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Forwards Corrected & Updated Research Chapter 10 for NRC 1981 Annual Rept
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Issue date:	02/26/1982
From:	Minogue R; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Bassett H; NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA)
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SUBJECT:

CORRECTED AND UPDATED RESEARCH CHAPTER FOR 1981 NRC ANNUAL REPORT Enclosed is the corrected and updated Chapter 10. " Nuclear Regulatory Research," of the 1981 NRC Annual Report. \\ ginalSiped by' Ori penwood r. noss,u. Robert B. Minogue, Director (Office of Nuclear Regulatory Research (sri tion-

Enclosure:

As stated f cc: Janes H. Dickson, SPB lkSDiv on Directors RES Deputy Division Directors RES Branch Chiefs RES Branch Representatives R. M. Scroggins F. Gillespie M. Slater E. L. Hill I L. E. Gallagher A. Eiss 3 S. Castro C. Johnson R. Hoskins E. Podolak v RES:D/DR RESID.y/DA'E r Arse @nruj RES: iSW, Bernero ' Bahbtt RES Task No. N/A g p g,,q g, g,4yggg g,g g,g, ( n RkkPA;ill. --d.RE,S.: .BESd IB d d... f $d$,..'s?.....RaE...........kk ..S, .S]aterb.SB g,.. RES p .R U,:. 0/.0F.Q..... h O tc....... Soller../.'4df.. 5a='h.k1]a e.. .81nogu..... i und orision o,. a, n, PM-l82-- .,,."/82 2/23/82 2/ /82 r %]82 2/ ? V 8203120096 S20226 'Gai RES SUBJ R2914g2 'FFICIAL RECORD COPY I

F i x aG bAb CHAPTER 10 N' CLEAR REGULATORY RESEARCH J The Office of Nuclear Regulatory Research and the Office of Standards Development were consolidated in April 1981 into a newly structured Office of Nuclear Regulatory Research. The new organization is designed to take the NRC research program more responsive to regulatory needs, provide for more effective application of research results in regulations and regulatory guides, and improve the use of staff resources. This chapter is organized to follow the reorganized office structure. Research and standards development work are combined under five categories: engineering technology, accident evaluation, risk analysis, facility i . operations, and health, siting, and waste management. REGULATIONS AND GUIDES NRC standards are primarily of two types: Regulations, setting forth in Title 10, Chapter I, of the Code of Federal Regulations requirements that must be met. Regulatory Guides, describing, primarily, methods acceptable to the NRC staff for implementing specific parts of the NRC's regulations. When NRC proposes new or amended regulstions, they are normally published in the Federal Fecister to allow interested citizens time for comment before they are adopted. This is required by the Administrative Procedure Act. Following the public comment period, the regulations are revised, as appropriate, to reflect the comwnts received. Once adopted by the NRC, they r deral Recister in final for1n with the date they become are published in the e effective. After that publication, rules are codified and included annual % in the Code of Federal Regulations. Some regulatory guides describe techniques used by the staff to evaluate specific situations. Others provide guidance to applicants concerning the information needed by the staff in its review of applications for permits and i 12/10/t') 1 CHAPTER 10 ANtmAt RIPDRT l l 7; i 1 i

T; v x n.,.- j J licenses. Many NRC guides refer to or endorse national standards (also called

consensus standards" or voluntary standards) that are developed by recognized national organi_zations, of tem with NRC participation. NRC seLes use of a nattenal st<.ndard in the regulatory process only after an tr<ependent review by the NRC staff and after public comment on NRC's planned use of the standard has been reviewed.

The NRC encourages comments and suggestions for improvements in regulatory guides and, befort staff review is completed, issues them for comment to many I l I individuals and organizations along with the value/ impact statements which indicate the objectives of each guide, along with its expected effectiveness and impact. To reduce the burden on the taxpayer, the NRC has an arrangement with the U.S. Government Printing Office to act as a consigned sales agtnt for certain j of its publications, including regulatory guides. Draft guides issued for public comment continue to receive free distribution, but the active guides are sold. NRC licensees receive pertinent draft and active guides at no cost. Regulations published during fiscal year 1981 are summarized in Appendix 4. Regulatory guides issued, revised, or withdrawn are listed in Appendix 5. ENGINEERING TECHNDLOGY ~ MECHANICAL / STRUCTURAL ENGINEERING t NRC's mechanical / structural engineering research program provides i technical information to support licensing decisions in the safety review of i nuclear power plants and fuel cycle facilities. The program also develops *.he bases for NRC positions reflected in national standards and NRC regulatory guides and regulations. The program addresses such areas of performance as piping, pumps, valves, snubbers, vessels, containment buildings, concrete structur'es, and soil media in a wide range of conditions.W. i ;;eb N 1: ' r:6M n 1;; ;. ;.c. U t'

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,, a,;/ : n.. :., a..m - ~ ~ ..a = =, = t.. W Sutyrograms are LP. ,,' m.m W.nm W o wi.ar w~.-s. m.. discussed below. 12/10/El 2 CHAPTER 10 ANNJAL RIPORT

O ( g-N -.m Seismic Research and Standards The Seismic Safety Margins Research Procram is a multiphase, long-range program to develop improved methods for seismic safety assessments of nuclear power plants, using a probabilistic coc "ation procedure. Phase I of the program was completed in 1981 with the development and demonstration of a methodology using three computer programs: HAZARD, which assesses the seismic hazard at a given site; SMACS, which computes in structure and subsystem seismic responses; and SEISIM, which calculates the probabilities of structural, component, and system failure and radioactive releases. This methoJology will be used to assess the effect of seissic events on nuclear power plant safety and to identify key areas of possible improvement to decrease risks from them. TM ' J.. vi uir c u. Joigwy..,.ooein 190% in n oiuai;...,, m.

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"m :;;. m.. n:9 My 'e%tWtems% Jrie r ad... ws r.ti.. ^ ;.,.7 ,, wi N 4m 7. q, e, a cyo -_, 4 Response Prediction for Soil-Structure Interactions. NRC's investigations of methods to calculate the modifications in earthquake motion caused by heavy, rigid power plant structures led to the development of a simplified computer i code, " Structure in Media" (SIM), for licensing use in checking license applications. The code was being verified at the end of the yea'r. Reinforced Concrete Panels and Seismic Cost Assessments. During the year, 1 NRC issued NUREG/CR-2049, which examines the strength and stiffness degracation of containment wall panels subjected to seismic cyclic loading, and i NUREG/CR-1508, which presents incremental costs of 1100-to-13DD MWe nuclear power plants as a function of a range of seismic design requirements. Seisele Response and Instrumentation. Two other achievements in the seismic research area included initiation of a study, due for completion in t' 1982, to evaluate the potential benefit of a seismic scram system that would t automatically t' rip the reactor upon sensing high-level seismic activity, and I the issuance for public comment of a proposed Revision 2 to Guide 1.12 on instrumentation for earthquakes. The guide describes the instrumentation aneptabletotheNRCforldeterminingj the seismic response of plant safety features. Fluid Systems and Components Gesearch anc' standardsf 12/10/81 3 CHAPTER 10 ANtfJAL REPCRT y I

e r N A ~ -. toad Combinations Procra suits in 1981 indicatgfatigue crack growth leading to couble-ended guillotine breaks in the primary systee piping of a PWR is extremely unlikely. This information affects licensing decisions and may lead to a relaxation of the requirement to design for simultaneous occurrente of an earthquake and a large loss-of-coolant accider.t. A panel of national experts has stated that reasons exist for concluding that further study will not change the findings already brought to light. @: ". _ ' "--* &#y-tN j$D al. & U y enterennto a cooperative research venture %11 Taipower at Taiwan's Kuosheng Nuclear Power Station, scheduled to be the world's first operating BWR/6 plant f'ng an advanced design pressure-suppression containment .3L (MarLK. Emphasis was given to low-level vibration testing of equipmenth near the suppression pool and to predictions of equipment and piping response to safety / relief valve discharge loads. Research at Heissdampfreaktor (HDR). At the deccommissioned HDR in West f Germary (see yp. 214, NRC 1960 Annual Report), investigations continued into computer code capabilities to estimate piping behavior under simulated seismic i 4 and thermal bydraulic transients. These have shown, in general, that even under controlled idealises situations, large differences may occur between a predictions and observations. Efforts continue to explain these differences and to develop more accurate methodologies. Loose Parts Detection. In May 1981, NRC issued Revision I to Guide 1.13 on the loose-parts detection program for the primary system of light water-cooled reactors. This guide contains guidance for programs intended to provide early detection of loose metallic parts and thus to provide the time required to avoid or sitigate damage to primary system components. Construction and Inservice Inspection Standards. Section 50.55a

Codes and Standards," of 10 CFR Part %, has been amel.ded to incorporate, by reference, certain sections and addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Cod ncerning nuclear power plant components and inservice inspection. This makes the quality assurance requirements consistent for Classes 1, 2 and 3 components, and clarifies both acceptance standards for flaws and the examinations for component supports.

12/10/El 4 CHAPl[R 10 ant 4UAL REPDR1 7 il

l J NRC also has issued for public comment a proposed amendment to update Section50.55abyincorporatingfurtherrefentaddendathrough1980aswellas k the 1980 Edition of the ASME Doller and Pressure vessel Code. Containment Pesearch and Standards Containment inteceity. A new program was begun to compare analytical predictions of containment behavict beyond the normal design loads with results from scale-model tests. The order of test loadings was determined -- static A pressure, dynamic (unsymmetrical) pressure, and lateral (simulated seismic) 9 pressure. The first containment typeg to be tested will be steel, to be followed by concrete. The results from 1981 studies will enable construction to begin on the first small-scale models in 1982. The test results should permit predictions about the ultimate capacities of containments and provide data against which analyses can be checked. Other work on containment response to dynamic loads addressed the sensitivity of response to a uniform Pydrogen 1 burn pressure. aisana6.- Containment Construction. NRC took another step toward endorsement of the ASME Boiler and Pressure Vessel Code's Section III, Division 2, ' Code for Concrete Reactor Vessels and Containments," with the issuance in June 1981 of Revision 2 to Guide 1.136 on materials, construction, and testing of concrete containments. Acceptance of this national standard made it possible to withdraw six regulatory guides: 1.10, 1.15, 1.18, s 1.19, 1.55, and 1.103 (see Appendix 5 for guide titles). ~ Structural Research and Standards Protability-Based Lead Criteria. NUREG/CR-1979, issued in fiscal year 1981, provides en in-depth review of the current use of probabilistic concepts and procedu used to determine the load combinations for the design of Category I structures, which are those structures designed to remain functional e if an earthquake producing the maximum vibratory ground motion should occur. Work also was s' tarted on a data base for various 1 cads and resistances. Safety Margins for Cateaory 1 Structures. The buildings (other than containment) that house safety-related equipment at nuclear power plants are massive concrete sher,r-wall structures, whict)because of their safety function, Go are subjected to loads and load combinations tg dif fer f rom f ramed structures. $ ne program will supply experimental information needed to assess the capability of such structural systems when Icaded beyond their design limits. 12/10/81 5 CHAPTER 10 ANNUAL REPORT .r

7.- sN l Structures considered in this program include fuel buildings, diesel generator buildings, and auxiliary buildings. The NRC published a plan for assessing the capability of Category I structural systems and began the first phase of the program with the fabrication and testing of small-scale models, f Other Concrete Structures Standaedy :' *

Guide 1.142, on safety-related concrete power plant structures (other than reactor vessels and wa.S containments),31ssued in October 1981. It endorsed an American Concrete Institute Standard (ACI 349-76, " Code Requirements for Safety-Related Concrete Structures")

d its 1979 supplement. Equipment Qualification Research and Standerds snubbers. As part of the NRC effort to improve the reliability of snubbers, a draft guide on qualification and acceptance tests was issued for public comment in February 1981. It provides guidance for functional specifications, for prototype snuboer qualification testing, and for acceptance tests of those that will actually be installed. Active Valve Assemblies. " Active valves" must, during or following n l postulatedaccidengperformamechanicalfunctiontoshutdowntheplant, maintain the plant in b.fe shutdown condition, or mitigate the Censequences of a pestulated event. In March 1981, NRC issued Guide 1.148 on the functional specifications for such valve assemblies, supplementing the ANSI f tandard which provides guidance for their minimum function and operability specifications. Safety and Relief Valves. A TMI-related inr'ustry test program to demonstrate the capability of safety anc relief valves to operate satisfactorily under all anticipated fluid conditior.s neared completion in 1921. The program incir#es testing of safety and relief valves, block valves, and associated piping. It is being monitored by NRC which will review and evaluate utility submittals on plant-specific valve and piping systems, identify codes and modeling techniques to confirm the adequacy of valves and piping, and verify hydraulic load calculations in valve and associated piping and supports. New test programs will be identified if required. 16'LilALS Ef.CINEERING I NRC's metallurgy and materials research program deals with the safety and serviceability of reactor pressure vessels, major piping, and steam generator 12/10.'31 6 CHAP 1ER 10 ANNUAL FIPORT i

7 N _, sb tubing - coarponerts of a reactor's primary system. The program incIudes studies of fracture wchanics for pressure vessel and piping applications, environmental operating ef fects, and nondestructive inspection techniques. A I -- t____.. These are discussed below. b. M,.,,Jc.L w b-n:1_% 0 F raee 14echanics) - ~ L m Q=_ w X =&~ M ^ ~__=_ t v - 9 g___ Fracture mechanics studies are directed at developing and validating methods for evaluating and ensuring reactor vessel and primary piping integrity. Areas of concern include thermal shock and pressurt2ed thermal shock to reactor pressure vessels, irradiation-induced loss of toughness in pressure vessels, and the capacity of degraded piping.to withstand earthquake and dynamic loadings, discussed below: Therwal Shock. The seventh in a series of thermal shock te'sts (see

p. 211, NRC 1980 Annual Report) was completed at Oak Rid, National Laboratory (DRNL) in 1981 to validate that thermal stresses alone will not drive a crack through a reactor pressure vessel wall. The test used a

. wall-thickness-to-vessel radius more representative of actual operating vessels. The tests have been aimed at validating linear elastic fracture mechanics concepts and evaluating the effects of loss-of-coolant-accident (LOCA) and thermal transients on pressure vessels. Recent developments suggest the need for further tests to evaluate the effects of reactor we'ssel cladding on cracks and on the propensity for short flaws to "run long" under thermal shock conditions. Planning for these tests was begun. Pressurited Thermal Shock. During 1981, researchers at DRNL designed a pressurized thermal shock facility to use an externally flawed test vessel. The external surface of the vessel Will be thermally shocked while pressure is applied internally, a test configuration that should permit duplication of a wide range cf transient and postulated accident conditions at little cost. Lonstruction is scheduled for completion in 1982, and the first test of a ~ series is tentatively scheduled for 1983. Under this program, researchers also produced special computer codes for use by license reviewers in heat transfer, thermal stress, and fracture mechanics calculations for reactor pressure vesstigami for probabilistic eva16.ations of reactor pressure sessel failure. Itastic-Plastic Fracture Mechanics. Fracture of steel used in reactor pressure vessels and piping can occur brittlely, ductilely, or in combination. Brittie fallute has long been analyzed by linear-elastic fracture mechanics. l 12/10/61 7 CHAPTIR 10 ANfluAL REP 0RT i

. p o e g a Elastic-plastic ' techniques for analyzing cVtile and mixed mode fractures are a more recent and rapidly developing area of research, !sportant for evaluating high-temperature conditions where tLe materi',is tenain in the uctile f ailure In 1981, work at DRNL, the Naval Research Laborato(rg, avid Tay s'RL range. hav41 Ship Research and Development Center, and the F 5. Naval Acadeey was designed to develop and validate test techniques and data bases. Benefits are applied directly to NRC licensing activities in fields such as reactor pressure ) vessel toughness (Generic Issue A-11), pressurized thermal. shock, and I leak-before-break in piping. Fracture Toughness Reovirements. On November 14, 1980, NRC issued for public comment general revisions of Appendix G ' Fracture Toughness Requirements," and Appendix H, ' Reactor Vessel Material Surveillance Program i Requirements,' to 10 CFR Part 50, clarifying the applicability of some require-ments, modifying others, and expanding the references to national standards. j 'In 1981, the public coments were resolved and the final rule prepared for management review. 2 E R M Degraded pipine and Protability of Failure. The 1981 programs addressing piping reliability used both deterministic and probabilistic approaches. The deterministic approach concentrated on elastic plastic fracture mechanics analyses techniques, fracture toughness data base development, and degraded pipe tests. Intermediate-s12 pipe tests were conducted at David Taylor, and a degraded-piping program was begun at Battelle Columbus Laboratory to demonstrate the capacity of degraded piping to withstand postulated acciderr and transient loadings and to evaluate the elastic plastic techniques in predicting load capacities and f ailure mades. In the probabilistic approach, a computer code for detersining the probability of failure or leak cefore break was expanded by Lawrence Livermore National Laboratory include additional variables such as stress cerrosion cracking and residual stresses. The code bas been used to generate input for the load combinations program and will be used for reevaluating the current' criteria for postulating pipe-break locations. A Oper6tino invironmental Effects Studies in the area of environmental effects include radiation effects on s.aterials, steas generator tube degradation, and stress corrosion cracking in primary piping. 12/10/e1 B CGPTIR 10 AtmuAL REP 0RT T. '

o s N ^ Irradiated Fracture Toughness Desimetry. and Fatigue Crack Groeth. Research in 1981 on the ef fects of radiation on reactor vessel stevis included irradiation and testing of fracture mechanics specimens to define the relationship between fluence and. reduction in fracture toughness, with emphasis on developing elastic plastic fracture touDhness data for irradiated specimens. This information is needed to demonstrate whe' her operating pressore vessels can maintain their integrity in both normal e ' sccident conditions. Work also NE L continued smder the irradiation-anneal-reirra istion program at the 4 eve 4* t - , on the ef fectiveness of annealing in restoring f ract.ure toughness to irradiated steels. NRC sponsored dosimetry work at DRNL ard the Hanford Engineering Developmer,t laboratory to establish benchmarks for validating and improving fluence calculation techniques. rk continued at NRL I 5 I on fatigue crack,-prowth rate for reactor vessel steels under various cyclic loading forms. Data from this program will be used to revise Section XI fatigue crack growth-rate curves of the A5ME Code. Steam Generator Tut., Integrity. The full-sir steam generator (see NRC 1979 Annual Report, p. 230, and NRC 1980 Annual Report, n 212) which the I w. sl 6 u s. k 6 NRC moved from Surry, Va., to Richland, Wash., in 1980 - _- a Fac444c I hee %we+44abontery-fac+Hty for a wide range of research studies. yen, skm ; e.. .*e/ Ce"dO invirossentally Assisted Pipe Cracking. ', - dr, /", .) from both normal and accident conditions can contribute to + = = ' " - cracking of reactor pipes in combination with the other conditions of metallurgy and loads. In 1981, the NRC published the Argonne National Laboratory's review of pipe-cracking literature (see p. 212, NRC 1980 Annual Report) and beDan new research on these problems. System and Component Criterie; Genera' "- ' "am4 December 1980 and in August 1981, Revisions 17 and 18 to Guides 1.B4 and 1.85, which list acceptable ASME Boiler and Pressure vessel Code, Sectioh III, Division 1 Code Cases as well s's those Code Cases annulled, revised, of reaffitted since ( 1 s t q

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ML jnteptionoftheseguides,[ereissued. n.d i,y.;- Ac n... ' 4., #.b ( ta r..,H '.D i. s Mi1 (d (** *9 ' D MmJ W.l.A hondestructive Imamination d g, d.,, 'f". jud ar 19 f f p This program includes studies of inservice inspection techniques to find and characterize flaws more easily and reliably and studies of methods for ' continuous monitoring for that purpose. S, ;-p n t d wde >-ceve ntyW E,,.

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e l F Flew Inspection by Ultrasonic Test. The improved ultrasonic testing (UT) method developed at the University of Michigan'W c- 0 .p.

196, see et 1978 Annual Report,and pp. 212-213,70980 nnual Report), called SAFT

($ynthetic Aperture Focusing Technique), has proved much better than earlier UT methods. The Southwest Research Institute, which has constructed a SAFT-UT inspection system for the NRC, was preparing to tak.e the system into the field for trials at year-end. However, until the new UT developments become standard, it is still necessary to determine the reliability of current methods. Batte11eI ific Northwest Laboratory (PNL) continued its efforts to define current inspection reliability and to deduce the best inspection methods. In 1981, PNL recomendations were being incorporated into the ASME Code for improving the reliability of inservice inspection. In addition, NRC issued Guide 1.150, which describes acceptable ultrasonic testing of reactor vessel welds during preservice and inservice examinationgin June 19Ep.d C.w 1.1", ." :.;...Mel. A n U.6... hie Ted C.A, L e n "!, P _

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"....., 1%; I CHEMICAL ENGINEERING NRC's fhemical fngineering fesearch frogram areas such as decomissioning, fuel storage, waste treatment and storage, criticality, ventilation, effluent treatment systems, hydrogen control, and fission product control. These and.. others are described in the sumary that follows. Decomrrissioning Technical studies for the NRC continue ~.o develop a decommissioning information base for light-water reactors and other nuclear f acilities. Four reports dealing with decomissioning were published in 1981 cover ling: (1) non-fuel-cycle nuclear facilities (NUREG/CR-1754), (2) uranium fuel f abrication pla ts (NUREG/CR-1266), (3) monitoring for compliance with cecomissioning e n.w i y g:,c, _ e p g crit 7tia (NUREG/CR-20B2), and (4) en adtendum to NUREG/CR-05707 on A envirtnmental surveillance programs for 13w-level waste burial grounds. Three other reports were nearing completion at yrar-end. The reports are part of NRC,*s continuing reevaluation of decommissitining policy. A report on assuring fand availe.v11ty and a draft generic 12/10/e1 10 CHAPTER ID AtmUAL L'. PORT T.

o, ab environmental statement on decommissionin nuclear facilities also were published. kwU68'> " C*"** 3' L;m 3c3 undeg de P W i. Ongoing research projects to help develop decommissioning standards and guides deal with long-lived activation products in reactor materialsg j decontamination methods to reduce occupational exposures, offsite relearts, and radioactive waste volumes; and radioactive contamination around typical LWR plants. A literature review on decontamination processes that are precursors to decomissioning (NULEG/CR-1915) was published during 1981. I%,m..ars d sd.ub oe. M.v.,6abow kav. Lw co m p t.bl nA A d.4C;.Lw uwebo a,d e=a be;g %Lbk k c% WR Lases. Spent Fuel Storage In November 1980, the NRC issued 10 CFR Part 72,

Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage j

Installation," as an effective rule. Revision 1 to Guide 3.44, providing the standard format and content for a safety analys reportforrwater-basigtype independent spent fuel storage installation (15F51), was issued in %s . December 1980. 4== ore t guides were issuen during fiscal year 1981: one providing the standard format and content for a safety analysis report for an 15FSI (dry storage)Ma*e-%e-e4w addressing spent fuel. heat _ generation in an.qh. dot.U.s ba 11 ^ Ww.M. su.z a.,k,A pp.. 6 M- *-

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_t,,, A..*~ of. - i Researchlo BetiraT6e Tiut11de1nventories and afterheats of LWR spent fuel _ was undertaken in order to provide standardized information to applicants concerning long-ters heat generation rates of power reactor spent fuel as a. function of burnup and decay time. The project data basis and SCALE system hab sateA codesgere compared to experimental measurements during 1981, and results will i be reflected in the appropriate active guide. l Nuclear Criticality Safety Guide 3.45, on nuclear criticality safety for pipe intersections containing sque'ous solutions of enriched uranyl nitrate, and a proposed w Revision I to Guide 3.1, on use of borosilicate glass raschig rings as y neutron absorbefin solutions of fissile material, were issued in Novecer 1980 f and May 1981, respectively. Also Guide 3.47, on nuclear criticality control and safety of homogeneous plutonium-uranium fuel mixtures outside reactors, was issued in July 1981.

1 12/10/E1 11 CHAPTER 10 ANNUAL REPCRT Ti l

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( = i Experiments to provide benchmark data on spent fuel storege, shippeng configurations, and process geon=tries using low-enriched uranium oxide continued to provide data used to validate NRC methods of analyzing licensee criticality safety programs. Effluent Treatment Systems Measurements continued at the Prairie Island plant to obtain radionuclide source term data for use with gaseous and liquid effluent models for LWR licensing. NURt'G/CR-1992, evaluating the effluent treatment systems at four operating LVRs was issued during fiscal year 1981. Hydropen Control g k* ene2e e A program was{ Me in 1981 to evaluate equipment con:epts and operating schemes proposed to pn. vent sudden flareups and/or detonations, and schemes to sitigate the effects of hydrogen burns in light water reactor plants. Fission Product Control A report (NUREG-0771) on the regulatory impact of nuclear resetw eccident source term assumptions was issued for public somment during fiscal year 1981. A program under development to facilitate review and evaluation of fission product control systems will examine engineered-safety-feature syster ~ effectiveness under various accident conditions, existing designs taking into account expected aerosol concentrations, fission product chemistry, and the behavior of iodine in chemical environments experienced in past incidents. ELECTRICAL ENGINEERING Qualification of Electric fouipment Research at Sandia P P Liv '" L, in 1981 improved NRC's understanding of ment qualification testing methodologies and aging techniques, as the 3aK facility was upgraded to accommodate larger equipment. Verification tests of connector and electrical penetration assemblies and certain cables were conducted in an BB-inch-high by 20-1/2-inch-diameter pressure vessel. Accelerated aging tests identified strong synergistic effects 12/1D/21 12 CHAPTER ID ANNUAL REPORT

c 4Y in certain insulation materials and demonstrated'the influence of the tes seguence on material degradation. he first tests were conducted in France as part of a joint U.S./ French test series under LOCA conditions to judge the relative degradation of elast metic materials with varying crygen concentrations. A proposed rule, " Environmental ualification of Electric Equipment for c$e Nuclear Power Plants," was ' de$e-In addition, the development of three regulatory guides dealing specifically with the qualification of lead storage batteries, motor control centers, and battery chargers and inverters was begun. Fire Protection hkA Fire protection research continued at 6eet as a full scale replication fire test was completed, and new research programs testing the validity of a 20-foot separation distance between redundant cabit trays were initiated. On February 17, 1981, new fire protection regulations became effective for nuclear power plants licensed prior to.!anuary 1,1979. Develcpment cf a comprehensive fire protection regulation for new nuclear power plants was begun. ACCIDENT EVALUATIDN ~ EXPERIMENTAL PRDCRAMS Experimental programs research' covers the integral systems and separate effects tests needed to support the reactor licensing effort. In 1981, n : 4 Anrssed continued to address items identified in the TMI-2 Test Action Flan, stressing small-break less-of coolant accidents (LOCAh) and the operational transients and accident de' ection instrumentation pointed up at TMI. t The following sections describe these research efforts. letc~el Systems Tests The NRC is the s.ajor source cf support for both the loss-of-Fluid-Test (LDFT) and Semiscale PWR test facilities (see previous Annual Reports). LDFT receiv'es approximately ten percent of its support from foreign countries.For 12/IO/EI i 13 CHAP 11R ID ANNUAL R[POR1 'i l 1 ,[i v 1 i )

a x 4 a third factitty, the Full Integral Simulation Test (FIST) BWR test facility, the NRC, the Electric Power Research Institute (EPRI), and the General Electric Cumpany (GE) share supporting costs almost equally. Te%t plans for all three facilities have been modified to include sa.all-break LOCAs and operational transient experiments as well as the original plans for large-break c0CAs. %L0ri Procram. During 1981 the LOFT program: Issued Research Infore.ation Letters on the small-break LOCA experiments performed, the Au ented Operator Capability Program, the Technical Support Center established since the TMI-2 accident, the operational transients performed, and an in-depth study comparing nuclear and electric heater rod performance. 1 Conducted experiments involving an open pressurizer power operated relief valve (PORV) in conjunction with a loss of all feedwater, a simulation of the Arkansas haclear One - Unit 2 turbine trip transient and associated effects, an intermediate-size-treak LOCA equivalent to the rupture of an accumulator pipe, and a core uncovery accident at high decay heat level. Conducted a modeling workshop in conjunction with Semiscalt to explain the experience gained in modeling the two facilhies to those analysts involved in code development, assess ent, a:td standard problem calculations. m. _ v.m 1 Finally,,1981 nr M H et W plans to close, d5 contaminate, and. 3 decommission the LOFT f acility following the test program in 1983, as directed by the Commission. Seeiscale. During 1981, several test series were completed on the Seeiscale test facility. (For a description of the facility er p. 198 .\\ A*A=.te W These included: f. Characterizat on analyses and ) NRC 1980 Annual Report.) tests that provided a component-by-component understanding of system heat loss in PWR's. Seven tests covering cold leg break configurations, with and without

6sz ca i

operation of upper head injection (UHI) subsystems. Breatf tested were d.m Akk 2.5, 5, and 10 percent s no significant core. heating was found in any g test. The $ percent break caused the greatest core uncovery, as had been predicted. Three cumulator configurations were evaluated: the standard non-UHI Westinghouse PWR setting (600 psi), the standard UH] NR 12/10/El 14 Cht#1[R 10 AWUtst PIPORT T 1

n o _a b Y setting (400 psi), and me) 400 psi without UNI. Although the results do not apply directly to a PWR because of some nontypical items, the test data can be extrapolated to PelR conditions. N hatural convection tests wart":"provided valuable information on the effects of noncundensible gas and two-phasj flow over a wide ranDe of c d ctem Quality. Several tests were also man to octain data under transient p een t, f conditions anga,a the infirence of emergency core coolant injection on natural convection behavior. l l The 1982 Semiscale program calls for further improvements in hardware and a program including tests of 25, 50, and 100 percent breats, evaluation of loss of station power and recovery methods, and study cf events such as steam generator tube breakage and Icss of malp. circulation pump seals. t-Q wg.ga.msenu %r I BWR FIST Facility. The BWP f1STJfacility in San Jose, Calif., is an h upgrade of the two-loop test apparatus (see p.199, WRC 1980 Annual Repert) to 'aprove the simulation cf various BWR transients. FIST, sponsored jointly by C E., . NRC, EPRI, and R : ~. M 11 use a single, full-sized electrically ? heated fuel bundle operating at typical BWR pressures and temperatures. Plans i for the facility were completed for 1982 and 1983 tests. BWR Counter Current Flow tirit Refill /Feflood Program. Plans for this t facility (cescribed in the NRC 1980 Annual Report see p.199) were carried out j in 1981. Simulations of the late phases of a BWR LOCA transient were e conducted, and the code models were produced fcr the BWR version of the TRAC.. code. (See the section on " Analytical Madels.") Separate Effects Experiments NRC separate ef fects research involves experiments in the FLECHT-SEASET facility shared with Westinghouse and EPRI, acquisition of model development data for use in computer codes, instrument development for use in experimental facilities, and the international 2D/3D program. FLECHT SEASET. In 1981 this program was expanded to include three major investigations: heat transfer ef fects of blockage in fuel bundles; separate effects of Ley componer,ts during reflood; and prisiary system behavior under M differ *g modes of natural circulation for long-term PWR cooling. The flow 4etf blockage ddressing the reauirement of Appendix K to ID CFR Part 50 to provide data to assess vendor licensing computer models for reflo[d, has been 12/10/BI 15 CHAPTER 10 ANNUAL RIPDRT I

r a S A completed. Thus/ far, both the 17 x 17 unblocked bundle and 21-rod blocked bundle tests have been completed. The 17 x 1? blocked bundle facility required to complete this series is under construction. Tho' separate effects test ta for the steam generator tests have been analyzed. The natural circulation Y f system effects test facility has been constructed, and tests were undepay at year-end to investigate the system behavior of single-phase, two-phase, and a reflux natural circulation. PWR Blowdown Heat Transfer Procram. A variety of film boiling and bundle f uncovery/ recovery tests were conducted in the Thermal Hydraulics Test Facility i at ORNL to obtain bundle heat transfer data for small-break LOCA conditions in a PVR. The final test series, completed in hoveeer 1980, produced data which have been stored at the data bank at Idaho National Engineering Laboratory. An initial analysis of the data indicates that steam cooling of an uncovered bundle can be adequately predicted using a modified CITTUS-BDELTER correlation i j i and that radiative absorption is significant at high steam pressures. Model Development. Most NRC model development research is funded at i universities. These programs are aimed at supplementing separate ef fects experiments, helping to interpret data from larger test programs, and Y developing correlatiy based on a phenomenological understanding. Some current efforts sponsored by NRC include (1) a program at Lehigh University, including p-rW A L. development of the necessary instrumentation, collection of data on heat flua boiling, and formulation of models and correlations; and (2) a program on phenomenological modeling of two-phase flow at Argonne National Laboratory. These models provide a basis for developing multichannel computer codes. Other experimental studies at Nerthwestern University are providing infors.stion that is serving as the basis for verifying models of containment flooding, emergency core r.coling penetretion, and pressure crop. Also, the State University of hew York at $ tony Brook started research to describe the heat transfer enhancement caused by flow blockages such as grid spacers. Advanced instrument Deveinpinent. Some of the research instrumentation N expertise and faci'ities described in the dBD A9nual Eepert have been used to j develop and evaluate ne power plant instruments, and the transfer of ibis technology to the industry vas emphasized in 1981. For enample, during fiscal year 1981, m m. =" heated thermocouple a ultrasonic ribbon liquid-level indicators were developed and tested by industry andgRC arranged 7 12/10/El 16 CHf,PTIR 10 AtitrJAL R[PDR1 l \\ 1

~ s w M . f for vendors to test and evaluate certain of their ema devices in conjunction g with scheduled hRC tests. Combustion Engineering tested a heated thermocouple sensor at DRNL, and the Westinghouse dif ferential pressure system nas installed at Semiscale, where it has been evaluated for various LDCAs and transients. The pulsed neutron generator was delivered by Sandia t. . R.. n b for use in in-situ instrument calibration and slow flow measurements at 2D/3D and other test f acilities. (See below.) 2D/3D Procram. The NRC has been participating in a joint nsearch program with Geesany and Japan since 1978 to st-@ various aspects of PoIR operation. Two integral systems test facilities an located in each country. NRC i furnishes advanced instrumentation and analyses for the testing programs. (See f 1980 Annual Report, p. 201.) l b The Japanese Atomic Energy Research Institute (JAERI), as part of the 2D/3D frogram, completed the first series of tests at the Cylindrical Core Test Facility with results essentially identical to those reported in 1980. The .Japar ese have also begun initial tests at the newly constructed JAERI Slab Core Test Facility to study full-scale flow behavior in the radial and axial directions. Results will be reported in 1982. The Federal Republic of Germany completed the design of the Upper Flenum Test Istility with a full-scale reactcr vessel and internals using a core simu1Geo ey a steam and water injection device. This facility effers a unique feature of studying, in full scale, de-entrainment in the upper plenum in the.. ~' reflood phase, the ECC water bypets in the refill phase, and the phase separation in het legs during a small-break LOCA. A large number of two-phase instruments developed in the U.S. under the 2D/3Dfrogram,anddescribedinthe198Dreport,performedsatisfactorityat the JAERI test f acilities. FUEL EEhAVIDR RESEARCH A a,ajor redirection of hRC's fuel behavior research program occer&d in S 1981 when the emphasis of the program changed f rom design basis and LDCA( to accidents involving severe core damage such as the event at TMI-2. LOCA a'nd Operttior.a1 Trendent Proteams 12/10/El 17 CHAT 11R 10 AtmuAt R[PDRT 3

a =. r (, ab Multirod Eurst Test (MdCT) Propram. The MRET program at OPNL to --.A.. ues.x ese.LE ~ invest' gate the behavior of Zircaloy cladding (see MC 1980 Annual repg, g

p. 2C2) featured continuation of the single-rod tests cescribed in 1980,

,d conduct of a multirod burst test ei e 6 x 6-rod bundle, and emarndtion of the p 8 x 6-rod bundle that mas burst' tested in 19P0. Final analyses of bloctage cata of both multirod bundle tests were under way at year's end, and a final WWD report on the MRET program is expected in 1982. gC/ Power Bu st Facility (PDF) Procram. At the PBF in Idaho (see 1977 Annual e "tepert ar.d 1 PD Annual Report, p C3), tests conducted in 1981 included A two simulating accident conditions expected in a large-break LDCA, and two i tests to otsstve the influence of thermoccuples used for measuring surface temperatures on the quenching behavior of the fuel rods during a LDCA. Trese latter two tests were conducted specifically to aid in the interpretatie9 cf data ottained from earlier LDFT facility tests. Plans and designs were oeveloped for severe fuel damage tests in the PBF in 1982-1983. The two LDCA tests support previous observations on circurferential strains during ballooring that ttnee strains faee a."d in irradiated fuel rods were only slightly greater than the strains unitraciated rods, though there are too few data points anallable to lead to a reliable Conclusion. The V tests on the influence of surf ace thermocouples on ehe' quenching behavior j showed that while the thermocouples mounted on the exterior surface of the fuel rod cladding did cause the rods to be quenched sligt:tly earlier and to produte. somewhat lower tergppratures trian for fuel rod without them, the errors hwit;ah 4

c. a. g e, 3

produced were nci p

y. mgh n ;-- as.e the effects ascribed to them during l

the LDFT tests previously conducted. i NRU Procram. Three joint NRC/ Canadian tests were performed this year in i i S c.se. be M *not ' the NRU reacter, Cha er, Canadag % e first in-rear,ter evaluation e- .. a. e., w. . m...a of LW thermel-t'ydraulic and 4*e mechanical ballooning and rupture,,meterialW 7 I i. - c',a fDI-length Pd fuel bundlej i Current commercial enrictments an:L fuel designs of a 17 x 17 Pe!R fuel bundle were used in the tests. t f,-. / t v** eat Halden Reactor Tests. Comprehensive data to rification of fuel performance computer coots were ottained in 19El from instrumented 0-rod test Q aan assembitagoesignedandconstructedatFacificNorthwestLaboratories(PNL)and irradiated in the Haldei reactor,in Porway. One of the three assemblies j 12/10/f1 IB CHAPl[R ID ANTR'At FEPORI i

>.4 - 1 e e 1 3 i \\ h } i-t d 'l I t I i . JU 1 4 .in se s.e.L The results of these tests indicate that nuclear-heated fuel rods Quench faster than anticipated. This can be attributed to the effect of a full-length fuel bundle. the effect of nuclear heating vs. electrical heating, and ballooned vs. defoned rods. The tests have also shown that i circumferential tenperature gradients of 250F to 30*F are connon. This is j important because it means reduced cladding defonation during ballooning. 1 I l 4 e i i e t l l I __- fj % d LT l I P> o-I

Fg o. r J J removed from the reactor had reached an average b'urnup of 30,000 MWWMTM. Two 44 w4 ee. other assemblies designed by INEL continued under irradiation. g Fuel Rod Analysis Procram (FRAP) Codes. The development and assessment of JtRC fuel behavior computer codes, FRAP T, used for the analysis of fuel rod response during off-armal reacter conditions, and FRAPCON, used for the steady-state analysis of fuel rod response during nonnal conditions, have been completed. Both codes were availablept the National Energy Softwan Center } or distr g t year's end. SeyereFuelDamace(SrD)Procrapp T.- ,\\.:.;. JOi:4 z 't 6: q-d G r.. M P: ht: ;.--J u.W.,legy fe,O etter u erstanding of core behavior in severe accidents addressing the following questions: (1) What are the physical and chemical states of a \\ / reactor core aft fuel rod temperatures have exceeded 1478'K (2200*F), (2) whether the severe amaged core can be cooled by ftflooding, and (3) what l eanagement ptocedures, r ty systems, instrunnts, and d .ostic information i . are necessary to *,ersinate i.e accident at different polvitt in the accident sequence. A report on SFD (NUREG-DB40) issued the d of the year recommended (1) N finding increased funding for the FEF SFD. s program and (2) supporting the early examination of the TMI-2 core. Th report was given an independent peer review. Steam frplosion. Thegojecti've of the steam explosian phenomena prog *am.. at Sancia is to develop criter to assess the probability and consequences of suchanexplosiondurin> ostulated reactor core meltdown accident. The s consequences of a steseh explosion la a specific reactor systemifZion Nuclear / \\ Station) were anal red, and the probability of containment failure due solely to the explosi was estimated. The upper-bound probability of con'tainment failure is stimated to be I percent, while the best-estimate value is p mbsh*4y .01 per nt. This does not mean that staan explosions can be disregarded dur g a postulated meltdoen accident, but rather that more emphasis should be s s~ n, 1m -,g, . :putawmt,.m a,,,,weeg soeg, ,e. < f n..d (;, A m d.the 5FD ,f Severe Cere tamace Analysis Pecksce (SCE,AP). Since success of bm A N; ar. I program is strongly dependert on >@ymr developn.enga comprehensive pode. IS 85 fu 7 mm SCDAP, htfing developed y xEt prsetet the following in an LWt fue14.=ent gp tamalt fuel rod temperatures as a function of time and axial position; the u r. Seaccc Ar.dded ce 64t 12/10/01 19 CHtJTER 10 ANNUAt REPORT t e J

e s v N i, i l ^ aO Ner 2. ial NRC task force was organized in 1981 to examine the needs for researth on severe fuel damage and the proper test facilities for such research, in response to the accident at TMI-2 and the recomendetions of the Comission that investigated that accident. The task force report (NUPIG 0640) concluded that research on the state end cooli.bility of a severely damayd core under Severe-accident conditions is :weded to form 1 a technical basis for licensing and rulemaking actions, accident sanagement planning, and probabilistic risk assessment. all for accident conditions beyond the design basis. The task force report received extensive peer 4 review including foreign representation. h dsk force recoi:nended an integrated four-part program of rJsearch to The t provide the needed inforwation (data base and verified analytical andels). ~ The first part consists cf integral, nulti ffects.13-pile tests in the PBF f to provide early scoping data on governing phenomena and later for proof tests of the severe fuel dcmage models and codes developed in the program. The second part consists of separate effects experiments on the governing phenomena, both in the ACRR test reactor and in the laboratory, to furnish a data base for sodel development. A Severe Core Damage Analysis Package (SCDAP) is the third part of the integrated program, and this includes development of severe fuel damage nodels from the experimental data base and their integration. There will be continuous active interaction and feedback between the analysis and experimental programs. The fourth part involves the information to be obtained from the TMI-2 core examination. i xplosions. The objective of the program of steam explosion research 4 at Sandia is to develop information for assessing the probability and the lm t i I *1Tc ET 2 _JQ o 7F I i

-e c N ? I' aN t consequentes of a steam explosion dhng a postulated core-meltdown accident, with primary emphasis upon steam-explosion failure of the contain-ment. In 1981. continued experiments with 20-tg-scale drops of core melt materials into mater substantially broadened the data base on the conditions under which steam explosions occur and on their severity. Analysis showed that the only risk-significant node of cor.cainment failurt by steam explosion is siissile generation by the steam-explosion failure of the reactor vessel during a core-meltdown accident. Probabilistic analysis of this process showed that this probability is actually more than an order of magnitude (factor of 10) less than the 1 percent used in the Reactor Safety Study (WASH-1400). t 1.ebe A C k de AJ 19 1 y 1

e w a p A total quantity and types of fission procucts released from the fuel; fuel rr[d deformation, the amount of hydrogen generated and released, and its axial distribution; amounts of lighied and resolidified cladding and fuel material; the amount of exidation of the cladding; the total mass of rubble debris and its distri utjQn; estigate f the flow blockage expecteG-d k feJa % d b 4C41 =L w e a Fission roductReleaseandTrensoort9rocramy i NRC's ongoing research programs on the release of fission products from overhetted and melting fuel, and their transport through primary reactor coolant system piping and containment buildings, are designed to provide the data and codes needed to estimate the potential consequences of severe (See pp. 204-5,p AC, accidents. 1980 Annual Report.) 4 A new facility has been constructed at Oak Ridge to measure the release of fission products from irradiated commerhl fuel rods to temperatures exceeding 2000*C. In a related prograru, shcrt fuel rod bundles with simulated fission products were heated to mehing to determine aerosol formation rates. Other j ORNL tests will try to measure the effect of steam condensatinn on the behavior of aerosol matgrjdu Ldedawe,e4als witbin the gentainment for use in aerosol models be Ment Le developed at NORNL also is investigating the chemistry of iodine and tellurium fission product species in aqueous reagetor solutions under the temperature and pH conditions expected during severe accidents. The cheristry of fission product species in high-temperature steam / hydrogen and steam / air environments expected in LWR coolant systems and containments during severe. accidents is under study at Sandia. Research programs are undgway or planned for 19E2 on the performance of engineered-safety-feature fission product removal systems under severe accident conditions. L* NUREG-0772, issued in June 1981, describes the best technical information available for estimating the release of radioactive material during postulated reactor accidents and to identify where gaps exist in our knowledge. It ob focuses on low prability-high consequence accidents involving severe damage to a the reactor core and core meltdown that dominate the risk to the public. Furthermore, in this report particular emphasis is placed on the accident behavior of radiciodine because (1) radiciodine is predicted to be a major contributor to public exposure, (2) current regulatory accident analysis precedures focus on iodine, and (3) several technical issues have been raised m 12/10/81 2D CHAPTER 10 ANNUAL RIPORT T

p, m recently about the sagnitude of todine release. 'Aeroscis in general also were investigated in some detail to assess their effect on fission product release estimates and to determine the perfors.ance of engineered safety features under accident conditions exceeding their design bases. (See Appendia 7 for a complete listing of NUREGs.) SEVERE ACCIDENT ASSESSMENT Severe Accident Seavence Analysis (SASA) Program The SASA research program focuses on possible sequences of events beyond e design basis accidents to calculate how power reactors and operators can function in order to prever.l. or mitigate adverse consequences to both the plant and the public. Four major laboratories are involved in the SASA program-- Idaho,LosAlamos,SandiyndOakRidgeNationalLaboratories. Three labs are investigating PWR accident sequence y : ? " - Unit 1-4 W th Los Alamos and Idaho analy2ing the " front-end" (up to core damage) and Sandia the 'back-end" (core damage through containment damage). Cak Ridge is focusing on BWR severe accident analyses, both front and back ends. The Los Alamos program involves calculations for

hands-off" accident scenarios (LOCAs) involving failure of the power-operated relief valve to recloseNnd the rupture of U-tubes in a steam generator.

The studies in Idaho address a matrix of four small breaks with and without high pressure-injection f ailur 2L. C

Qd the results have been analyzed regarding options available to an operator.

A study a) Sandia /n&w. Mar e,alyzes,a hypothetical core m A4 5

Ila **

"e w h't ofoff-sitepowegatZionUnit1. Some key findings include: (1) relatively brief operation of containment sprays before vessel breach significantly reduces radiological consequences; (2) containment pressure reductions following vesse'l breach should be carefully controlled (preferably with sprays to avoid H burne); and (3) following core uncovering, safety features should 2 be operable before restoring reactor coolant e.akeup. Another study dealing with small-break LDCAs for Zion Unit I revealed that: (1) f an coolers can prevent gross containment f ailure caused by overpressuritation or H burning; (2) partial injection failures do not 2 12n0/81 21 CHAPTER 10 AtmDAL REPORT 7-

p. c a necessarily lead to core melt; and (3) with three or more fan coolers operat-ing, containment sprays are not required. AN4 ze$. C.a y A study L;; '- 7at Oak Ridge tation blackout M Browns g W Ferry Unit gtype. The blackout is assumed to persis-beyond the point of battery exhaustion to core meltdown and subsequent containment failure. The analysis of fission product transport makes up a major part of the study. Hydrogen Proc am j s1 The NRC research program on turdrogen is aimed at better understanding the phenomena associated with hydrogen burns, the methods to prevent / mitigate severe accidents and effects of burns on equipment. In 1981 experiments were conducted to quantify the H / air limits on combustion using igniters similar to 2 those proposed for use in nuclear power plants to control hydrogen. Work was begun to assess the effects of mitigating sieasures (water fogs and foam) in controlling the pressure and temperature of hydrogen burns. This research in 1982 will be expanded to include examination of pre-inerting and oxygen depletion as well as such eitigating schemes as post-accident CO inerting and 2 use of hydrogen getters. The program has been useful in Ifunsing work in assessing the hydrogen control systems for the Sequoyah and Grand Gulf power stations. As part of the regular hydrogen program, analyses have been done for Zion (large dry PWR) and Sequoyah (ice condenser) and are currently planned for Grand Gulf (BWR Mark III) and Surry (subatmospheric PWR). The analytical part of the program e$ will improve understanding of the entire role Q hydrogen in a potential accident. Core Melt Technolotv This program at Sandia V ;ma un.te, aks to develop the technology to quantitatively analyze severe core melts, using a large-capacity melt facility (200 to 500 kg of fuel and structural material). The structure features a complete redesign of the melt crucible and furnace geometry coupled with new temperature sensors to provide reliable spatial temperature distributions within the melt. Ultrasonic thermometry provides several axial temperature measurements within the melt, and a rugged fluid thermometer backs up the ultrasonic 12/10/81 22 CHAPT [R 10 ANNUAL FIPCRT T

4 Yl N .s measurement. Other features of the melt facility include a pressing capability, which enhances melting, spinning, welding, and flame spraying with tungsten for the melt crucitriesj and a crack-detecting technique for ceramic bricts, which are used for core retention designs. At the end of 1981, crucibles and charges were being assembled. The computer program, CORCDN, which is being developed and verified at Sandia, kH 1 model phenomena governing molten core / concrete interaction after } an accident. The first version is operational, although its application is limited to early stages of an accident since only pure molten materials are considerod. Tae behavior of solid or partially solid debris will be included A in a later version. The users' manual (NUREG/CR-214f) has been published. TMI-2 Post-Jccident Examinations The cooperative NRC/DDE/ Electric Power Research Institute / General Public Utilities effort to conduct post-accident examinations of TMI-2 resulted in f . cost-significant efforts outside the reactor in 1981, and in the planning for f examinations of primary system internals and fuel which will occur in subsequent years. (Seep.210,NRC1980AnnualReportg.) About 15 research reports have been prepared on results from some of the i six technical tasks and a seminar was scheduled in December 19EI to discuss Men.s. Arp.m then with industry and utility representatives. g ~ Advanced Safety Technoloey Research g[ g NRC's advanced safety technology research programj sw J ' - - - pp. 207-210, NRC 19BD Annual Report; focuses on liquid metal fast breeder I reactors (LMTERs) and high temperature gas-cooled reactors (HTGRs). Liouid Metal Fest Ereeder Eeactors. Work in 1981 under the LMTER program consisted mainly of projects in analysis, accident threats to the primary system and containment, and aerosol release and transport. Much of this effert continued as described in the 1980 report. Newer developments included: i was s es.ss%. +.~, d 4 - 1. Analysis. The - ..M gf a 4 CDtf4IX-1A code to analysis of the United Kingdom Protetype fast Reactor invessel flow anomalies and invessel analysis of the Fast Flux Test Facility (FFTIK - -- -- Q' &, - +. L n'. Brookhaven National Laboratory's Super System Code (SSC) simulates the hydrasile behavior of an entire nuclear plant. It has been available in 4 12/10/El 21 CHAPTEL to ANNUAL REPORT t.

s ~ s a various forms including %e SSC-L code for loop-type LMTEM Another version, the SSC-P code, for poor.ype LMrERs, was completed and will be documented and ready for general use in 1981. Plant modeling for use with SSC-L was also completed for the Clinch River Breeder and other reactors and validation of SSC-L in 1981 focused on the comparison of calculations of the FFTF tests with the experimental data. The comparisons are good. Plant modeling also was extended in 1981 to a generic steam turbine electrical system model, applicable to LWRs and HTGRs as well as LMTERs. Los Alamos continued work on the SIMER code in 1981 (see p. 234, NRC 1979 Annual Report) with emphasis on verification experiments. % W ss N, y \\ 2. Accident Threat to Primary System.,fhis research/ is 64ee6 44, developiaf - s 4 .a the data and codes ta assess.tt'e impact of severe core-disassembly accidents on M..- %, & ae.

s LMFER primary system (the reactor vessel and piping)--nctably the threats from Mt.

h Q pm.- s.t t 63 energy released in the accident and the heat

*': rM t m

.; core ..g debris. Mo t accident energetics experiments are performed with test reactor fuel irl the Annular Core Research Reactor (ACRR) at Sandia. In 1981, a series of new experiments was started in the ACRR on the, streaming & se.s.;se r: and freezing of molten y a; r aeu cs: sa*s.,s, fuel during the '^"- " transition pt.as -fuel motion which determines the energy release associated with that phase. Preparations also were completed ~~ for new experiments to determine whether a propagating thermal explosion can.. occur with molten reactor fuel and liquid-sodium coolant. If such explosions can occur, they may significantly increase the' damage potential of postulated severe ancidents. The ACRR coded-aperture-imaging diagnostics system (see

p. 202, kRC 1978 Annual Repert) was significantly improved in 1981. This unique diagnostics system produces images of th? displaced test fuel from gamma rays emitted by fission in the test fuel and is used in some of the experiments in ACRR.

A series of unique experiments in ACRR on core-debris coolability 6n 19E8*- k 1%I became a. joint international program with EURATDM and Japa, with the foreign participants carrying most of the program costs. The. fif th experisent of the i i series showed that a stratified debris bed with the finer debris at the top (as j would naturally occur in an accident) as considerably lower coolability lielts than the unstratified beds previously studied. An analytical model of l 01/26/E2 24 CHAF1ER 10 ANNUAL REPDRT 1

i ~ _. - - -... -. .. ~ - - - - 4 debris bed coolability limits that best fits available data is now in general use. It has been used in safety analyses of LWRs for the TMI-2 accident and in the Zion / Indian Point studies. 3. Accident Threat-to Containment. This research addresses the threat to a tantainment from sodium and postaccident core debris that have penetrated the reacter primary system. In both cases, the primary threat is from gas pressure generated by interaction with basemat concrete and not from penetration of the basemat per se. In 1981, tests on these chemical interactions between liquid sodium and different concretes showed that in some circur. stances the reaction can ce quite rapid. Although considerable understanding of the complicated chemistry involved in these interactions has been developed, they are not yet sufficiently understood for reliable prediction. In 1981, work on the Large-Melt Facility (LMF) at Sandia was finished. This facility can produce pours of up to 500 kg (1100 lbI) of molten reactor . fuel onto concrete or similar materials or into reactor coolant. Experiments to expand the data base on core-melt interactions are now possible with the LMF. Development also continued on improved models of the CORCON code for the analysis of core-melt / concrete interactions. Duririg 1981, experiments at Brookhaven National Laboratory provided important data on heat transfer between liquid layers subject to bubbling gas flow, about which little has been knowg) ~ An improved model developed from these results was added to CORCON. In general, NRC concludes that the program of research on sodium / concrete interactions produced major results'in 1981. In addition to the items mentioned above, a new model of concrete attack and ablation (SCAM) was developed, a large-scale test showed tnat the energetic reaction with limestor* concrete could be prolonged by sodium additions, and energetic reactions were initiated in sir.all-scale (lyft dia.) tests by careful balancing heat loss, interface velocity, constraint, and pressure.

4. Aerosol Release and Transport.

During 1981, tests were conducted at Oak Ridge with uranium oxide aerosols and (For other details of this program, see p. 209,$ M' 1980 Annual Report.) steam. High Termerature Cas-Cooled Reactors. For two years, Wt.i budget / MJ As eben.ahed the funds reouested by NRC for gas-cooled reactor research; however, DW MW the present Adreinistrativ has hW.d-52 stillion annually for tFe next 0106/82 25 CHAPTER 10 ANWAL REPORT o

4 several years. As repcrted in 1980, plans had been made to curtail or discontinue some projects, but Congress identified certain funds and specified certain programs that were not to be tersinated. At NRC's request, in the event of project terminations, the national laboratories prepared summarit; of all the research work up to and including 1980, and these will be available early in 1982. Some programs of importance to the fort Saint vrain reactor in Colorado were continued in skeletal form at several national laborator ANALYTICAL MODEL5 Computer codes, as defined on page 205g1980 Annual Report, are desig to assist in the resolution of ifcensing issues. In 19El, the fo110 wing codes were completed avj released: (1) TRAC-PFI, used primarily in the analysis of small-break LDCAs in PWRs, and certain non-LOCA transients and accidents; (2) TRAC-BDI, for analysis of a variety of both LOCA and non-LDCA transients and accidents in Be'Rs; (3) COBRA /TRAKu$ed to analyze LOCAs in Westingh that feature the upper headyinjection form of the emergency cooling system; and g (4) RELAP-5/CD1, for one-dimensional analysis of LWR accidents and transients. Plans for 1982 include completion of the PWR version of the code. Other work in 1981 included ef forts (estimated for completion in 1982) tDward adaptation of the COBRA-TF subchannel code to LWR containment subcompartment 1 cad analysis, and the initiation of work to adapt an existing ~ multidimensional code (50LA-3D) to analysis of hydrogen transport and distribution in LWR containments. Independent code assessments of the TRAC-PD2 and RELAP-5/ODI codes indicated that TRAC-PD2 is much more accurate and reliable than its predecessor, TRAC-P1A. The RELAP-5MDDI code is so new that not enough information concerning its predictive capabilities could be assembled. TRAC and RELAP codes werY used extensively in severe accident seguence analyses and studies of pumps on/off conseQuencesp sf.n.u-w s. w y V .,,Ji ~ Y, mete =ee licensing issues, such as overcooling tran ients and station blackout. RISK ANALYSIS de " M T AC-El d RISK METHDDDLOGY DEVELOPMENT gg h p..;;/enoo t o.e b ""b gg

y u te b e,wlvka yo.

01/26/B2 26 CHAP!ER ID ANNUAL REPORT

s c ~ ~ 1 I N .-.---,a ~~ "The. In NRC's development of methodology for probabilistic risk analysis continued in 1981 with special emphasis on safety goals for nuclear power plants. Prioritycontinuedonthedevelopmentofformalizeddeejifonmaking n icR.d > wsgg approaches (using risk analysis). in licensing and inspections. T "- - " -- aeltak;h4 l produced met. (and some sof tware) for evaluating time-dependent h I

**U*T'd b-: .. +; u, p Vgamon-cause failure probabilities, 2nd estimating flood probabilities and risks, as well as for analyzing component reliability data.

m%ede le e w A total of 11 risk assessmentguREGs Wre published in 1981 (see Appendix 7). REACTOR RISK Anticipated Transients Without Scram The Commission voted on June 16, 1981, to issue two proposed alterr.ative rules on Anticipated Transients Without Scram (ATVS) for public commentfnw . Eederal-4eeMM 0ne would establish design requirements to reduce the likelihood and mitigate the consequences of ATV5 events. The other would less endansbr. require licensee reliability assurance programs and 4ertagdesign changes. gggo m b W e c - -t: F 'll b; ::::= d-ef4M9-dey-periodA h ulf ' *.

ME C e stb ed d @.c Ldl e &se ts aM +%e A fM. C de M i %%

sk [AhM L pt:c co .~d 6, peaCter Accident Consecuence AnalysisAca;l e

M;a'

~ a. t*... A A K h, a c owh a.; In.1981, NRC released the Calculations of Reactor Accident Consequence-2. (CRAC-2) Model, featuring significar.t improvements over the original CEAC model in emergency response modeling capabilities and meteorological dispersion modeling techniques (see p. 219, NRC 1980 Annual Repert). Studies were initiated to review and revise as necessary the health effects models used in j 63'et sk=eA<a ' p the 1974 AEC Reactor Safety Study (WASH-1400f. pome 30 orgaf zations representing 16 countries W in an international comparison of consecuence models sponsored by the MEfa#M (Orgar.'zation for Economic Cooperation and Ih Development / Nuclear Energy Agency -Committee on the Safety of Nuclear A Installations. Emeroency Planninn At the request of the Federal Emergency Management Agency, NRC undertook studies in 1981 to (1) quantify the potential benefits of household items such 01/26/82 27 CHAPTER 10 AtmUAL REPORT m. B

m. - - - - - - ~ ~ -. - - - N as towels, sheets, shirts, and handkerchiefs as filters to protect the respiratory system and (2) assess the relative worth of various protective I actions in different reactor accidents, TaA? @ cay ' heat Removal {dee.UTS pe A Preliminary results of NRC research on alternative decay heat removal concepts for Ifght water reactors were published (NUREG/CR-1556) in April 1981. The research includes studies of current decay heat removal systems and the des'gn criteria used in both U.S. and non-U.S. light water reactors. The ' report sets forth various concepts to increase the reliability of the decay heat removal function for further consideration by industry and licensing p.aJ.34 Y. # #, ~~~ Y

d

.'s e q t Ag'f" authorities. A a, cs w :fe /.., Mg or ..w. hh eactor $vstems Analyris and Licensino Support Work continued on the Reactor Safety Study Methodology Applicatiens b ac. e 4 n e m. Program (see p. 219. NRC 1980 Annual Repcrt uth - - it

at'A 1

clumes : T. c' ' of NUREG/CR-1 59 t,:,d i. n eli % ned taIm. which discuss the four plants studied, p a p ill be published m in ' __.,,= 1982. w A Work on Phase I of the Interim Reliability Evaluation P'rogram will be completed by early 19E2, when NRC expects to publish results for each of the p.An k four plants studied. (See,WRC,1980 Annual Report,for program description.) The NRC prowice fir.ancial assistance to the Institute of Electrical and ~ Electronics Engineers and the American Nuclear Society to cocrt.inate develop- \\,s NmeEt of a procedurefCR 2.h was guide for probab lis i l hanabsisofsafetyofnuclearek maw /WI. JUM c.4.en et-IMEG 1 a ;; i)ower plants.{@Qll'be available in mid-1982.~( ede y. Development has completed on two computer codes to model the physical 5 b processes of core meltdown accideng. The MARCH code (see NRC 1980 Annual g Report) n.as released in late 1980; [the CORRAL code, used in concert with

[public We-bt! vde i *Y te !a.s s^ e.

MARCH i 3 E2. Research has supported activities of the Office for Analysis and anin s Evaluation of Operational Data by providing benchmarkgesults cA Licensee Event Reports (LERs) in the Accident Seguente Precursor Program. This program has resulted in a screening of all LERs (some 22,000) f rom 1969 through 1980 to I identify precursors of significance to core damage. Analysisofsignificant als sn4eb ans%n neiidil;4 JL,- * -+ t 'i h(Inn 4 Ly 'is one nt,a ct er 4 fa ey e n"'s. trend d n* ; / a S 4' N w.t7 hg nd me ' N e ed **on. d)

c.M.

c.d

bM k 02/03/61 28 CHAPTER 10 ANNUAL RIPORT Tl i

I

mm m- -.ow, m.m I ,1 e / i 's ~ ') SERT Alternative Containnent Concepts l ition to the investigatian of alternative decay heat renoval concepts. studies also continued to examine the merits cf alternative containment concepts, especially filter-vent containment systems (FVCS) and molten core retention devices. Final reports on the risk-reduction benefit and costs of the former are expected in the latter half of I fiscal year 1982. A report on the risk-reduction potential of the latter was issued in 1981 (NUREG/CR-2155). In 1981. wort was begun to merge tSese two programs with the alternative decay heat removal concepts e i I program. The single resulting program is systematically investigating the risk-reduction benefits and costs of these concepts (and combinations ? ) of them) along with other concepts. The report of the first semiquantitative analysis of these concepts is expected in the sumer of 1982. i I I i g ~ I t i: 1-f 5 ~ f ~ .g e .r~ i e 9 e = h I s

o. .L % [4mewe ka% ] (400ad4. quantitative accuracy for forecasting the likelihood and the j improved topology of core damage accidents. The program is also indicating the nature used M of various multiple failure scenarios being experienced that can be "

m., -;' upgraded operator training, plant design, and licensing safdy review.

j Alternative Containment Concerts t the merit of vented, filtered containment designs is Investigation e the results and any needed follow-on analysis will be merged winding down into a new program to support NRC's work n -,am., s' -; d e-b define the relative merit of a broader set of possible plant

. ;r modifications for improved safety.

'N w RISK METH3DOLDGY DEVELOPMENT Ir The NRC's development of methodology for probabilistic risk analysis continued in 1981 with special emphasis on safety goals for nuclear power Priority continued on the development of formalized decision \\ plants. These efforts approaches (using risk analysis) in licensing and inspectioits. I produced methods end software for evaluating time-dependent failure effects, t Quantifying common-cause f ailure protabilities, and estimating flood occurrence probabilities and risks as well as for analy7ing component reliebility data. / Eleven risk assessment NUREGs were published in 1981 (See Appendix 7).- L Transportation Safety Researca The transportation safety research program focused on two main issues: eveloped $b e4) to deterrine if mode-dependent transportation regulat (1 to ead improwbag the technical basis for protecting public health and safet and (2) to establish a data base for assessing the potential consequences of explosive attacks on irradiated fuel shipping containers. The intent of the first program is to establish package performance tasts for severe accidents and to combine these requirements with an appropriate set Testing of road and rail transport packages 4 Met of post-test acceptance standards. l g,anc (baatu fed-S is plannec 1+te-MM r early 19E,3. A similar process to assess air and siarine i A J transpert modes began late in the year. C% APTER 10 ANNUAL REPORT 29 il

n a e e \\ 0 N JYl The secon.1 prograrphich characterized the radiologhal releases resulting T .keed.d ansiast fromspecifickindsofexplosiN rradiated fuel shipping casks, has been completed. This program included ieveral *first of a kind" experiments, some of which were carried out in the experi. mental configuration shown on page _. (Figure to be supolied f using this configuration, the effects of a shared charge attack on irradiated fuel were assessed. A flash x-ray shoMng passepe of the explosively formed " jet" through a row of fuel pins is shown on page _. (Figure to.a supplied.) The results from this program have f indicated that the effects of explosives on irradiated fuel are less than had been previously assumed. NRC decisions on safeguards meas;res required for irradiated fuel shipments are being reviewed in light of these results. (A comprehensive discussion of transportation regulation, including regulatory standards and guides, appears in Chapter 4.)

Apphch.as i

Radioisotopel *::.- n' I' ag t.Mi e a s II utteie s I NRC *e+++e in radioisotope m p:: -- included work in the following areas: ?' incandescent Ces Mantles. Investigation of the potential radiation doses M de otwunl P*Wc. ~ Wom incandescent gas mantles impregnated witt, thorium compounds continued I. b _1 ,i J L c u s gh 1961.,Thjs study is part of several dealing with radiation doses. Ge8Ja** throu .e g a c:es

c. ta; a t w *** lese:W n m ess;ls.

Instrument (.alibration Sources. For many years Commissien regulatit,ns have exempted usa e' a small source in a radiation-neasuring istrument. In 1981, that exemption was expanoed to permit users to obtain instruments with severalsourcesofdifferent(radionuclidesaswellasmultipledetectorsona single instrument. The changes permit faster and more reliable measurements. Contaminoted Smelted A11ovs. In 1981 NRC continued to accept public & u g me d c.e,ntamia.iwi tr

e a m *'* -. 44 *Oen comments on proposed amendments dealing with technetium 091 (see p. 195, 1980 Annual Report). More than 3600 letters, postcards, and telegrams had been received at year's end.

Well-Loooino Sources. In 1981 NRC completed its assessment of rists in 9 insch.sileva'n'.e. wdl.in6. 6eus ce s f reopening wells containing = r '..:n . see p. 195, / e.x 1980 Annual Report)) and concluded that the^Ieches reduction in fadiological risk does ,,u. e4 tee. tau:ualy not warrantg : '...Qcposed procedures 0 at udondt well-logging 3 b8 NE """* sources. Thus, no regulatory action will be reovired 02/03/81 30 ChMTER 10 ANNUAL REPORT 7

o o [ ~_. . ~.. AN s'e n en[fk C.eg ed b ACE h.% tag s PL 7,-. Fuel Cycle Risk Assessment I hRC's development of methodologies to assess risks from nuclear fuel cycle activities, other than reactors. cor.tinued in 1981. The development and demonstration of the high-level umste (HLW) risk assessment methodology continued on the bedded salt reference repository s..e. Similar methodology Work ses initiated is being developed for preclosure an. spent fuel isolation. to expand the HLW risk assessment methodology to other geologic media including The Interoffice Waste Management basalt. welded tuff, domed saignd granite. Modeling Group (IbMG) (see 1979 and 1980 Annual Reports) continued gaining experience in applying the HLW risk assessment methodology by exercising problems on the geosphere transport, biosphere transport, dosimetry and health effect nd statistical codes. An IlNG Program Plan was formulated outlining l the steps for developing expertise on the application of the HLW risk JLo assessment methodology. More than tenamey NLfREG/CR reports and technical articles have been publi;hed since the program began in 1976. An independent The technical review group continued its review of the published products. Fuel Cycle Risk Assessment program was initiated to scope the risks from all elements of the nuclear fuel cyIle and to develop risk assessment methodologies L e d a.1. 40 t.*... .. w"... for the high risk elements or elements amee need immediate risk tool @Q REGULATORY ANALYSIS ~ kM e.%u MM ....:.. m Regulatory analysis is p e w to Ae R;'s regulatory +rocess "em ad c.c.- em,, coherent, and understancabigg Toward this end, a number of 4 ctivities are being pursued, including the 5 of h M ations te -'. development of procedures and methodologies to ) identify the costs and bew!its of e proposed e unneeeeseey buroent the periodic review of existing lregulatoryactio regulations, and be implementation of procedures inat MMeelopme to comply with statutory requirements in this area / such as the Paperwork Reduction Act (P.L. 96-511) and the Regulatory Flexibility Act (P.L. 96-354). MCEaf h".'d"Rulemaking U R ~" S 4 e m..L en As an outgrowth of TMI-2 accident studies, the NRC is tmtisting / rulemaking to consider to what extent, if any, nuclear power plants should be designed to deal efisctively with degraded core and core melt accidents and to sitigate the conseguences thereof'. An advance notice of proposed rulemaking 02/04/82 31 CHAPTER 10 ANNUAi REPORT v e l

.- e p e N ~. - - - - - - was putlished in the Federal Secister in October 1980 to solicit public comments on several Questions related to the development of the rule. In a related action, the NRC has developed an interim rule to improve hydrogen management in some light-water reactcrs and to provide specific design and' cther requirements to mitigate the consequences of accidents re.ulting in a degraded core. A notice of proposed rulemaking on this interim rule was published in October 1980. The sections of this interim rule relating to j design consicerations to mitigate degraded core accidents were later ) incorporated into a proposed rule for' operator license applicants published in i i May 1981. A final rule on hydrogen control in Mark I and II boiling water reactors (BWRs) was published in October 1981. This rule requires the inerting of these reactors and also requires bydrogen recombiner capability for plants that previously relied on venting. Currently.uncer development is a propcsed rule to require hydrogen control syster.s for BWRs with Mark-III-type containments and for p tssurized water reactors (NRs) with ice-condenser-type containments and to establish specific criteria for equipment survivability during a bydrogen burn. - 1. the A ) FACILITY OFERATIONS ~ HUMAN FACTORS ME } The objective of h factors research is to provide a technical basis to m {dy support reguptory needs -(n applying human factors engineering to nuclear

yQdo, o63 facilities. Ewpta.e% incluoe human performance data, analytical methods,

.25 4 t assessment of new concepts, and design and evaluation criteria. -Data to ,3 g u*7*. support the implementation of improvements in the operator / machine inte{ ace are n-- eD%" especially needed, as well as improved Quantitative estimates cf human c--m e u s e, ,,, g reliability to helpleduce large uncertainties in risk analyses. v 3g - e-g - s uw E6E3 NRC's human factors research activities was consolicated in Mwe% / 1981 into a program addressing human f acters systems engineering, human = reliability, plant procedures itensee Qualificatio

, ;- A

,UfY w g.;,; g**e h r

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.,a human factors review of nuclear i

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) J f0 E E Q T t b RESEAR"H TO IMPROVE REACTOR $ATETY% A plan for research to improve reactor safety was described in April 1978 im NUREG-0438, a retsrt to Congress. The report called for t $183 million of effort spread over a 3+ year period to begin NRC t4 search on (w roved eJ ety. f This work had just begun when the Three Mile Island accident pve far greater emphasis to this area of improved reactor tnfety. In 1981, this research included projects in the areas cf alternative detpy heat removalconcepts,alternatedontainmentconcepts.humanfactors.and instrumentation and control. These areas are described in other parts 1 of this chapter on research. L In the future we plan to drop the designation 'Research to Improve Reactor Safety" becausegince trie TMI accidenj in a broad sense, much of MRC's work, particularly the research program, has been directed towar mprcving reactor safety. The [mproved [afety [ystems /esearch program has served its purpose of starting work in this direction. i i 4 t a m n,. k J. b ~*. g

o + - s i -= iY, behdeedtsAck 01cm ed, NUREG/'8-2147]( - - % SelSots, S ~ - rgs -"~ power plant alars. c: r-affecting pump and valve operability, using nalyses of human error n__

t ent Report (LER) data and hun.an error prediction models,..-

Licensee wee

=

as - {. f! .';---d- ' q(NURES/CR-1279 and NJREG/CR-JEBO) P A survey of A ,M. Ma AC Ca. a s mem ~' d Ib *oeti&W **T1og S

- ", and practices elated to operator 3m q _ ~- e-.<,_- eNi rA1G - p% N.

' '. E

l selection, training, and utilization was completedg$hattended

'd q

'~s t ge e..

Over 150 nuclear engineers and human factoes? e_n~ - - b iae / hw tt &BE A. a.f hRC-sponsored weeka ear workshop on hum factors standards and safety. A N. Ic Tr _r -e pecorsed Revision 2 to Guide 1.8 on personnel qualificat 8 nand training,pEendorses American Nuclear Society 5tandard ANS 3.1,[tas ,; issuedforpubliccommentinOckber198g ice 1.149 ar powe 6m. ~ U 7', J1 ant simulator $ for use in operator trainingi~ch endorses American Nuclear . E E 'I I

n ey $ociety Standard ANSI /ANS 3.919Elg% issued in April ;M1.

E I. E 5 }

"3 :' ;

y IQuality Assurance In this fiscal year 1981, NRC developed proposed regulations and regulstory guides accressing quality assuann (QA) criteria for the dispesai of high-level radicactise wastes in geologic repositories; reporting changes to QA programs f ar nuclear power plants; and updating QA guidance for the oesigr., construction, and operation of nuclear power plants, with completion scheduled for fiscal year ?.9E* Prelieinary plans are under way to begin research in fiscal year 1982 to better deterziae those nuclear sw plant structures, ~ systems, and components coMdered important to safety and to develop a L: -mat o s.. methodology for applying the QA progray, requirements in a graded manner. A% m.e - proposed rule will be publisher' in ++s%2962 to clarify the relationship between Appendices A and B to 10 CFR Part 50 for the application cf QA requirements to nuclear power structures, systems, and components with the effective rule scheduled for late =? 3 2-1982. Additionally, efforts ~ are under way to endorse the Irgtute of Electronic and Electrical Engineers e program for accreditation of laboratories conducting qualification testing. A proposed Revision 3 to, Guide 1.28, on QA program requirements during design and construction, was issued for comment in March 19E1. Proposed rulemaking concerning reporting changes to QA programs for nuclear power plants mas published in the Federal Recir,tg on July 2,1981. 1 Quality assurance criteria were developed for proposed Part 60, on the C2/04/E2 33 CHAPTER 10 ANNUAL REPORT T,

g - - - _.. _ _ _. _. _ ,d i-technical criteria f or the disposal of high-level radicactive wastes a geologic repositories, which was also published for comment in July 19E1. t Emerceney preparedness NRC research and standards activities within the emergency preparedness _ L- ~ area have concentrated on the following projects: 'O q (2) upgrading of emergency preparedness regulations for cer a. g fuel cycle and material license the upgrading or clarificatien of ropriate emergency preparedness regulatory guides and regulationg w.d marning system capabilities NRC is now N U*f' EN= ..m y, g;--. ..-,...;.. 7 m in help establish u detailed criteria for implementing the emergency.creparednesg.M,s40ed

p t 5 % In September 1981, the Commission

% i.tM; M; A published in the Feceral Register a proposed rule change that would celay for one year the cate for prov ding the cepability for prompt public notification. .J A Atsaci L "': dtfsiculties and uncertainties regarding This delay was e64-designing, prcturing, and installing appropriate warning syster.s. The Commission published in the Feceral Recister on June 3,19E1,

n Advance Notice of Propcsed Rulemaking (f.6 FR 29712) announcing that consideration is being given to specifying strengthened emergency preparedness requirements for those fuel cycle and materials licensees having the pctertialW Publication cf as for accigents that could thrtalen public health and safety.

~ g h it. (> e.e s ts t =e44M4we@t+es. is expected in 1982. In parallel vith upgrading the regulations on emergency prepareoness, the ~ A a **. s. staf f is upgrading appropriate regulatory guides to correspond tofreculatiorfs % Guide 1.101, on emergency planning for nuclear power plants, was published in September 1981. INSTRUMENTATION AND CONTRDL L AAfdk j I As part of the NRC research pregram at the Dak R10ge National laboratory (DRNL) on noise surveillante and diagncstic techniques, the study on use of noise analysis methods for cetecting, locating, and characterizing loose parts e in nuclear power plants was completed. This study assisted in ceveloping 34 CHAPTER 10 ANNUAL REPDET 02/04/82 l \\ f

F ab, k ] r -m Lprar 2d 1, 4 i l The etdective of the NRC research and standards development program in aba fnstrumentationandf.ont 1 emma is to provide the technical bases to support - a c.- w the regulatory progra. for operating plants as well as these under Itcensing g re view. The r8esrch effcrt prie,arily consist of the development of surveillance and disg'.ostic techniques. including noise ann?ysis methods. ~~ +. evaluation of instruments for following the course of an accideit, assessment cf instrument components under severe environmental conditions, ani the initiation of a program on the safety implications of control systrs. The standards develcpment effort primarily consisted of the issuance of a revision to Regulatory Guide 1.g7 and continuation of work on standards and a regulatory puide for the qualification of electrical equipment in nuclear power plants. _L ear aA p q u y. T

p-s ~ _... _ _....... - _ b t i ANSI /ANp.5-1980,*CriteriaforAccidentMonitoringFunctionsin LigSt-Water-Cooled Reactors," is encorsed by Guide 1.97. Work is continuing on evaluatinD the adecuacy and effectiveness of this guide and standard, and revisions to the guice will be issued when considered necessary. Work continued on standards and guices for the qualification of electrical equipment in nuclear power plants. A draft guide on qualification testing of cable penetration fire stops is under review by user groups before being issued as an active guide. Proposed Revision 1 to Guide 1.131 on qualification testing of electric cables and splices l's also undergoing final review. (be 3 p.184, NRC 1980 Annual Repert.) DCCUFATIONAL RADIATION PROTECTION Health Physics Measurement _s During fiscal year 1981, research and standards development in improving health physics measurements required to protect workers from radiation centered O on upgrading personnel cosimetry programs, developing and testing health physics survey instrument per'ormance standard m# :-' --; :-d testig '% 2:- ~ A oeweloping and b't. Wp iu w uy n.:, b,.im ge-feero m W w testing performance standardy for bicastgy laboratories. The Health Physics A Society Standards Committee and the, Amyr.ican hational Standards Institute htv-develoged draf t standards, with KRC participation, fo* the performance of health physics survey instruments and bicassay laboratories. A technical %r* assistance contract, te+e jointly funded and managed by NRC and DDE, was established to test the standards for applicability to the radiation protection programs of both agencies. Work continued on a program for the accreditation of personnel dosimetry processors who provide devices used to measure the radiatien doses received by workers in NRC-licensed activities. Plans were made for additional testing of processors against a revised A.':51 performance standard, and site visits to 36 dosimetryprocessorswereconductedtodeterminereasonsforhoor lier performance. (See 1979 and 1980 NRC Annual Reports for results o* earlier b tests.) W final round of tests will provide assurance that the standard, as revised, is an appropriate basis,for accreditation. 02/04/B2 36 CHAPTER 10 ANNUAL U POR1

[ l .AO Guide 1.133 on the locse part cetection program for the prie.ary system of LWRs. An on-line neutron noise surveillance and diagncstic cemonstration system with continuous measurement capability was installed at the Sequeyah Unit I reactor and has been gathering signature data since & P__ _2 April 19E1. Abnorr.a1 operating conditions noise data were ottained as part of LDFT and Semiscale tests d are being used in assessing the feasibility of using pressure, neutron, and temperature noise to detect anomalies at power plants. Nuclear power plant instrumentation performance will be evaluated in a new program by the Idaho National Engineering Laboratories, using criteria in Guide 1.97 that defines the instrumentation recommentations for following the course of an accident. Sandia M t:n:! L t :*/ " 'S' conduct a series of instrument component assessments focused on identifying degradation and failure modes of instruments and electrical equipment important to safety under design tasis accident conditions. This research is intended to improve Quality assurance guidelines for the design, installation, and n.aintenance of instrumentation and other electric equipment importagdt t, safety. . ~, In another study started at Mgclear plant alars and annunciater systems will be evaluated to confirm their adecuacy and to assess the feasibility of setting priorities for the required operator responses. A new progras at CRNL has begun to study the safety implications of ~~) ~ conteel systems and rela'.ed plant dyncaics. W peccident stauences that U s.sy be outside the design basis envelope assumed for all plants will be idantified and studied. A methodology for assessing the failure modes and effects of contr.ol systems on the basis of common cause, common mode, and cther multiple failures such as casgade failures will be developed. $s.n o.* w A related program at est was also initiated in fiscal year 19El to oevelop methods for assessing the adequacy of nuclear power plant electrical systerrs with regard to system interactions (particularly with control systems) and cascaded failures. k w . _ -m. Revision 2 to Guide 1.97, on instrumentation for light-water-cooled nuclear power plants to assess ptant and environs conditions curing and following an accident, was issued as an active guide in December 1980. 02/04/E2 35 CHAPTER 10 ANNUAL E{PDRT 'T

p A I Guide E.28, on the selection and use of ascible alarm cosiaeters, was published in September 19E1. It provides information on acceptable uses of warning dosimeters and limitations on their use. Radiation Protection Trairino Guide 8.27, on radiation protection training of workers at light water-cooled reactors, was issued in April 19E1, and Guide B.29, providing instruction on risks from occupatior.al radiation exposure, was issued in July n V V 1981. The Igtter guidgV written in a question and-answer forn.a+p presents material acceptable to the ERC staff to satisfy requirements for biological risk training. A safety traini9g s nual for radiographers entitled " Working Safely in W r.4 Gamma Radiography" i+-4. sag prepered as a text for use in training industrial radiographers. Respiratory Protection In 1981, the ERC completed two vicectape/ training manual units on the proper use of air-purifying restf tors and atmosphere-supplying respiratorse Ceakon A ma dand released a third unit on cleaninkmaintenanteg*and stcrage of respirato Y pWglating respiratory protection to emergency preparedness.as initiated. 1.* g Ptiysics Survevs Revision 1 to Guice E.23. on radiation safety surveys at medical institutions, was issued in January 19E1. The guise describes acceptable methods for implementing and enducting radiation survey programs for medical licensees, licensineGuidag The NRC staff provi45 guidance on reovirements fc+ applicaticas fer r' . various types of licenses iur the use of radioactive raterials. Two accitional guides in this series were iss'ued: Revision 3 to Guide 10.8, a guice for the preparation of applications for medical programs, in October 1980, and Revision 1 to Guide 10.5, guidance for appilcants for type A licenses of broad 1 scope, ** 2nuary 1981. 02/03/E1 37 CHAPTER 10 ANNUAL MPORT l

f.

i

p c l i__ _,_ l 5AFEGUARDS RESEARCH i During 19E1, the NRC continued to work on computer-based evaluation methodologies in support of the domestic safeguards licensing proDram. The agency also supported studies to optisi2e defense strategies to analyze insicer eV movements required to sabotage vital components / and to provide computational support for vital area identification. A safeguards measurement handbook (NUREG/CR-2078) pubitshed during 1981 describes the capabilities and limitations of nuclear material measurement methods currently in use, and a >A3 description of the principle.s, problemsg nd applicatior of nondestructive assay V ,2 4A t-measurements we gublished in NUREG/CR-0602. Several es regulatory guides a J suMed e4.;.whuctivt 4%cw q (ce se.cl i ~

trc:. '- U.: ;;

' ::' ;.; ::/c.:, ; ;.: wfere p ta.64+d in 19EJ,. Chapter 9- _ _ ....vt-a ~ t k 4afegwa % P - - M.s.'n s, c h', ko a s, k d*=. L o u ^ co. c-a M g g{e y ads M t=% ' a. 4 p gnam.) k m Statistical Treatment of Accountability Data Considerable emphasis was given to the treatment of dataf<,a accountability of special nu' clear material in 1981. A contracter study of criteria for bias correctiMalternatives (NUREG/CR-2205) was completed, and ancther on the u seal effects that certain : ' ' ' dw - -; assumptions can have on the calculation cf the tek p R variances of inventory differences was,p;'p. e d'NUREG/CR-197Q g A New methods for verifying an inventory through the use cf statistical sampling plans were being investigated in 1981, and efforts to develop a M4 atherA statistical test powerful enough to cetect diversion oser-$everal loss scenaries were pursued. The project reporf will be published in 19E2. Several projects to improve the statistical treatment of accountabihts data were started in 19El, including a three year effort to update and improve statistical methods for nuclear material accourta,bility, a new contractor study g aimecE*.>[n analysis cf to develop procedures to resolve shipper / receive ees d cumulative shipper / receiver differences which will help cetermine when problems exist in corrrspondent shipper / receiver accounts over multiple shigtunts. J 3 s e C2/04/E2 3B CHAPTER 10 ANNUAL F[PCET j n e

. _ _. -.. - ~. - - - HEALTH, SITING AND M STE HA MCEMINT j SITlNI, AND [WVlRDNMEET The siting and environmental program covers research and standards 4 regarding the siting of nuclear facilities, the assessment of environmental impacts from the construction and operation of thefe f agflities, and the & Mn. e k ms evaluation of the environmental pathways for radioactive materialg. Activities 4 in this area during 19El incluced the following: Site Sa'ety Tsu A.cel s w o*(i M b Thimon-etetqhe rulemaking, en Reactor Siting Criteria (see uu, u.e6ew e.d;, p g NRC 1980 Annual tenert, p. 180) pi+Meynggtica of r; tert to prepare an es?. ew..:

en ironmental iscact statement in wccmber 1980. "-' - +'

' Honimenth ^

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, popea% m,,, l rn.,.., s *. gejhk v u en en,;:w.m.ci m..^,m. .-t we e % - - - d.. -h_ ut ,,.,,r.,,, o dt! be The nem regulations esa oesigned to establish cuantitative demographic g M-criteria for propcsed sites for nuclear power plarts. Mcstyro,isions or the y Cornission's present siting rega tions/ (e.g., to consider seismicity naar the a / site) will t,e retained. /' &t 0 l p, D Ervieonmental Radiation Standards The NRC issued a final rule amending its radiation prctection standarcs to incorpcrate the environmental radiation standards cf,the Environmental Protection Agency (EFA) fer the uranium fuel cycle and,,in a related action, cenied a petition from the American Mining Congress (AMC) to stay the implementation cf these standards to uranium tilling operations. The AMC has also petitioned EFA for a review of the standards. On May 22, 1921, the AMC and the Kerr McGee Corporation filed separtte ' suits against the ERC in the Dr.ited States Court of Appeals for the 10th Circuit for a stay of the implementation cate. On July 17, 1981, the court consolicated the cases and held them in abeyance pending further aaministrative L proceedings before the EPA. 02/03/El 39 Ch@TER 10 ANNUAL K'CET v i

l __ sU l 1 Socioeconomic Icart Evaluation A report (NUREG/CR-20E3) was published on the isyatt of the TMI-2 a cice / on residential property values in the vicinity of Middletown. Fa. rt 5 m::deling syster to predict the ograph{cimpct f plant construction on the n.- -~*s "A loca1' community was developed- ". NJREG/CR-200 D activities in this area incluoed work to revise the CONCEFT/DM DST code which estis.ates capital and nonfuel operating costs of nuclear power plants by incorperating tqe ef fects of new THI-related safety regulation >r h A isaa !=di.4ms ~ $ research project to analy2e pestlicensing populgtion censity and land use g % 4 11 changes around nuclear power plant sitegen gid in oeweloping methods of i forecasting small-area demographic and land use changes. (4m b g -. gu;r ae r- . q rg%%% b PN 5even research reports, published in 1981 by the College of Fisheries of f the University of bashingte provide measurements of radionuclide distribution coefficients in acuatic ecosystems. Onsfivgvolumer-port (NUpIG/CR-IE52) b + I-k "^^- studies of coversthemethodolcta% m;.u.A Nandu : ..c.o,o -. a surernen . k., tesiur137, strontium-L5, plutonium ano americiun, and curium g The other.. p .p \\-Y two volumes (NURE0/CR-1E53) deal with the effects of organic compounds on radionuclide uptake by sediments and with the distribution of racioni.clices among suspended sediments, phytoplank;on, organic dotritus, and filtered sea. ster. A study by the Woods Hole Oceanographic Institution of tSe behavior in a s.arine environment of transuranic racionuclides released from nuclear power plants was published as NUREG/CR-1658. The behavior and release of iron-55, cobalt-60, cesium-134, and cesium-137 also were studied. i Battelle Facific herthwest Laboratory published a critical review of sedimert and radionuclide transport models, water-guality mathematical l ( modeling, and radionuclide adsorption / desorption mechanisms (NUREC/CR-1322) Radionuclide Urtake in Acro-Ecosystems At the Savannah River Ecology Laboratory, agricultural scientists studied radionuclide uptake by plants prom in soil contaminated for 25 years by airborne ef fluents from a fuel re'proces,ing facility. The chemical forms of the C2/04/82 40 CHAPTER 10 ANNUAL R[ PORT l'

pp I -~ R radionuclices, products of natural weathering processes. present a unicue oppercunity for investigation of conditions that eight be associated with the { act.idental releases., Preliminary results of analyses of wheat and soybeans inditated that americium and curium are much more readily taken po through plant roots than plutonium. Research continued at year's end on the urtake of s'.her radionuclides and other important edible plant species as wen as the effects cf notsei spricultural practices and soil treatments on radionuclide uptake. act.atic Ecoine' cal Impact $tudies ' %e b[e.ws ? WRC't, b4Wtic ecological is@act research program covered a mide variety of activities in 1981, including the following: Rec.ults of copper tc.vicity tests at the Lawrence Livermore hational Laboratory mith various life stages of the Facific oyster and carp (NUREC/CR-0747, -1086, and -10E9) indicate that little, if any, vffect would I , cesult from the levels of copper se.isured duritig operations of a power station, but that highe pulsed releases (e.g., during a startup) s.ay cause more signir. cant imracts. Cak Ridge kational Laboratory (DRNL) efferts to provide' better tools for assessing the inGacts cf cooling system operations on fisher'es produced two reportsp t C y u _ y. er', - gon statis,tical methods for analyzing (tWfN;Mt :M & ax.a _ e r2 ~ stock-recruitment relationships (elA % t E. 4 j Q stimates cf M ;-- ;-. /q - y ), entrainment mortality cf ichthyoplankt Felated work at ORNL proviced infors.ation for analyzing impacts en the threadfin shad when nuclear generating -4 facilities are sited on reserveirs (NJREG/CP-1043)., e,a Ee. e.c. ele.*c.16g c,-s 4 maca-as 4RI c o Other studies done undir NEC contract dealt wit]the entrainment of Zooplankton at operating nuclear pont stations (hew York e University--NUREC/CR-2091)N

Athe usefulness and validity cf f,isheries mocels fct Ana it,1 4.',,. 7 e impact assessment (University cf Washington--NUREG/CR-2016);gt.k,*me. -

eb;td 4Ie.tse.5 -4.yg*eewr+ n nuclear generating station ef fluents (Facific Nortt@Ost g Laboratory-+NUREG/CR-0892 en chronic entorine toxicity tests with rainbow trout, NUREG/CR-1297 on bronofore toxicity tests with various a.arier organisms, N 4 dd., h wunac. c i and NUREG/CR-1299 on halogent.ted byproducifs in 4 .}Qeispect I of nuclear power station geration on the ocener nce f.athogenic amoeba in Ev =r. k

c. e rrn su t we cooling tower water, (ORNL--NUREG/CR-1761% inel+eet% uw W-4m 7

de is. d atb +psbMr e dh, s.,6c e.n c J c 4 A, sa.m,( 4ww md adb. Non< Ca' ndad. c, g.. . m u 02/03/t1 41 CHAPliR 10 ANKUAL REPORT e j W

? s AY \\ ~~ (T y f T N k'-n .... :n Of the use of subsurface redor techniQuet for surveyiraj ow-level nuclear waste disposal sites (Geo-CeDters Inc.)_ demonstrated that the radar techniques can detect objects, anomalle and trench boundaries to depths of from one to 30 meters. HEALTH EFFECTS RESEARCH Projeds and results of NRC activity in health effects research incluoed in 1981: [ A report on the health status of former thorium workers (NUREG/CR-1420) revealing that the causes of death of 511 workers showed little association with thorium exposure d that thorium ceposition was detected in 131 of the 194 living persons examined. The wo'rkers with highest exposures are receiving follo p medical examinations. (

2. A new research program to improve estimates of risk from neutron exposures i

I at occupational dose levels, in which large populations of mice were exposed to pure fission neutrons or pure gamma rays at doses comparable to the pemissible occupational limits. Both somatic and genetic ef fects were being evaluated at the end f the year. (*Mti. A - 1 1 W )\\ / J, A nalysis of the adult petion of the Tri-State teukemia Survey Cata at Argonne National Laboratory indicating that the x-ray-etisted excess leukemia risk is sraller than previously sugguted and thct it is limited to cases of males with acute and chronic myeloid leukemia and more than 40 t.rmk creys. The children's pertion of thefnalysis was under way at year's entL Radiation Protection Standards NRC has undertaken a major revision cf its tasic tadiation pectection standards (10 CFR Part 20) in order to implement certain recomendations of the International Commission on Radiological Protection. An NRC pamphlet explaining misadministratice reporting requirements,that became effective November 10, 1980, was sent to medical licenseegand the reports received up to the end of 1981 from NRC licensees show that aboct EOD of the 5 eillion annual administrations of radioactive material are n.ishandled. This rate of.01 percent compares favorably to an estimated 15 percent aisadministration rate for all drugs administ in hospitals. Also 02/03/El 42 CHAPTER 10 ANNUAL REPORT

{, '. 4 published was a proposed rule that would require medical licensees to measure radiopharmaceutical desages before adr4nistration to patients. NRC amended its regulations to permit local disposal as nonradioactive maste of certain tiomedical wastes containing tracer amounts of hydrogen-3 and caebon-14 instead of sending them to licensed waste burial grounds. This change vill save medical and academic institutions an estia.ated $13 sillion annually. The e' fort initiated in 1979 to establish a TMI Radiation Worker Registry continued as members of the health effects staff monitored the exposure data on the TMI work force and provided it to the hational Institutes of health TMI ^ "" *'" d'"["t"d dbcommittee.:mu.. mLc= f. *;4. "A Follo ab mic b*$ ' mi., sb .wAAl:w.A_ ~ t_ A 4c. 7., yu.u k.qm.p. Ak, d*w % toa.h ' WASTE MANAGEMENT RESEARCH a +3nssas, b k h d NRC'c waste management research program g *, geprovpmeasurement and prediction methods; confirfdata bases; and develo(regulatory standards to support the licensing of high-level waste repositories, shallow-land burial sites, and uranium mill tailing operations. Nicb-level Vaste M -# TW, ; ;- ;; '. ;;." r f mj- < f S " ,m... 6 m i s d*tre in theN 19 " ' - " E:p:-t 2!Mi&i.- ^i w on c.;nnfhi inhd ihe N r ~~ f; t o,[ f a.)ts h+ l k:" : b,esearch into the rate of corrosive attack on metal r 2 A canisters, and on backfill testing methods for evaluating radionuclide containment by backfills. Evaluation of the effectiveness of radionuclide containment by a h c.eE.inan*W reposito*y site by assessin;r methods for measuring and predicting the 4 ao m., o geochemical capability of rocks to limit radionuclide movement, enQaluation of other characteristics of proposed waste disposal sites. huhlicatian-a# ethnical criteria for the disposal of high-level, w ,..u. radioactive waste in geologic repositories (10 CFR Part 60hfor public cumment kw. .w e~. in July 1981; :n;ur-the procedural requirements by the Commission in 3 February 1981; and ' un ' ; 2 :-

/c' the draf t standard fo 02/D4/82 43 CHAPTER 10 ANNUAL REPORT

L f l l l 3 m e3M' G.T The emphasGW1evel-waste research is en estabihbing confidence that t this waste can be isolated from the bioenvironment for long periods in geciogic j iewpositories. The program investigates waste fore and container materials. I

peological and f ydrological sciences reposito'y engineering and design assess '

t r ^ Iment, and developernt of mathematical models and statistical methods that fore e^ h hasis af a rist methanology for assessirng repository safety. I 1; Activities in the anterials science m in fiscal par 19El --" --. t % e~s~ m .s-sordse experteentally the durability of satrices and pacLages for wastes, and, team ne. N ~ W the relationship petween potential storage envi neents and the rates at strich salidified wastes could leerh__into ground water $ a f O N 7c; N T u N n_ i "r; e l I i i 1 \\ l

L

entent guide to be used in the DDE site characterization of the geologic Ga6 a s M be

\\ c o~ e.TD repositcry in April 81.

tow-level Waste This program identifies better ways to predict and monitor migration of radionuclides from disposal facilities and to find alternatives to shallow-land """' "'"O'i','1 sL6 d.., u:sbt.iy =Ltles !. J Le J eh .g uake e%.n 1981,bhe uRe - _, m ; __t 1- _t:.:.e a+ wt : a.d. n u+u.w u b .~d.L *ess:aa 4w ..a a. c: ..i s n; N " -:j ' 1! _ i t - W a: ^- mprovfuep decommissioning and 3 ponte5%d siting criterigt..,, yiquidlorle/efwastesthathavebeensolidified prior to burial are tested for stability and retention of radionuclides when immersed in water. The staff continued its development of regulatory guides to support proposed rule 10 CFR Part 61 published in the Federal Recister on July 24, 1981. Other guides were being developed to address such areas as fore.at and content for license applications and environmental reports, site selection, site suitability and chaf acterization, waste classification, and monitoring. Uranium Recovery ~ Dranium recovery research in 19E1 me glaboratory and field tests of methods for determining the radon attenuation properties of natural cover saterials and the development of attenuation models based on simple p5ysical tests; evaluations of clay liners and unlined sites for limiting seepage over long periods of time; and assessment of the long-term stabilization of tailings by rock covers. a S hew projects in 1921 included assessments egn situ mining o minimize St en ground water contaminahogA nterim.sta[iliyation of tailings to reduce i y wehant ;tted airborne contaminati cnemical eep to limit contaminant mobility below the j, water tablet tailing dewatering techniques; and monitoring methods and A instrumentation for detecting contamination. I: 02/04/82 44 CHAPTER 10 ANNUAL REPOR1

O EARTH SCIENCES RESEARCH Hydmlocy A generic study was under',aken in 1981 dealing with un;sturated flow and MLW) L transport through fractured rock related to high-level maste repositorieg en g en reactor sitingresearch continued,the monitoring of hurricane surges along the 6.,.t.e1 N E*xad N **N*g:c aJ yAel ec f eaea** ";i ^f

b I ,'d I
- 'J i

d d E*I T *l morica Lain.C b l i 7 ~ Qt.A a sik ;.%y = u L M ."t %y, e'1. 4 .4. de " n-3 - Q,A, t.~ > ueo pov and Seismolony In situ testing needed for high-level weste repositories was evaluated, and a list of underground openings that could be used for test facilities has been compiled. The types of tests that may be used to evaluate coupled thermomechanical and hydrological ef fects in repository rocks and barkfill materials were outlined. ip

' ' Q uun;c-c.u 5tudies of geophysical methods used to minimize borehole intrusio,a 7-produced the outline of a new method for processing geotomograptly data. As a result of the May 21, 1980 ount St. Helens' volcanic cruption, olcanic hazards study program has been started. This program will study and attempt to estimate potential volcanic hazards to nuclear power plant sites in the I

northwestern United States. Additional work concerning the study of geology i and faults in north central Dregon has also been initiated. Methodologies are being studied 'or use in detrreining the recurrence interval between earthquakes at nuclear power plant sites and to rank the f eeka!4ues kc*frW'ugs to determine those intervals. heteorolooy Ys Based on information developed by i ;. 4 , - r: m:rch progra EG/CR-2260 which provides the technical basis for Guide 1.145, on atmospheric dispersion models for accident consequence assessments; and Revision I to Guide 1.23, on the meteorological research programs for nucleer s . power plants, weye publishej! or being revised at yearJend. T amolda 3.+eaeca % h mde.) ea.LauA Idaho field Experiment consisting of nine tests mas conduc#get y = = ted between July 15 and July 31, 1921. Each test involved eight-hour release of tracer materials, with plume trajectories octemined by oil fog tracers and radar-tracked tetroons. m. dp MMo d kes a C. lCleno6 RNGc. - m 1r see -- uM:e a,,+ e. n7..,.s.Mh b emaec C CHAPTER 10 ANNUAL N P0RT ~ ~ ~ 9,t g.. d.=,:. % 4 ' & bQ m.. t 1 h

o a r* e ' s l aN b est 4 l I w Four draft International Atanic Energy Agency (IAEAT safety guides dealing Mth i ydrology utre nrieved. 51ptificant contrhttions were also provided in j, t the development of proposed nle 10 CFR Part ED on HLW paologic repositories; !i .sf pmposed nie 2D CFR Part El en land disposal of low-level radioactive unste, amuf of a draft guide providing standarul format and content for site characteriza- + tion reports for ifLW geologic repos,itories. 1 d 6 i O b#M ST" = i W g \\) t i I; i I l J

ac:t':=nch q K,&gl:yn.n a m. gLAm 50 *l0ss< - s I 9-g [* '. 4 ' c_e bd d e(Y M Lb ELE d

h G E. *>

O report, NUREG/CE-2252, on national thunderstorm frequencies for the contiguous United States, issued in September 1981,5.r4wt: " " n t':- 'c % I -tAe-et**f.e ic - t v : _

m

T, a bas proved useful i,to 4

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c, g; d- ,f % Lltt.se % eil,*h. ed IAEA REACTOR SAFfTY STANDARDS c[ k See page NRC 3980 Annual Report for a description cf this program. In 1981, the Senior A$ory Group, Technical Review Committees, and their working groups forwarded three draft guides and eight completed safety guides to the Director General of the IAEA. Some 53 of the 56 planned IAEA safety I guides are undergoing review, with the NRC tcsearch staff coordinating the reviews within the U.S. l SERT I 8 f NATIONAL STANDARDS PROGRAM i The national standar'ds program is conducted under the aegis of the American . National Standards Institute (ANSI). ANSI acts as a clearinghouse to coordinate I l the work of standards development in the private sector. W. Q Th'e NRC staff is active in the national standards prograt, particularly with respect to setting priorities so that regulatory views are known regarding the standards that can be most useful in protecting the public health and safety. NRC participation is based on the need for national standards to define acceptable l l 1 ways of implementing the NRC's basic safety regulations. I i l The actual drafting of standards is done by experts, most of whom are members 1 of the pertinent technical and professional societies. Approximately 250 NRC i staff members serve on working groups organized by technical and professional societies. National standards are used in the regulatory process only after 'I ( independent review for suitability by the NRC staff and after public comments on their intended use have been solicited and considered. - s / 02/03/El 46 CHAPTER 20 ANNUAL REPDRT 9. .nm. em f}}

ML19310B144

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