NUltiMWit-5512 was pubhshed for cornment to pro.
Seco (California) nuclear power plants and the Ship-vide the technical modchng basis for the establishrnent of pingport reactor. ihese data will cover costs, radiation residual contamination cnteru for surface soils and for doses, and low lesel waste resulting from decommission-stiuctures. 'Ihe Imal report will becorne the basis for mg actiuties. Identification of radionuclide source terms mtenm resiJual contananation coteria that should be for decommissioning and low level waste burial is con-available in 1991. Separately, during 1 Y 1990, the NitC tinuing, using materals from the Shippmpport reactor, a staf f imtiated a contract to develop crocru for sun eying decommissioned West German reactor, and several U.S.
sites to determine cornpliance with residual contamina-operating reactors.
tion enter u.
llegulatory guides for 1 Wits and other nuclear facilities were issued as final guides related to the assurance of l>uring I Y IWO, the NitC staf f imtuted efforts to de-funds for decommissioning financing. The final regula-selop a renene ennronmental impact staternent as the tory guide for decommissioning recordkeepmg has been first step in the preparation of rulemaking for dose crite-delayed awatting ongoing rulemaking activities in this tu f or Occornmissioning. This activ.ty will be contmued area. The fmal regulatory guide on standard format and for the nest sescral years. culmmating in a rulemaking content of plans for reactor decommissioning as well as and criteria for decontanunatmg and deconuniwioning of the draft regulatory guide on I. Wit methods for facil-nuclear Iacilines.
itating decommissioning are in progrew.
47 NUlt!!G-1266
Al'I'ENDIX FY 1990 ItEGULATOltY l'ItODUCTS FROM Tile OFFICE OF NUCLEAlt itEGULATOllY llESEAllCll I) ate llegulatory Product Description 6
Integrity of Reactor Components July 1990 Regulatory Guide 1.35, Provided guidance on insenice inspection of Reusion 3 ungrouted tendons in prestressed concrete containments.
July 1990 llegulatory Guide 1.35.1 Provided guidance for deterrmning prestressing forces for insenice inspection of prestressed i
concrete containments.
July 1990 NURiiG-1377, llevision 1 Provided a listing and summaries of reports issued i
l through May 1990 on the NRC research program on plant aging.
i Presenting Damage to Reactor Cores Api t! 1990 Supplement 2 to Provided industry with candidate accident manage-l Generic l etter 88-20 ment strategies for use in preparing individual plant examinations.
Reactor Containment Performance August 1990 NURIIG-1420 Provided peer review committee comments and recommended uses of methods, data, and results set forth in NUREG-ll50," Severe Accident Risks: An Assessment for Five U.S Nuclear Power Plants."
l Assessing Safety of Nuclear Waste Disposal i
l'ebruary 1990 Proposed Rule NRC licens;ng authority for the custody and long term I
l care of reclaimed or closed uranium or thorium mill tailings sites after remedial action or closure has been completed (55 FR 3970). (Final rule was issued in j
October 1990 (55 FR 45591).)
i June 1990 Petition Denial Denial of petition for rulemaking from the Sierra Club pertaining to uranium mill tailings sites to require an NRC license for the possession of material being cleaned up under Title I of the Uranium Mill Tailings Resource Conservation Act (55 FR 25670).
September 1990 Draft Regulatory Guide Provided format and content guidance for high level waste geologic repository license application.
Resohing Safety issues and Developing Regulations October 1989 -
Generic Safety Issues For generic safety issues resolved in FY 1990 September 1990 see Table 5.2.
A-1 NUREG-1266
Appelidi\\ (Coillilitied) l) ale llegulatory Product llescliplinn October 1959 l' mal itule
'lhe NitC crula, ion (10 ('l 11 Pait 3$) on paladium-103 for m'erstitial ticatment of cancer was revised in response to a petition for rulemalmp.
October 1989 NUltlG1370 lleported on h>dioren control measures and ofIcets of hydrogen bmn on safety equipment.
O,tober 1459 Genette 1.ctter 69 22 Provided resolution f or G1-103 on inaumum probable precipitation.
t Nmember 1459 Proposed Itute The proposed rule would amend NitC regulations (10 Cl It Part 34L "ASN I' Certification of industrial Itadiographers," which retornites the radiortaphie certiheation prograin and encourares beensees to use the prograin bemy developed by the Arnerican Society for Nondestructis c 'l esting ( ASNT).
1)cecmber 1939 NUlti G-1316 Piovided technical hndmps and regulatory anal) sis for G1-70 on evaluating power operated rehef valve and block valve schabihty m PWits.
l>ccember 1989 NUlti!G-1326 Provided terulatory anal) sis f or resolving Gl-94 on adJitionallow temperature oserpiessure protectioii for 1, Wits.
1)cecmber 1%4 Policy Statement issued a revised l'ohey Statetuent on Maintenance of Nucicar Poact Plants to state the Cornmission's expectations in inaintenance and to indicate the Commission's intention to hold rulemaking in l
abeyance for an 1S month period while it monitored j
iridustry initiatives and propress in maintenance.
l)uring this period, the Commission indicated that it l
would continue to develop a maintenance rule.
January 1990 Proposed Itule!!)ralt The proposed rule would amend NRC reputations, llegulatory Guide 10 Cl It Part 35, and an accompanying draf t regula-tory ruide was issued. 'ihe rule and regulatory guide would require medical use licensees to implement quality assurance (QA) programs and would revise inisadministration reporting requirements. 'the impact and effectiveness of the rule will be evaluated during a pilot study involving approximately 70 medical use licensees. The proposed arnendmetus would entiance patient safety whde allowing the fleuluhty necessary for proper medical care.
January 1990 1 mal ltule The NitC regulations (10 Cl lt Part 34) on safety requirements for industrial radir. graphic equipment were revised. This elfoilis in response to a Commis-sion directive regardmg improved safety in industrial tadiography.
N Ulti!G-1266 A-2 i
sie mu ms Appeltdi\\ (Contit)urd)
- 1) ale llegulatar) Ptodutt flestription i chinaty Paa i mal ilule The N140 tegulatnins (10 UI It Paits 70 and 74) were icun J to t entialue omtenal eonholimJ accounting (h1CA A) ha nunr and m>per tion acliutics for non-teactor laohtics. 'this r u1t making comp!cled the i
phased centialvation of rcrional h1CA A mtintics in llead;.piarters f or norocactor facihues. 'lhn, action was necessary because, rnen the unall number of facilities m the respe(in e tryions. the rertonal olhces could not suppotI and inaint;un the ful! Spectr um of
\\nowedre, skills, and dnuphnes needed to conduct h1CA A bcensing and in' pectien.
I etu uary 14"O l-mal lh port Pubinhed an Nitr.conttactcd NUl(P lleport 106, "llot Par ticles on the Skm.' llus NCl(P wmnuttee g
repoti recominends a special knut for e.sposures from j
hot patucles on the sLm of 75 Fi hr and hem other i
umuibutors and will serre as tuhnicallusc for future N1(C rulemalmp.
Apnlluna N1Iltl'G-1333 1(sued the hnal rvpotI, in support of the Comnus-(
non's rulemakmp cllort on m.untenance, on the rev:cw of inaintenance approxhes and practices in selected forcirn noticat power ptograms and in other U.S. mdustries.
hta) 1%D Pt oposed 1(ule
'1he N1(C epulanons,10 Cl l( Patts 20,30,40,and s
70, on notifications of modents were reused. 'lhe rule would resise licensee reportmg requirements regard-iny the notifications of incidents related to rachatiori Ndel). 'this rule is needed to ensure prompt evalu-ation of a heen'ce's attions by the Comnmsion to protect the public health and safety, Ntay luna 1(eputaiory Guide 3.58 Issued rcrolatory puidante on criticahty safety for handhnr, stonny, and transpotting IMll fuel at fuels and inaterials fanhtics.
June 1440 iteputatory Guide 3hh issued reputatory pmdance on standard format and j
(January 1440)
(Draft lieputatory Guide) content of financial assurance methanistns acquired g
for decomtniwoning under 10 CI'll Parts 30,40,70, i.
and 72.
June two NUlllLG-137?
Provided reg.te.ory analysis for resolung GI C-8 on main steam iso ation valve leakape and leakape con-trol sptem failure.
June 1940 NUllliG-1334 Pronded tesolution for GI-29 on bolting degradation or failure in nuclear power plants.
JuneIWO Generic 1 etter 90-6 issued as patt of res.olution of GI-94 to provide inf ormation on additional low-temperature overpres-sure protection for light-water reactors.
Jul; 1000 Draft Supplement to Pionded puidance en individual plant esaminations Generic I etter 88-20 for esternal events.
A-3 NUltliG-126h
A ppetidis (Conliittled)
Date ikgulatory Product 1)escription July 1990 NUltl'G-1407 Draft report to describe the proposed scope, accept-able methods, and reporting requirements for individ-ual plant esaminations for external events.
July 1940 linal 1(ule
'Ihe NI(C regulations,10 Cl'It Parts 50,72, and 170, to allow holders of nuclear power reactor operating licenses to store spent fuel in NitC approved casks at reactor sites under a general bcense were revned.
This rule, which provides procedures and cntena for N1(C approval of spent fuel stoiage cask design, was undertaken in response to the requirements of ti,:
Nuclear Waste P,licy Act of 1982.
Ju;y IWO pohey Staternent issued a fmal pe":y statement on below regulatory concern (llRC). 'lhis policy applies to the use er dis-l posal of radioactivc materials containing suth small
{
quantities of radionuclides that they do not need to be further regulated.
l July 1900 Supplement 3 to llequested that containment pc.cmance I
Generic 1. citer t&20 improvements be considered tr adividual plant examinations (for all plant types except Mark is).
July 1490 hopos ed llule issued 10 Cl 11 Part $4, proposed icpulation for nuclear power plant lic ense renewah July 1000 NUltlE1362 issued draft regulatoiy analysis to ptovide supporting information for proposed rule on nuclear tiower plant license renewal.
July 1990 NUlti!G-1398 1ssued draft environmental assessment for proposed i
rule on nuclear power plant license renewal.
I i
July 1940 NUlt!!G-1411 Provided NRC's responses to public comments result-j ing from public workshop on nuclear power plant l:ense renewal, July 1940 NUlti;G-1412 issued oi.dt Nument to provide the foundation for
]
the adequacy a.he licensing basis for license renewal applications.
July 1990 Advance Notice of Provided notice of intent to moJify 10 Cl:R Part $1 Proposed 1(ulemaking because of relicensmg action August 1940 Interim l'inal Itute The N1(C regulations,10 Cith )k 35, were revised in response to a petition for rulemaking from SNM/
ACNP that requested departures from l'D A-approved instructions for radiopharmaceuticals so that physi-cians may provide proper medical care. This rule, which was coordinated with the 1 DA in its develop-mont, allows certain departures from 1:DA-approved instructions while providing reasonable assurance of radiological safety as well as a balance between adequate controls and avoidance of undue interfer-ence in medical judgment.
NUREG-1266 A-4
Appendix (Continued) llate Ittgulatory Product lies.oriptiori August 1990 Regulatory Guide 1.1$9 Provided guidance on assuring the availability of funds for decomrnissioning nudear reactors.
4 j
September 1990 I) raft Reputatory Guide Provided guidance on the f tandard format and content j
for emergency plans for fuel gcle and materials facili-1 ties.
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The goal of this office is to ensure that safety-related research provides the technical bases for rulemaking and for related deetsions m support of NRC licensing and inspection activities. This research is necessary to make certain that the regulations that are imposed on licensees provide an adequate margin of safety so as to protect the health and safety of the public. This report describes both the direct contributions to scientific and technical knowledge with regard to nuclear safety and their regulatory applications.
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13 A VAttABiUTY ST AI LMENT Unlimited nuclear regulatorv research
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