ML20247E252

NRC Safety Research in Support of Regulation - 1988

ML20247E252
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Issue date:	05/31/1989
From:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-1266, NUREG-1266-V03, NUREG-1266-V3, NUDOCS 8905260187
Download: ML20247E252 (49)

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e l NUREG-1266 l Vol. 3 NRC Safety Research in Support of Regulation 1988 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research sa atc oq as kj$uj i;P 228Ai?

1266 R " PDR

s AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publ ations Most cocuments cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room, 2120 L Street, NW, Lower Level, Washington, DC 20555

2. The Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082.

Washingtor', DC 20013-7082

3. The National Technical Information Service, Springfield, VA 22161 Although the 1, sting that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Document Roorn include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulietins, circulars, information notices, inspection and investi-gation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceed-ings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regula-tions in the Code of Federal Regulations, and Nuclear Regulatory Commission issuances.

Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and stata legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of information Resources Management, Distribution Section, U.S.

Nuclear Regulatory Commission, Washington, DC 20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory i process are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, and I are available there for reference use by the public. Codes and standards are usually copy-righted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10018.

NUREG-1266 Vol. 3 NRC Safety Research in Support of Regulation - 1988  :

Manuscript Completed: April 1989 Date Published: May 1989 Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 l

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ABSTRACT This report, the fourth in a series of annual reports, was for related decisions in support of NRC licensing and in-prepared in response to congressional inquiries concern- spection activities.This research is necessary to make cer-mg how nuclear regulatory research is used. It sum- tain that the regulations that are imposed on licensees marizes the accomplishments of the Office of Nuclear provide an adequate margin of safety so as to protect the Regulatory Research during 1988. health and safety of the public.His report describes both the direct contnbutions to scientific and technical knowl.

The goal of this office is to ensure that safety-related re- edge with regard to nuclear safety and their regulatory ap- 1 search provides the technical bases for rulemaking and plications.

iii NUREG-1266

TABLE OF CONTENTS All STR AC T . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii EXE CtJI1 VE S UM M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii 1 IN'IT;G RrfY OF REACTOR COMPONENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1 Reactor Vessel and Piping Integrity . . . . . . . . . . . ............................................... I 1.2 Aging of R eactor Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..5 1.3 R eactor Equipment Q qualification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.4 Seismic Safety . .......... .... .. ..... ..... . ........... ..... .............. . .... 7 2 PREVENTING D AM AG E TO REACTOR CORES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2.1 Plan t Perfo rm an ce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 2.2 I luman Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... ........... ... 13 2.3 Accident Management and Individual Plant Examinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3 REACFOR CONTAINMENT PERFORMANCE AND PUllLIC PROTEC110N FROM RADIATION . . 17 3.1 Core Melt and Reactor Coolant System Pailure . . . . . . . . . . . . . ...... .............. .. . . 17 3.2 Reactor Containment Safety . . . . . . . .. . . ......... .......... . . ...... ...... .. 18 3.3 Containment Structural Integrity . . . . . . . . . . . . . . . . . . . . . . .......... ...................... 20 3.4 Reactor Accident Risk Analysis . . . . . . . . . . . . . . . . . .. ..... .. .. ... ........ .. ... 21 3.5 Application of Severe Accident Research . . . . . . . .. ... .......... ................ .... ... 22 3.6 Radiation Protection and 11ealth Effects . . . . . . . . . . . . . . . . . . . . . ............. ...... .... 23 4 CONFIRMING SAFETY OF NUCLEAR WASTE DISPOSAL . .. ..... ... ........ ..... .. 27 4.1 I Iigh-Ixvel Waste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... .. . ...................... 27 4.2 low-lxvel Waste . .. .. .. . . . ....... . ......... ........... ..... ... 28 5 RESOINING SAFETY ISSUES AND DEVELOPING REGULATIONS . . ..... ................. 31 5.1 Generic and Unresolved Safety Issues . . . . . . . . . . . . .. .. .. . ... .. .... ........ .. 31 5.2 Standardized and Advanced Reactors . . ....... .. .... . . ....... . .. ............ 33 5.3 Fuel Cycle, Transportation, and Safeguards .... .. .. . ... . . .. .. . . .. ..... 34 5.4 Developing and Improving Regulations .. . . . . .. ..... . .. . .. .... 35 APPENDIX-1988 Regulatory Products from the Office of Nuclear Regulatory Research . . . . . .. . .... .... . ... . . .. . .... . .... 37 l

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EXECUTIVE

SUMMARY

NRC safety research is vital for implementing a large financial assurance, license termination, number of the agency's programs. Research provides the content of decommissioning plans, and bases for timely rulemaking and related licensing and in- recordkeeping.

spection activities that are based on NRC's longstanding philosophy of defense in depth. 'Ihis philosophy provides 2. Need for Change a clear and logical structure for the safety research mis-sion area, which eonsists of five majorprograms: Integrity a. Studies of steel from a test reactor indi-of Reactor Components, Preventing Damage to Reactor catcJ Aat embrittlement of material ex-Cores, Reactor Containment Performance and Public posed to low neutron flux at low tem-

~ Protection from Radiation, Confirming Safety of Nuclear perature might be substantially higher Waste Disposal, and Resolving Safety issues and Devel-than had been previously supposed. As oping Regulations.

a result, NRC conducted an urgent short-Provided herein is a summary of findings, results, and ac-term review of the embrittlement of complishments of the safety research mission areas that reactor supports and support materials (1) have led to, or are being mcorporated in, specific Com. that are exposed under low-flux, low-mission actions to ensure or enhance the levelcf safety in temperature conditions. Ihc review re-activities or facilities being regulated; (2) demonstrate a vealed no immediate safety problem but necdforchanges in regulations or regulatory approach; determined that the issue needs more and (3)confinn or support the regulations or regulatory ap- study.

proach.'Ihis summany contains five parts that correspond to the five safety research programs indicated above. b. Aging research studies were completed Regulatory products emerging from these programs are on specific safety-related equipment in hsted in the appendix to this report. order to (1) identify failure mechanisms resulting from aging and service wear; A. Integrity of Reactor Components (2) recommend maintenance, inspec-tion, surveillance, testing and condition

1. Specific Commission Actions to Enhance monitoring to ensure openttional readi-Level of Safety ness; and (3) establish degradation pat-terns for use in detecting incipient fail-

a. Radiation embrittlement of the reactor ures. Among the components and vessel is characterized by an increase in structures evaluated were pumps, pres-the temperature range in which the ves- surizers, control rod drive mechanisms, sel steel changes from relatively brittle to feedwater and main steam lines, and tough and ductile (nil-ductility transition BWR containments.

temperature). In May 1988, the NRC published Regulatory Guide 1.99, Revi- 3. Confirmation of Regulations sion 2," Radiation Embrittlement of Re-

a. 'Ihe degraded piping program, which ex-actor Vessel Materials," containing a amined the load-carrying capacity of correlation of the shift m rderence tem' p pes c<mtaining cracks, was completed.

pensture to neutron fluence and copper 'Ihe results validated the flaw evaluation and nickel contents. procedures for stainless steel pipe and welds in ASME's Boiler and Pressure

b. Based on results of the recently com- Vessel Code,Section XI, which are the pleted steam generator research pro- bases for the NRC regulations.

gram, draft revisions of Regulatory i Guides 1.83 and 1.121 on inservice in- b. Seismic research on the Mccrs Fault in I spection and plugging of steam generator southwestern Oklahoma showed that the I

tubes were prepared. Also, value-impact Meers Fault is a capable fault, i.e., there analyses for implementation of the im- have been ground displacements at the proved recommendations were initiated. carth's surface within the last 35,000 years.This is the first fault located east of

e. Final rule amendments on decommis- the Rocky Mountains that has been iden-sioning nuclear facilities were issued. tified as capable. These results were Regulatory guidance is in preparation to transmitted to NRR in Research Infor-provide additional information on imple- mation I etter No.151, "Results of mentation of the rule in the areas of Mccrs Fault Investigation."

i B. Preventing Damage to Reactor C. Reactor Containment Perfor-Cores mance and Public Protection

1. Specific Commission Actions to Enhance rom Radiation Level of Safety 1. Specific Commission Actions to Enhance Level of Safety

a. In 1988,58 tests were conducted in the a. The NRC is revising its regulations re-joint NHC-industry experimental facility garding standards for protection against called MIST. These tests verified the ionizing radiation (10 CFR Part 20) to best estimate thermal-hydraulic code, ensure that Part 20 continues to provide  :

RELAP5/ MOD 2. These results were adequate protection of public health and then used by the utility to analyze a de- safety.The final comprehensive revision sign change, at the Davis-Besse nuclear was sent to the Commission in Novem-power plant, to resolve NRC concerns. ber 1988. This revision modifies NRC's radiation protection standards to reflect

b. Generic Letter 88-20, requesting all li- developments in the principt::s and sci-censees to perform an Individual Plart entific knowledge that have occurred Examination for severe accident vul- smce Part 20 w:s issued. These develop-nerabilities, was issued. mets not only melude updated scientific information on radionuclides update and metabolism, but also reflect changes in

c. A staff evaluation of the methooologyfor the basic philosophy of radiation protec-Individual Plant Examinations devel- tion. This revision also brings the NRC oped iy the IDCOR (Industrial De- standards into conformance with the graded Core Rulemaking) program was Presidential guidance issued in 1987 and provided. with radiation protection recommenda-tions from the International Commission

2. Need for Change on Radiation Protection (ICRP) and the National Council on Radiation Protec-tion and Measurements (NCRP). The

a. On September 16, 19"- tv NRC Commission intends to issue the final amended its regulations to ahow de use rule m early 1989.

of new methods to demonstrate the ef-

b. A final rule was published to establish festiveness of the ECCS during a design emergency planning and preparedness basis loss-of-coolant accident. The new requirements needed for fuel loading best-estimate calculations methods' and low-power testing on nuclear power based on research since 1974, replace the plants.

conservative, unnecessarily restrictive, old Appendix K rule. This will permit c. A probabilistic risk analysis (PRA) re-more econoroical plant operations with- view was provided for the draft safety-out compromising safety, evaluation report for the Westinghouse SP-90 application for Preliminary De- i

b. A workshop was he:d from which Siga Approval.

l emerged recommendations for short- 2. Need for Char.ge term and long-term research on organi-L l zation and rna"agement performance in Ir.terim recommendations were provided to l plant and uti'dty settings, the Commission for the Mark 1 Contain-ment Performance Improvement program.

l c. Guidelinesior p'mt operationalreliabil 3. Confirmation of Regulations ity programs were deveioped and pub-lished, therey resolving TMI Action The following findings are in support of the Plan item II.CA ugarding the applica. closure of severe accident issues:

tion of relhbility engineering to mein-taining operational safety throughoct the a. Calculations were performed to suggest life cycle of the plant. that, in an unrecovered station blackout I NUREG-1266 viii E l

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accident at a PWR reactor, natural reliability of the engineered alternatives circulation of steam within the reactor designs.

coolant system may result in failure of

' the hot leg piping prior to core slumping. E. Resolving Safety Issues and Such failure would depressurize the .

reactor vessel, thereby avoiding the Develop.ing Regulations high-pressure ejection of molten core 1. Specific Commission Actions to Enhance material into the reactor containment. Ixvel of Safety This result has been factored into the latest version of the reactor risk docu- a. The Commission issued a Policy State-ment, NUREG-1150. ment on the Maintenance of Nuclear Power Plants in March 1988. In this pol-

b. A core-concrete experiment was con- icy statement, the Commission indicated ducted, among other things, to test an as- its intention to pursue a rulemaking on sumption about the chemistry of such in- maintenance. In developing this pro-teractions present in most models of posed rule, the staff had extensive inter-core / concrete interactions. The result of actions with U.S. industry (airline and the test was fairly dramatic evidence that nuclear) and foreign nuclear mainte-a key chemical feature that had been nance programs and practices. In addi-omitted from most models was needed to tion, a 3-day public workshop was held in appropriately describe what is likely to July 1988 to solicit feedback on rulemak-take place in a severe accident involving ing. options. Information gathered in reactor vessel meltthrough. Models are these interactions and from the work-now being revised to account for this fea- shop was used in formulating the pro-ture, which has relevance to both PWR posed rule and its supporting regulatory and BWR containment performance. guide. The proposed rule was published for public comment in the Federal Regis-ter in November 1988. 'Ihe Commission D* ConfirminE SafetY of Nuclear mtends to issue the final maintenance Waste Disposal rule and a proposed draft regulatory

1. SPceific Commission Actions to Enhance gup(w pm s guMance for cond plymg with the rule) m early 1989.

Level of Safety

b. The Department of Energy has submit.

Research Information Letter (RIL) 152 was ted for NRC review three advanced reac-issued reporting and summarizing the.re- tor conceptual designs.The designs con-sults of a 5-year investigation of isotopic and sist of one modular high-temperature geochemical methods of dating ground gas-cooled reactor (MIITGR) and two water to complement dynamic hydrologic advanced liquid metal reactors (Sodium testing for characterizing the hydrology of a Advanced Fast Reactor [SAFR] and repository site. Specific attention was given Power Reactor Inherently Safe Module to appfying the technique to the Yucca [ PRISM]).The purpose of these reviews Mountain site. is to determine the licensability of these unique adv: meed reactor designs. The

2. Need for Change NRC staff nas completed the draft safety evaluation report (SER) for M11TGR, i' g a. NRC completed a 5-year study or, the PRISM, and SAFR.1hese advanced re-long-term performance of high-level- actor SFRs have been reviewed and com-7 waste (HLW) packages. *Ihis work made mented on by the ACRS. The Commis- .

definitive contributions to the technical sion decided to issue the MIITOR SER i base of NRC's1ILW regulatory program. to the Departrner.t of Enera for review and commenClhesuffclar stoissueth  !

SERs on the two renaicing litpid metal j J b. Research was completed on the reliaSil- advar.ced reactors in early 19F. i lf hy of engineered enhancements for t.hal. 4

c. The Commission issued an Advanced I l low land burial of low-level waste (11W)

Tr.e rescarch indicated that the cover Notice of Proposed Rolemaking(ANPR) component was most responsiblefor the to inform the public that the NRC is ix NUREG-1266 I

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considering amending its regulations re- - GI B-5, " Buckling Behavior of garding enhanced professional or educa . Steel Containments."

tional credentials for senior nuclear power plant operating personnel. The - GI-99," Loss of RHR Capability in proposed requirements are intended to PWRs." A combined strategy of further ensure the protection of the hardware and procedural changes health and safety of the public by improv- was confirmed to be cost-effective.

ing the capability of shift operating crews to effectively respond to offnormal situ- - GI-II.E.43, " Containment Integ-ations and could also add operating expe- rity Check." Nobackfit or modifica-rience to plant management by opening tion to plants was required.

for senior operators a career path into plant management. The proposed rule - GI-93," Steam Binding of Auxiliary was published for pubhc comment in the Feedwater Pumps." Increased sur-Federal Register m December 1988. The veillance requirements were con-Commission mtends to make its decision firmed to be cost-effective.

on whether to issue the final rule and an associated policy statement or just to is- Need for Change 2.

sue a pohey statement in early 1989.

d. Final resolution was achieved for the fol- a. An advanced notice of proposed lowing safety issues: rulemaking on the regulatory options for nuclear plant license renewals was pub-

- USI A-45, " Decay Heat Removal lished.

Requirements." Using risk-based cost / benefit methodology, a deci- b Proposed resolutions were published for sion was made to require plant-the following safety issues:

specific analyses under the Individ-ual Plant Evaluation (IPE) program. - USI A-47, " Safety implications of Control Systems."

- USI A-44,* Station Blackout." The - USI A-40, " Seismic Design Crite-Commission amended its regula.

ria." Revisions to the standard re-tions to require that plants be capa. view plan are proposed.

ble of withstanding a totalloss of ac power for a specified duration of '

tim ;

- GI-103, " Design for Maximum Probable Precipitation." Revisions

- GI 125.11.7, "Re-evaluate Provision to the standard review plan are pro-to Automatically Isolate Feedwater posed.

from Steam Generator in ljne Break." No backfit or modification - GI-II.E.6.1, "In-Situ Testing of to plants was required. Valves."

NUREG-1266 x

1 INTEGRITY OF REACTOR COMPONENTS This program is conducted to ensure that reactor plant ditions for failure and to ensure that an adequate experi-systems and related components perform as designed mental basis exists to validate those procedures.The most d uring both normal operations and accidents in order that critical facet of pressure vessel integrity is that of their functional integrity and operability can be main- embrittlement of the pressure vessel steel as a result of tained over the life of the plant. Reactor safety depends bombardment by neutrons escaping from the fuel core on maintaining the reactor system pressure boundary un- during normal service. Experiments are thus conducted damaged and leaktight. Failure to maintain pressure to develop a base ofinformation on all the factors that will boundary irdegrity could compromise the ability to cool cause this embrittlement to increase during service life.

the reactot tre and could lead to a loss-of-coolant acci- Much work is done to establish correlations between dent accompanied by release of hazardous fission prod- small-specimen behavior and thick-section behavior to ucts. ensure that the analyses performed to assess structural in-teg"rity are valid. Thus, use is made of large-scale " mod-els that realistically represent the,true components.

1.1 Reactor Vessel and Piping Similarly, the ability to predict integnty m pipmg has re-Integr.ty i quired testing of full-sized sections of pipe havmg a vari-ety of cracks that could develop in service to determine if such cracks could cause failure during either normal serv-1.1.1 Statement of Problem ice or an accident. For both vessels and piping, knowledge of the late at which cracks grow is veiy important to en-The primary system of a light-water reactor (LWR)is the sure that a comp principal boundary enclosing the nuclear fuel core and operational pen,onent od. Ihus, many will not fail durmpts forthcomi the water coolant used both to maintain suitably low tem- ducted on a wide vancty of ment pert,matenals expenmentsunder are a con-eratures on the fuel cladding and to conduct the heat very wide range of typical and ex cted exposure condi-from the fission reaction to a heat exchanger (for a PWR) tions to determme the maximum undmg rates of crack where it can be converted into steam for electricity gen- growth. Detection and sizing of flaws and cracks in all pri-cration.The primary system incluoes the reactor pressure m ry system, co,mponents are conducted by the industry vessel, primaiy coolant piping, primary p(um s, andthrough penodicthe steamRs), mscrvice pri, mspections at shutdowns. To generators. For boiling water reactors H ensure that the inspections rehably detect and accurately mary system must include the steam line at least out to the size the flaws, extensive tests are conducted with inspec-first isolation valve. This boundary must be kept intact tion teams drawn from the mdustry using typical equip-and fully serviceable at all times to ensure that water cool- ment and techniques on samples whose flaw conditions ant is always available to cover the fuel core so that heat, are known.1 rom the results, it is possible to determine either from direct power generation or from decay follow- which techmques are effect,ive and the magnitude of the ing shutdown, can always be safely conducted away, thus error bands for both detection and sizmg. Improvements precluding a core meltdown accident. The principles of in methods are propos,cd and qualification rocedurcs de-of the primary system ensuring the structural integritbe elements of fracture veloped that can provide better assurance or not missing com nents are embodied m t flaws m future mspections and for sizmg flaws more accu-mec nics procedures used to predict conditions for fail- rately. Use is made of materials and components removed ure. These elements are (1) knowledge of the material from actual service to measure the real condition of mate-properties (strength, toughness, embrittlement, etc.), es- Pf0ferties resulting from years of service, to establish U3I theas rea corrosion state, and to pecially a consequence the chanfes in operations; o nuclear those properties that can (2) knowledge of occur flaws that have been " called, validate the existence,of and estimated m size lied the pressure and other stress loadings that can be apfromthrough nondestructive examination procedures.

to the components either from nonnal operations or accidents: and (3) knowledge of the presence and size of cracks or other flaws in the components. The regulations, 1.1.3 Research Accomplishments in 1988 codes, guides, etc., that pertain to the st:ucturalintegrity i of LWRs are focused to ensure thst . cossible corabma- 1.13.1 Pressure W ssel Safety tions of material properties, loads, and flaws viu >icld Wiciheds for evalua'm.g the potential for vessel fractcre adequate margins agamst failure of primary tplem com.

ponents.1he goat of the Reacter Vcssel and Piping Integ. must encompass both normal operattog constions and rity element is to ensure that appropriate analytical p;o. postulated accident conditions. fhey also must encom-cedures exist for assessing the safety of componcnh pass the full range of material behavior-fully ductile to

)' fupy brittle ~and the reactoroperatmg environment. In during normal service and accidents and that s'afticient, this regard, there ate three areas that have been empha-

! critical experiments are conducted to validate those pre.

cedures, sized ir NRC-spontared research during 1988: fracture evaluation, radiation embrittlement, and survetlLmce dosimetry. I l

1.1.2 Program Strategy The NRC's fracture c<aluation rerearch includes both The approach used for this element is to deve7op ar alyti- analyiice.1 and experimental efforts. During 1988, this re-cal procedures for predicting continuing integr;ty or con- search in luded work on developing and refining analysis i

1 NUREG-1266

1 Component Integrity methods and evaluation criteria for reactor pressure ves- tainty in several of the analysis assumptions and input pa-sels fabricated with welds that could be susceptible to low- rameters. Research has continued on several fronts to energy ductile fracture; developing crack arrest data and validate these analysis assumptions and inputs and to de-analyses; and designing pressunzed thermal shock experi- termine the actual margin against failure inherent in the ments (IrrSEs) to evaluate low-energy ductile fracture l'fS analyses.

and stainless steel cladding effects.

The HSST program continues to perform most of the NRC's regulations require that precautions be taken to NRC's irrS research. That research in recent years has  ;

avoid nonductile failure of the reactor pressure vessel. been focused on erack arrest evaluations and on perform-They also require that the ductile fracture resistance re, ing benchmark experiments to evaluate specific details of mam above a specified limit as measured by the material's postulated I'rS accidents and the possible vessel fracture Charpy V-notch upper-shelf energy. If the upper-shelf associated with them. R,cactor pressure vessel analyses energy falls below the 50 ft-lb regulatory limit, a detailed for postulated FTS loadmg have shown that the stect's analysis must be performed to demonstrate that an ade- abihty to arrest a rapidly propagatmg crack, termed the quate margin agamst failure is ensured or the vessel might crack arrest toughness, can be very important m demon-have to be thermally anneated. There are some vessels stratmg adequate margm agamst fCure of some pressure currently in service with welds in which the Charpy V- vessels.The NRC's crack arrest research seeks to advance notch upper-shelf energy is projected to fall below the II.ie state of the art m crack arrest analysis models, to pro-existing regulatory limit before the end of the vessel's vide the experimental data needed to validate these design life. 'Ihese welds are commonly termed low upper- analyses, and to justify changes in the ex,istmg ASME shelf welds. Research has been started to provide a firm crack arrest toughness curves. As part of this effort,large technicalbasisforevaluating continuedoperation below specimen tests-the so-called wide plate crack arrest the 50 ft-lb limit and to vahdate the beneficial effects of tests-were imtiated to provide the needed crack arrest thermal annealing. data. 'the second series of wide plate expenments was completed in 1988, and the results suggest that the ASME curves could be modified and extended to higher Dun.ng 1988, the Oak Ridge , National Laboratory crack arrest toughness values. Analysis of these results (ORNL) under the IIcavy Section Steel Technology will continue, with the possibility that a few additional (HS$r) program performed a detailed review of the basis tests will be needed, to develop the technical bases for the 50 ft-lb limit and the margins that have been in- needed to justify changes to the ASME curves.

cluded in the evaluation criteria developed by the Ameri-can Socie y of Mechanical Engineers (ASME). The re-Neutron radiation embrittlement of reactor vessels has suits of this review show that, the 50 ft-Ib limit has a firm been found to be higher in many plants than previously techmcal basis, but the way m which margins have been mcluded in the proposed evaluation entena may be un- thought. The NRC's regulatory documents are being up-dated to reflect this new information. Also, research is be-necessarily restrictive. Also during 1988, work was per- ing performed to examine the factors that control neutron formed jomtly by the U.S. Navy's David Taylor Research radiation embrittlement and to develop additional data Center at Annapohs and the U.S. Naval Academy to that can be used in updating the regulatory documents.

evaluate and revise the methods for determmmg a,stect's As a related effort, the effects of low-temperature, low-ductile fracture resistance from laboratory specimens. flux irradiation on the integrity of reactor pressure vessel This work also exammed methods for exmipolating labo- supports are being evaluated.

ratory test results to pressure vessel evaluations. The re-sults show that the current American Society for Testing and Materials (AS'I M) test and data analysis procedures The embrittlement of reactor vessel materials is charac-are overly restrictive and that more realistic limits on the terized by changes in a " reference temperature for nil-ductility transition," which can be defined as follows. For test data lead to a reasonable,yet conservative, procedure for extrapolating the data. m my reactors now in operation, toughness of the beltline materials at room temperature is too low to permit full pressurization of the vessel with adequate safety margins.

Under certain postulated accident conditions a pressur- As temperature is raised, toughness increases slowly at i:.ed water reactor (PWR) pressure vessel couM be sub- first;but, at the refer ence temperature, toughness begins jected to severe cooling rates coupled with a high internal to incretse much more rapidly. At normal opcrating tem- '

pressure. This combmation of thennal and pressure peratures, vessel materials are cluite tough.

stress, called pressurized thennal shock (l'TS), could pose >

a serious chauenge to tne integrity of some older pr essure Tc monitor radiation errfor;ttlement in reactor vessels, J tessels that have developed a significant dqrce of specbens of the most radiation-scitsitive materials src j embrittlement because of neutron irradiation. In 198f exposed ir, stirveillaace capsules positioned inside the j the NitC's regulations were amended to establish a limit tessel near the wall. Destructive tests of these specimens l en irradiation damage that could not he exceeded unless when tbc capsule is withdrawn after several years of ex- q ddailed analyses show that continued operation would be posure previde a data base for studies of the relationship .

safe. In 1987, regulatory guidar.ce on performing these of cmbnttlement to neutron fluence and material compo-analyses was issued. Although the rule amendment and sition.

regulatory guidance provide reasonable assurance that -

potemial l'rS accidents will not lead to PWR vessel fai!- In May 19f,8, the NRC published Regolatory Goide 1.99, ure, the actual margin against failure is clouded by uncer- Revismn 2,"Radir, tion Embnttlement of Reactor Vessel NUREG-1266 2

1 Component Integrity Materials," containing a correlation of the shift in refer. In 1988, based on results from this research program, enee temperature to neutron fluence and copper and draft revisions of Regulatory Guides 1.83 and 1.121 for nickel contents. Regression analyses of the surveillance improved guidance on inservice inspection and plugging data base by two independent investigators provided the of steam generator tubes were prepared. Also, value-technical basis for the guide. Peer review by public com- impact analyses for implementation of the improved ree-ments gave general agreement. 'Ihere was further peer commendations were initiated.

review by two national standards committees that were using the guide as a basis for their standards. Recently, 1.1.3.3 Piping Integrity the guide was checked against the considerable body of surveillance data received since the original correlations A vety si nificant problem encountered in boiling water were made and found to be satisfactory. Ihe maintenance reactors IWRs) has been the intergranular stress corro-and analysis of this data base are done by ORNL on con- sion crac ing of austenitic stainless steel piping at weld-tract with the NRC. ments. 'this condition has been responsible for over 400 pipe-cracking incidents throughout the world over the in addition to analyzing the surveillance ca;wule speci. last 10 years. Because these problems have resulted in men data, the NRC is evaluating radiation embrittlement extended and unscheduled outages-with extensive in-

'in research programs.These research efforts use materi- SPCFtions, repairs and replacements, and sigmficant occu-als test reactors to provide accelerated embrittlement of pation exposures-the NRC and the electnc utilityindus-various reactor pressure vessel materials so that many dif-try have devoted much research to their resolution.

ferent variables can be evaluated in a relatively short pc- . .

riod of time. In 1988, results from the 5th series of tqst NRC research in this area is focused on developing the reactor irradiations performed in the IISST program capability ,to predict dress corrosion crackmg m 13WRs were analyzed. These results indicated that the ASME's and to vertfy the acceptability of proposed fixes.

method for accounting for radiation embrittlement ef-The use of alternat.ive materials and other proposed ac-fccts on fracture toughness may slightly underpredict the tions to mitigate mter actual loss in fracture toughness. I he impact of these re. has been investigated. granular stress corrosion crackin sults is being evaluated, and the possibility of a change in the ASME procedures is being considered. Other studies steel Type 316 NG, Type 347, and CF-3, have been are under way to evaluate the effects of neutron irradia. evaluated under a van,cty of environmental and mecham-tion on crack arrest toughness, stainless steel cladding cal loading conditions and found to be significantly more fracture toughness, and low upper-shelf weld fracture resistant to the cracking problem than the materials com-toughness. monly used in the past for nuclear plant pipmg. IJowever, tests have shown that under certam water chemistry con-ditions even these superior materials can become suscep-An important aspect of the surveillance program to deter- tible to stress corrosion crack propagation.

mme the degree of embrittlement in the pressure vessel of an operating nuclear power plant is the prediction of An extensive program has been carried out to demon-the amount of neutron radiation exposure (neutron strate the strong interactions amony dissolved oxygen and fluence) of,the vessel. Fluence determmations are made sarious impurities, as well as the cifects of individual im-by calculations to compute the fluence, dosimetry meas- purity species on stress corrosion of sensitized Type 304 urements at key surveillance locations, and a consolida- SS in low-oxygen, high-temperature water.The data pro-tion of the measurements and calculations to reduce un* vide the basis for affirming the benefits of good vater certainties of predictions at critical locations of the vessel. quality and the roa, of different impurities in stress corro-It is necessary that these predictions be reasonably accu

• sion cracking of sensitized austemtic stainless steels. By rate to ensure that the plant is operating in conformance removing certain deleterious species (which have been with NRC safety regulations. from the water, crack growth ctm be sup-identified) pressed or halted. A phenomenological model t.as been A draft regulatory guide that identifies methods and as. developed to aid in understanding and interpreting these sumptions for establishing premire vessel fluence is be- data. I robably the tnost significant proposed action to ing resised based to NRC m/iew prior '.o publication for mitigate stress carrosion cracking in liWR stair.lcsnteel

! pdblic comment. The gu.dc makes tssi of thc develop- piping is the use of " hydrogen water chemist ry," ivhich m-ments generated by the servei!!auce 60simetry program, c!udes additions of hydrogen to lower oxy gen leve;s in the i coohnt and rmintaining very low levels of innpuritics.

LL3.2 Steam Generator totegrity A thermal aging program was initiated in 1982 to evaluate the long-term eff ects on degradation of tougimets in cast The Steam Generator Group Project af flattelle-Pacific stainless steet as a function of time of exposme, tempera-Northwest Laboratories (PNL) has used a re, ired-from- ture, and material composition. Through 1988, resuLs service neem generator from an actual PWR facihty as 4 have been accumtilating to a!!cw a curfitative evalu-test bed fa measuring the effcc.iveness of eddy current ation of the degree and significate of toughness loss at

! 'nspection techniques to detect and size flaws in stcrm reactor operatmg temperatures and operational times.

generator tubing. In t.ddition, tube serments removed hness loss from the generator were burst tested to validate empiri- Also.

are the mechanisms being identified by evaluating responsible forl the tour'aboratory-both cal models of remaining tube integrity developed earlier. exposed specimens and specimens removed from actual 3 N UREG-1266 L

F 1 Component Integrity components in nuclear power plants. A heat treatment continuous monitoring techniques (using acoustic emis-has been identified for recovery of toughness loss. How- sion) for crack growth and leak detection.

ever, reembrittlement during subsequent exposure oc-1 curs at a much faster rate than the initial aging embrittle- An improved method for more reliably detecting flaws ment. and sizmg them with greater accuracy in light-water reac-tor primary circuit components is called the SAFT-UT Very significant pipe wall thinning has occurred in a (Synthetic Aperture Focusing Technique for Ultrasonic number of steel pipmg systems of nuclear plants because 'I esting).The S AFI'-UT technology is based on the physi-of crosion-corrosion of the material by high-velocity cal pnnciples of ultrasonic wave propagation and uses single-phase coolant water.This problem was highlighted computers to process the data to produce high-resolution, .

at the Suny Unit 2 plant where part of the feedwater pip- three-dimensional images of flaws to aid the inspector in l ing was thinned so severely that the pipe failed cata- locating and sizing the flaw (s). After several years of field trophically. A survey was performed of 28 U.S. plants testing and real-time operation at operating reactors, the and two foreign plants to determine general expenence SAFI-UT technology was transferred to Sandia National witn crosion-corrosion and to establish the significant Laboratories and pulled together into a package for easy variables that might be related to the problem. These transfer to the nuclear industry. In 1988 system operation variables included feedwater velocities, pressures, tem- and demonstration was published, and the technology was peratures, water chemistry histories, and materials. Re- evaluated for inspection of cast stainless steel.

suite established that the problem exists to a significant degree. Research work, national and international studies, and field experience over the last severalyears have indicated A state-of-the-art review was performed on the available that inservice inspection, as currently practiced,is not al-data and current mechanistic understanding of erosion- ways reliable or effective. NRC research results have indi-corrosion. It was observed that susceptibility depends cated a need for qualification of the entire inservice strong 1y on the interaction of flow and environmental and inspection (ISI process, including the personnel, procc-dures, and eqm)pment. Research has been conducte matenal variables. Thus, one cannot usefully identify the critical limit of one variable such as velocity or pH or ge- criteria developed for proper qualification of the ISI proc-ometry for crosion-corrosion but must consider all these ess. In 1988, one of the appendices to Section XI on per-factors in an integrated manner. A qualitative under- sonnel training and qualification was approved and incor-standing has been developed of the interaction ofimpor- porated into the ASME Code. Another appendix on tant vanables. Quantitative predictive methods have criteria for performance demonstrations has been ap-been devcloped but are subject to large uncertainties. proved through the major committees and is in its final stages of approval and adoption.

The Degraded Piping Program, conducted by Batte!!c's Columbus Division, has been the NRC's primary piping Research has been under way to develop the use of acous-fracture research program. This 4-year program, imtiated tic emission (AE) for the continuous online monitoring in 1984, was com leted in 1988, and the final report will of reactors to detect and locate crack growth and to esti-be issued in earl 1989.The research examined the load- mate the severity of the cracking from the AE signals. In carrying capacit of. pipes containing cracks. Various pip- 1988, calibration studies were conducted en the Watts mg matenals and sizes were tested under typical reactor Bar Unit 1 reactor m preparation for con'muous moni-operating temperature and pressure. The results from toring during its operation. The availability and proper the analyses used to predict the load-carrying capacity use of this technology will mean that reac' ors can be con-were compared to the experimental results, and improve- tinuously monitored and that any cracks'. hat develop can ments were made to the analyses in several areas. The re- be detected and evaluated. In this way, proper and timely sults of this research have been used to validate the action can be taken to avoid extensive crack growth or ASME's flaw evaluation procedures c(mtained in Sec- component failure.

tion XI of the Boiler and Pressure Vessel Code for stain-less steel pipe and welds They also were used in develop- Evaluation of a stand-alone " Smart" system for AE leak ing similar evaluation procedures for flaws in carbon steel monitonng was completed in 1988. 'Ihe system is capable pipe and welds. The research program has produced six of accurate detection h> cation and sizing of leaks m the summary reports,11 topicai reports,and detai!cd data te- pressure boundary. A detailed topical report was pub-cords for 67 oipe fracture experiments. A user-friendly lished to give detadt of the equipment, calibration and op- {

computer code has been produced for analy ew m cked cratior. procedures, and data analysis and evaluation pro- {

pipe that can be used in various regulatory ara - md it cedures. i has significantly expardec,1 the material prt ; uta j base lor mtclear reactor piping matcrials. g 1.1.3.4 Inspection Precedures and Technologies On June 27,1988, final rule amendments on decommis- '

sioning nuclear facilities were issued (53 FR 24018).

This program includes studies af moroved methods for Regulatory guidance is in preparation to provide addi- l the detectior and sizing of flaws during inservice inspec- tional information on implementation of the rule in the l areas of financial assurance, license termination, content tion of carbon steel, wrought, and cast stainless steel pip-ing and pressure vessels. It also includes studies of onhne of decommissioning plans, and recordkeeping.

NUREG-1266 4

1 Component Integrity The final rule amendments on licensing requirements for crating capacity Utilities are currently planning to apply the independent storage of spent nuclear fuel and for license renewals and have outlined a tentative sched.

high-level radioactive waste were published on Aug- ule for several steps in the process.The first submittal to  ;

ust 19,1988 (53 FR 31651). Regulatory guidance is m the NRC is expected in 1991, with a large number of addi-preparation on storage of spent fuel. tional submittals to follow shortly thereafter. To keep pace with these industry plans, the NRC will need to de-

' A program was initiated to determine the effect oflow- vote effort over the next several years to license renewal.

i temperature, low-flux irradiation on the mechanical A firm NRC policy on the terms and conditions oflicense properties of the neutron shield tank of the Shippingport renewal applications can then be completed by early 1 reactor. As a result of the findings of embrittlement in the 1991. Review of these applications at an early stage will  !

- High-Flux Isotope Reactor pressure vessel at Oak Ridge, provide an indication to the industry of the viability of the there was an urgent need to assess the possible embrittle- life extension option in sufficient time to elect an alterna-raent in present-day reactor supports. The Shippingport tive option if necessary. j neutron shield tank (vessel support structure) provided an excellent opportunity to check for such an effect be- 1.2.2 Program Strategy cause its construction ma;erial is equivalent to the mate-rial used in present-day core support structures. NRC staff effort in aging is being pursued in several areas, including technical and scientific research to identify the l A coring tool procedure was developed for extracting effects of aging on the key safety-related components of 6-inch disc samples from the outer and inner wall of the the plant and to examine methods for mitigating such ef-shield tank.The two walls of the neutron shield tank were fects. Specifically, the strategy is to achieve relative to separated by 3 feet of concrete, and this made the coring each component the following results:

and retrievmg of the 6-inch-diameter samples a rather difficult operation. Twenty-four cores were taken from 1. Identify and characterize aging and service wear ef - l the tank-12 from the outer wall and 12 from the inner fccts that, if unmitigated, could cause degradation of wall. Two of the cores represented material from weld- structures, components, and systems and thereby 1 ments while the others were from the base metal. The . .

imp r pIant safetp samples were also taken from locations that would repre-sent different levels of fluence.

2. Develop methods of inspection, surveillance, and Preliminary results in the area of base metal indicated monitoring and of evaluating residuallife of struc-that the low-temperature, low-flux irradiation has signifi. tures, components, and systems that will permit cantly lowered the toughness (Charpy energy) of the compensatory action to counter sigmficant agmg ef-shield tank (support) material. fects prior to loss of safety function.

L2 Aging of Reactor Components 3. Evaluate the effectiveness of storage, maintenance, repair, and replacement practices, current and pro-Posed, in mitigating the effects and diminishing the 1.2.1 Statement of Problem rate and the extent of degradation caused by agmg.

Aging,affects all reactor components, systems, and struc- I tures m various degrees and has the potential to increase The program covers the electrical and mecham. cal com- '

risk to public health and safety if its effects are not con. ponents and systems important to the safe operation of trolled. In order to ensure continuous safe operation, the plant and the containment struce re, measures must be taken to monitor key components, sys-  :

tems, and structures and interfaces to detect aging degra- 1.2.3 Research Accomplishments in 1988 i dation and to mitigate its effects through maintenance, i repair, or replacement. For an older plant approaching Research studies were completed in 1988 on specific the end ofits design life and for w hich extended operation safety-related equipment in order to (1) identify failure beyond the initial license period of 40 years is contem- mechanisms resulting from aging and service wear; plated, aging becomes a crttical concern and will clearly (2) recommend maintenance, mspe.: tion, surveillance, be crucial to any assessment of the safety implications of testing, and condition monitoring to ensure operational license renewal. readiness; and (3) establish degradation patterns for use in detecting incipient fnitures.

Recently, the nuclear industry has initiated a significant effort aimed at extending the hfe of e :isting plants beyond The Shippingport (Pa.) nuclear pov'er plant, now under-theiroriginal license term of 40 yea *s. According to a De- going Jecommissiomng, ts a major source of ratatall partment of Energy study, the projected nct ber.efit to the aged equipment for the nucicar plant aging resear j

! United States economy can be on the order of $230 billion (NPAR) prograrn. As the first U.S. large-scale, central- ,

. through the year 2030, assumina a 20 year life extension station nuclear plat, the Shippingport reacter is simfiar j for current plants. If a 40-year life extension is j udged fea- to current commercial PWRs m design and opel e tion. It*, I sible, the benefit is even larger.The benefit reflects both 25 years of servree exceed the operating times of most q the lower fuel cost of nurieer plants compared to fouil- currently active nuclear power plantr<. Alsa, because of fncied plants and reduced oatlays for t eplaccinent of gen- substantial modifications during the mid-1960's and 5 NURiiG-1266

1 Component Integrity 1970's, Shippingport offers uni se examples of identical prioritization for indepth aging assessment. Prioritization or similar equipment used side side, but representing was primarily based upon criteria derived from a specially different vintages and degrees o aging. developed nsk-based methodology.The methodology in-corporates the effect upon plant risk of both component The removal of more than 140 electrical and mechanical aging and the effectiveness of current industry aging man-components and samples of naturally aged materials from agement practices in mitigating that aging.

the Shippingport Ator& Power Station has been com-pleted. I hese components a?d samples of materials have An international nuclear power plant aging symposium been shipped to various naticial laboratories for postser_ was held in 1988. Over 500 representatives from 16 coun-vice exammations and tests to i lentify aging mechanisms tries participated.

and to evaluate component performance in aged condi. managementof agmThe g m commercial discussion centered nuclear power plants around tion. Data to be generated from this element of the of all ages, that is, for the currently licensed plants and for NPAR program will be usefulin evaluating plant condi. plants s.eckmg license extension. A significant number of tions of all ages, including those seeking license renewal / the , topics discussed at the symposium addressed the ef-life extension. fcctiveness of mamtenance to manage agmg. Major par-ticipants at the symposium included those from Japan and Maintenance, in its broadest sense, is one of the ke s for I;rance who discussed their respective life-extension pro-managing plant aging and will play a pivotal role in life ex-grams. U.S. mdustry, DOE, and El RI also made sigmfi-tension /hcense renewal. Recommendations were devel- can! contributions toward the overall success of the sym-oped for preferred maintenance techniques tocounteract P"S* *

• age-related degradation in valves, pumps, and emer ency diesel generators. These, recommendations inc uded I)cgradation and failure of swing check valves had led to a (1) use of signature analysis techniques mvolvi motor need to identify method (s) to predict performance and current signatures and thrust momtormg meth( s to de- degradation of these valves in nuclear power plant sys-tect and differentiate motor-operated alve abnormali- tems. As a part of the nuclear plant aging research pro-ties resultin from aging and mamtenance,(2)identifica- gram, methods have been identified to predict the stabil-tion of de iciencies m the current mscruce testmg ity of the check valve disk when pipe flow disturbances, program for auxihary feedwater pumps and publication of such as elbows, reducers, and generalized turbulence recommendations forimprovements m the mscryice test- sources, are present.

ing programs for these pumps, (3) publication of recom' menda, tions for preferred testing and mamtenance of Time-temft has been usecrature super cmcrgency diesel generators and development of techm- proach th in polymers for moreosition than is an em cal data to support the Generic Safety Issue B-56 on die- years to make thermal aging predictions. Using data de- ,

sel reliability. Appropnate ASMhandIEEEcommittees rogram on nuc! car veloped power plant from the cable materials, radiation agin four f > berent materialshave have been informed of the research results and considera-tion is being given to incorporating the results in revised been studied by DOE to develop nrocedures for time-codes and standards documents. temperature-dose rate effects. These materials are hy-palon, neoprene, polyethylene, and PVC jacket material.

l'or two of these materials, extrapolated predictions An evaluation was completed of the use of the based on the superimposed data were found to be in ex-NUREG-0956 mechamstic source term models for calcu- cellent agreement with 12-year, low-dose-rate results de-lating accident radiation closest to safety-related equiP- rived from the operating nuclear power plants.

ment.The models used in Regulatory Guide 1.89 for de-sign basis loss-of-coolant accidents were found to giv comparable integrated accident doses to equipment. be- 1.3 Reactor Equipment Qualification l vere accidents were found capable of giving doses higher by a factor of 10. 1.3.1 Statement of Problem The aging assessment of 11 major light-water-reactor As the result of the Three Mile Island (FMI) accident, concerns and questions were raised regarding the oper-components and structures was cornplcted. Sigmficant ability and structural integr'ty of components during stressors contributing to agmg and wear, degradation sites carthquake and loss-of-coolant accider,t (LOCA) envi-and aging mecha nsms, and pot enmd failure modes were ronments. Although design criteria and loading defini-identified, and curr ent mscruce mspection requirements were evaluated. Components and structures studier were: tions have changed over the years to improve tne integrity of these comp (ments, the concerns z.nd questions dealt di-for PWRs, pumps, pressurizer. Surge and spray lines, rectly with the adequacj of the component qualifications.

charging and safety snychon no7zles, feedwater hnes, control tod drive mechtmtms, and reactor mternals; for Therefore, those items that were identified as high prior-ity were gisen irnmediate research attention and action. It

, BWRs, containment, feedwater and main steam lines, centrol rod dris e m(chanisms, and reactor internals. Also was also intended that the result.s of the research would be mcorporated into standards.

studied were eles trictd cables and emergency dicsci gen-cratcrs- {

Subsequent to the TMI research activity, other safety is-sues were identified and where these had impact on A report was issued to document the results of an expert equipment qualifications, research effort was proposed to panel workshop established to perform the component develop the data base to aid in the resolution of these NUREG-1266 6

V l O

l 1 Component Integrity-  !

high-priority safety problems. Current effort is address- containment when very high velocity flows will develop in ,

ing one of these generic safety issues. Another effort is the pipe. If unchecked, with valves that do not close, the  !

providing guidelines for improving valve qualification leakage can cause serious consequences not only because i standards, of steam release outside containment, but also because other emergency equipment may be exposed to the harsh 1.3.2 Program Strategy 8' "* "I"* "* ""d I"U' The results of the first series of experiments showed that NRC staff effort in the equipment qualification program valve closure was achieved during each blowdown test. l s currently involved with developmg the basis for resolv- llowever, evidence exists to show that numerical values -

mg a high-pnonty generic safety issue (GSI 87) related to used in the analytical methods in the past may not be con- l the operabdity of motor-operated valves (M,OVs). Im- servative for all valve applications. Although all tests portant information will be provided to chmmate the were successful for the hot water conditions, there is an question as to whether gate valve thrust requirements important need to understand and to validate the existing should be based on openmg or on closmg motion. 'lhese method for sizing actuators. In addition, it is necessary to i

' results will be incorporated m the valve qualification stan- develop procedures that can be used to demonstrate and/ i dard and will clarify one of the areas the staff believes may or assess whether older valves in operating plants will also !

be contributing to MOV problems. Another research ef- close under adverse break conditions. 'Iherefore, re- i fort is being devoted to understanding the effects oflarge search effort will continue in 1989 by conducting addi-

, earthquake load,s on the operability of an aged gate valve. tional tests involving other fluid environments and devel. ,

lhe effects on pipmg, support, snubbers, and anchors will oping procedures for evaluating installed valves. l also be determmed. Ihis cooperative program with the '

Germans uses one of their test facilities.

1333 Dynamic Qualification of Mechanical Equipment Future effort in the equipment qualification program will address new problems and safety issues consistent with The seismic testing program was completed in 1988. A safety and licensing needs. Since industry is expected to typical U.S. piping system was considered, with a 30-year-  ;

become more involved in solving some of the pressing old (aged) valve mstalled. The objective was to excite the  !

valve safety problems, some NRC cffort will be devoted piping to mcreasing levels of carthquake loads (to a maxi- l to following this work and evaluating the results with re- mum of eight times the typicallevel)in order to deter- '

gard to licensing applications. mine whether the valve operability would be affected by these large dynamic loads. There was first the concern The results from current and future efforts will be incor- whether the valve internal parts might bind under such porated in appropriate qualification standards to provide loads, preventing the valve from operating. Another part the basis for ensuring safer components. of the test was aimed to obtain pipe stress data for validat-ing a pipe-design computer program.The third important goal was to determine the response characteristics of the i' 1.3.3 Research Accomplishments m. 1988 pipe supports, snubbers, and anchors as a result of these large dynamic loads.

1.33.1 Electrical Equipment Qualification The test data records have not been processed at this An assessment of mechanistic models for estimating time; however, preliminary results indicate that the aged source terms, as described in NUREG-0956, was com- valve operated successfully during and after each test. It is pleted. The radiation dose to equipment from beta and not known whether the valve stroke times changed at the gamma radiation (generated by fission product activity), high excitation levels.

released to containment during accidents, was calculated and, presented in NUREG/CR-5175. The integra,ted ra- 1.4 Seismic Safety diation dose received by safety-related m-contamment equipment from this mechanistic model was shown to be m reasonable agreement with the 200 Mrad integrated 1.4.1 Stalement of Problem dose calculated by the deterministic methods in Regula- Earth uakes are among the most severe of the natural tory Guide 1.89.That dose wasfre entl usedinthe ast hazarks faced by nuclear power plants. Very large earth-in qualification tests for a design - sis OCA acci. ent, Severe , accident scenarios were also calculated usmg quakes would simultaneously challenge the ability of a!!

platt saf-ty systerrs to function and, coupled with the mechamstic models and gave integrated radiation doses to equipment an order of magnitude greater than the likely loss of oIffsite power and dependent safety systems, 1 could pose a unique threat to pubac safety. As with rrany I above calculated for the design basis accideat.

potentially severe conditions there is much uncertamty )

associated with the . design and evaluation of nuclear -

133.2 Environment! QualiGcatiun of Mechanical niants for carth_luakes. Seismic hazard in Ceritral and Equipment hastern United States ren:ains an issue that is not likely to be casily reso'ved These regions contain the highest per- 3 Experiments were conducted ta determine whether centage of nuclear power planti.in the United States. His-valves in high-energy pipes will close as they stonid to toricahy, the h.rgest earthquakes in the United States prevent leakage during a pipe break accident outside the heve occurred at New Madrid, Mo., and at Charleston, 7 NUREG-1266

1 Component Integrity S.C.'Ihe geology of the central and castern regions makes issues, including a strong basis for scismic zonation, it difficult to establish earthquake magnitudes or scismic source mechanisms, characteristics of ground motions, parameters for specific locations or to ensure a proper de- and site-specific response. 'lhe NRC is addressing these sign basis for individual power plants. Recent information uncertainties through research that encompasses sus-from the United States Geological Survey (USGS) sug- tained seismic monitoring, geologic and tectonic studies, gests that many of the currently crating nuclear power neotectonic investigations, exploring the carth's crust at plants could be subjected to h er seismic loads than hypocentral depths, and conducting ground motion stud-were specified when these plants were designed. ies.

Generally, uncertainties involved in scismic hazard analy. The backb(me of the NRC program in the Eastern United sis are not reduced quickl or with a single effort. While it States has been the scismo raphic networks dept d is possible for a technica breakthrough to provide a de. throughout the Eastern and entral United States. he finitive conclusion,it is more likely that the continuing ac. NRC is currently funding seismographic networks in the cumulation of data and understanding will result in a con. following regions: Northeastern Umted States, Virginia, comitant gradual increase in the level of confidence held Charleston, S.C., the Southern Appalachian region, the in the design basis of nuclear power plants for the Central New Madrid (Mo.) region, Ohio and Indiana, castera and Eastern United States. This will require a continued Kansas, and Oklahoma. An agreement was reached in level of effort in field investigation, data collection, and 1986 between the USGS and the NRC to jointly support analysis. the establishment of the eastern portion of a national seismographic network. The national network is sched-uled to be fully in place fiscal year 1992. In the mean-In the 1970's and before, our interest in nuclear plant time, the currenti NRC- unded networks in the Eastern scismic design was mainly) limited to response atnddesignlevels Central Unit d States (e.g., will OBE be gradually and SSI:phased out. knowledg and our primaril based on analytical techni ucs and assump-tions. I the 1980's, a considerable eff rt has been made In twent yeaty, the NRC has supported scismic, testing to better predict the potential response of nuclear plants and the collection of carth uake expenence data m order to earthquakes reater than those considered in design. to improve and, gain con dcnce m the use of scismic PR As and scismic margm studies. lhese data are also be-Our understand ng has been increased greatly by the test-rt pro)osed improvements to scismic ing to failure of equ' ment and structurcs, by the gather- design criteria.p"8.used ,to su , he, cart quake resist ing and synthesis o earth uake experience data from non-nuclear facilitics, and b the large number of scismic equipment, and piping has been found, m general, to be hi her than previousi hought. Major efforts in this area probabilistic risk assessments that have been made. But w I be completed in 990, and the results are being suc-research can only do so much; without having real nuclear cessfully used in licensing actions.

plant experience of earthquakes in the very severe range, some uncertainties will remain concerning plant resis-tance to these events, l.J.3 Research Accomplishments in 1988 M1 Earth Sciences 1.4.2 Program Strategy Columbia University and Pennsylvania State University The strat to resolve the seismic roblem involves re- have been investigating for the past several years seismi-scarch to velop the methods and ata that will support cally active regions in the northeast for evidence of Qua-ternary surface or near-surface tectonic deformation. Co-the necessary the evaluation seismic tools. criteriaisdevelopment The research and p(rovide focused on 1)im- lumbia University has concentrated on specific areas proving estimates of earthquake hazards by identifying where the likelihood of identifying recent surface or near-l potential carthquake sources and determining the propa- surface tcetonic deformation or palcoscismic features gation of seismic energy with distance, (2) estimatmg the was expected to be high such as the epicentral area of the possible range and likelihood of scismic round motions 1983 Goodnow carthquake in the Adirondack Mountains, at nuclear plant sites, and (3) assessing th effect of these Cape Ann, Mass., the I;mcaster, Pa., scismic zone, the ground motions on soil, structurcs, equipment, and sys- Lower fludson Valley-Eastern Newark Basin scismic tems ofIhe lants.The integrated results of this research rone, and the New Jersey Coastal Plain. Pennsylvania will be use ta quanti the risk to nuclear plants from State University has focused its research activities on the cadhquakes, to assess t e scismic rafety mar msinherent lancaster, Pa., and Moodus, Conn., seismic zones.

in current or future plant design, and to het identify and l set priorities for what improvements are needed in plant The identification of surface or near-surface tectonic designs, or what parts of scismic design criteria may be re- structures associated with current seismicity can contrib-laxed. ute substantially toward defining carthquake source s rectures or seismic source zones m the Eastern United A major focus of the NRC reuarch;nograms in geology, States. The palcoseismic investigations, by providing iso-seismology, and geophysics continues to be identifymg prehistoric carthquales, have the and defining potentid carthquake sources or source topic dates notential of iar{ ing deterministic guidance for calcu-for provi 1 zones in the Eastern United Mates r.nd using that infor- lating return periods of large earthquakes in the North-

! mation in assessing seismic hazards with resiv:ct to nu- eastern United Stater.This would be a majm step in as-clear power plants. Mr.ny mknowns exist regarding these sessing scistaic hazards in the Eastern Unitul S'ates.

l NUREG-1266 8

1 Component Integrity Field studies associated with these projects were com- may then provide an indication of belts of larger motion pleted during the past year and the final reports are being that may exist and may therefore help to define areas with prepared. higher seismic hazards.

~

'The NRC has funded over the past few years studies by 1.4.3.2 Component Response to Earthquakes the USGS and the University of South Carolina of soil de-the 1886 carthquakeand of The last static test in a series of large reinforced concrete formed similar, butbyolder, liquefaction features during(paleoliquefaction models features) representing a portion of a nuclear power plant that were apparently formed by prehistoric earthquakes was completed this year. Results from this test, like those of about the same size. The finding of this program-that obtained from the two tests performed in 1987, provided paleoliquefaction features occur less frequently and be. excellent agreement with analyses typically performed by come smaller in size the farther away from the Charleston design engmeers.

meizoscismal area they are located-indicates that the palcoscismic events occurred in the vicinity of the 1886 Although testing in the Category I struct ures program has carthquake, but could have been larger. To date no demonstrated that nuclear plant shear wall structures paleoliquefaction features have been identified north of have high seismic strength, it has been shown that their Southport, N.C. Findings related to this program so far stiffnesses may be significantly less than considered in de-support the NRC position used in past licensmg decisions, sign.'lhe latter findmg will be incorporated in new ASCE i.e., the Charleston seismic area is unique. design guidance that affects the seismic design of equip-ment and piping as well as concrete structures. A study is Results of research by thc University of South Carolina in being conducted to uantif duced shear wall sti ness. y the risk significance of re-the Charleston carthquake epicentral area suggest that smici se to be concentrated at the intersection of Computer Analysis for Rapid Evaluation of Structures (CARES), a personal computer based system, has been recenti developed by the Brookhaven National 12bora-Research m. the central Virginia scismi,c zone by the tory (bbl) for the NRC to perform evaluations of struc-Virguna Polytechmc Institute, which consisted of surface tural behavior and the capability of nuclear power plant geologic mappmg, seismic reflection prof,iling (bo,th S within the seismic zone and outside ofit), monitormg seis- facilities under earthquake loads. staff CARES system wi to assess micity and analyzing carthquake data, and reprocessm, g a analysis methods used for structural safety evaluations.

USGS-acquired scismic reflection profile along Route I-64, led to a reinterpretation of the under the Piedmont and Coastal Plam, basement Two main objectives of the EPRI/NRC piping and fittin

.The more impor-structure dynamic reliability program (PFDRP) were met in 198k tant findings of this research mclude: (1) the crust that with the com compnses the seismic zone contams a greater number of and analysis.pletion of allconsistently PFDRP testing, data reduc The test results showed that subsurface reflectors than the aseismic areas.This is m- pip ng has very high resistance to dynamic inertial loads; ter reted to indicate that the seismic area crust is more ,

typicall failure was produced only by dynamic in hig ily faulted and sheared than the, nonseismic crust; and scaled f5 to 30 times higher than design Thelevels.putloa fail-(2) the seismic zone overhes a section of the crust where ure mechanisms were different from what was assumed the mantle is considerably more shallow than surroundmg areas. when the current ASME Code piping design rules were developed. That is, piping rupture was caused by ratchet-ting and fatigue effects, and cross-sectional collapse did Research on the Meers Fault in southwestern Oklahoma not occur.

showed that the ' eers Fault is a capable fault in terms defined in Appei *ix A to 10 CFR Iart 100. This is the Altogether,41 piping component failure tests were com-first fault located cast of the Rocky Mountains that has pleted. Two piping systems were ruptured by high seis-been identified as capable. These results were transmit- mic-like loads, and one of these systems was retested.

ted to NRR in Research litformation Letter .No.151, Testing over 140 fatigue ratchetting specimens was com-

"Results of Meers Fault Investigation." pleted, and waterhammer pipe systems tests were per-formed.The results of this p;ogram are being used to re-In cooperation with the Nationa: Geodetic Survey, t,he vise the piping design critena of the ASMb Boiler and NRC is Fponsoring a Crustal Modon Network of 45 sta- Pressure Vessel Code.

tions coveritig the eastern two-thirds of the United States.

Highly accurate positions of these stations will be meas- The seismic capacity of most electrical equipment is gov-ured with the Global Po itioning System at intervals of erned by malfundion of re'ays. An evahtation of the exi.st-about 2 years. A first set of measurements was perfor med ing relay pst da9 bcsc has indicated that the schmi,: fra-this year. Results were very encouraging in that ac- gifity of a relay msydepend on various pa:ameters related caracies of a few parts in 108 were achieved. 'With this to the des!gr. or the input motion In particular, tne elec-rneasurement accuracy, errors over a baseline of 2,000 f ricr! mode, contact state, adjustment, chatter dernion kilometers amount to ontv a few centimeters. by acceptance limit, and the frequency and the direction ef remeasuring the stations ove'r a lcrgth of time, it should the vibratica input have been considered to influence the be passible to directly detern ino crustal motion in the relay fragility level. In order to investigate the effect of Eastern and Central United States. Such riersurements most of these parameters or. the seismic fragility level, 9 NUREG-1266

1 Component Integrity

. NRC has conducted a relay test program. A total of 46 tance to high dynamic lads was demonstrated, and specimens of 19 popular relay models from three manu- valuable dynamic responw m were obtained.

facturers have been tested. For 10 models, more then one specimen has been used to study the consistency of re- 3. Se smic tests of a 1/2.4 scale model of a PWR piping suhs. loop were performed on the large shaker table in Tadotsu, Japan, in collaix) ration with the Japanese The test results confirm that the performance of a par-ticular relay specimen is significantly influenced by vari.

Ministry for International Irade and Industry ous parameters. Since there was a wide variation of the (Mill). The experiment, carried out m April 1988, capacity levels of different specimens of the same model, was successful in exciting the modified piping loop testing of multiple specimens appears to be necessary for model well into the inelastic range. Both ratchetting establishing a fragility level with reasonable confidence. and dynamic crack growth occurred.

Over 2900 tests were performed during this test program and a vast amount of ata was obtame . 1.433 Seismic Design Margin Methods

'Ihe NRC's participation in international seismic test pr* Seismic margins review procedures have been found to be grams is beneficial m the sharing of both research re- an effective and efficient way to assess the capability of i;ources and different perspectives on seismic design is- nuclear power plants to safely withstand earthquakes sues. Ihe pooling of resources allows the development of larger than their design basis level.The results of scismic bigger, rnore complex, test articles. These larger-scale margins evaluations can be used to answer questions re-tests are an important element in the validation of meth- gardmg the effects of higher scismic hazard at a site or to ods to predict the seismic response behavior of nuclear identify what systems and plant functions are most relied plant systems. upon to lessen the probability of core damage resulting

. . from earthquake events.

The NRC is cooperatmg m. three such programs:

I

1. A soil-structure interaction (SSI) experiment has , e sc'i n ic argi rvw I tch Uni I n IN8 na been completed at a site m Lotung,'Iaiwan.This ef* dition to review and comment by the NRC staff, the NRC fort is in collaboration with the Electric Power Re- is contributing by sponsoring an independent flatch Peer search Institute (EPRI) and the Taiwan Power Com- Review Group (five expert consultants with expertise in pany. The objective of the experiment is to obtain plant systems and seismic evaluation) and a separate fault measured earthquake response data from a soft soil tree analysis to compicment the success path analysis site that will validate the accuracy of analytical pre. sponsored by,EPRI. Georgia Power is, performing its dictions of SSI effects. Fourteen earthquakes have USI A-46 (seismic equipment qualificatto been recorded, three of which exceeded the Richter P.I ants) review of components and tanks m,n m o conjunction with the margm review. The lessons learned from this magnitude of 6.0. sharing of efforts is of great interest to the NRC staff.

2. A piping loop was subjected to severe seismic load-ings at the Heissdampfreaktor (IIDR) facility in Other cation ofseismic margin study a BWR systems tictivities in 1988 include (NUREG/CR-507 , up- the[)ubli Kahl, Federal Republic of Germany, in collabora- dating the PRA f' agility data base (Revision 1 of tion with Kernforschungszentrum Karlsruhe (KfK). UCID-50571), and ;he first phase of a study to compare Although failure of piping did not occur, the resis- two methods for pr.:dicting component scismic capacities.

i l

l t

NUREG-1266 10

i l

2 PREVENTING DAMAGE TO REACTOR CORES His program encompasses research pertaining to the op- llabcock and Wilcox Company. Ilecause of the accident, erations of the reactor as a system, including controlling the NRC reviewed the analytical predictions of feedwater power level, maintaining core cooling and heat removal, transients and small-break 1 OCAs for the purpose of en-and maintaining proper coolant temperatures and pres- suring the continued safe operation of all operating reac-sures under both normal and abnormal conditions of op- tors. As a result of the review, nuclear steam supply sys-eration. 'Ihis program also includes consideration of op- tem vendors and fuel suppliers were required to provide l crator actions as an integral part of the reactor system. A experimental verification of the various modes of single- i complete understanding of the reactor operating as a sys- phase and two-phase natural circulation predicted to oc-tem makes it possible to define the conditions of opera. cur in each vendor's reactor during small-break LOCAs.

tion that prevent core damage and also actions to mini- This requirement is delineated in NUREG-0737, "Clari- ,

mize the consequences of a core damage event should one fication of TMI Action Plan Requirements." (Task Ac- I occur This research program emphasizes severe accident tion Plan Item II.K.3.30," Revised SBLOCA Methods to prevention and mitigation by enhancing the understand- show compliance with 10 CFR Part 50, Appendix K.")

ing of both plant and human behavior during accidents The expenmental data on small-break LOCAs from the and transients. This information is used to ensure that Semiscale and the LOFF facilities were applicable to the regulatory requirements exist that suitably ensure that Westinghouse and Combustion Engineering (CE) de-plant equipment, operating procedures, and training of signed reactors but not to the liabcock and Wilcox (B&W) personnel can deal with operating events and prevent se- design. This is because the B&W once-through steam rious accidents or can mitigate the consequences of an ac- generator (OUG) presents a significantly different con-cident, should one occur. figuration from U-tube steam generators employed in the Westinghouse and CE designs. Neither the Semiscale nor 2.1 Plant Performance the LOFr facilities modeled the unique B&W hot leg configuration of the O'l SG and, as a result, did not simu-late the appropriate natural circulation conditions. In 2.1.1 Statement of Problem particular, there was uncertainty about the effects of two-

. . phase flow, noncondensible gases, anJ the validity of the A wide range of reactor plant design variations exists m boiler-condenser mode of heat removal. In addition, the the United StaMs, and the safety of these plants for a wide hydraulic stability, effects of high point vents, and internal range of normal and abnormal operations must be en- reactor vessel vent valves were items of interest.

sured. The NRC is required to independently assess cact licensee's assertians and performance of his responsibil- Section 50.46 of 10 CFR Part 50 requires that calculations ity to design, construct, and operate a reactor with respect be performed to show that the emergency core cooling to the safety of the plant for the complete spectrum of will adequately cool the reactor in the credible operating conditions and events. systems (ECCS}A. Appendix is to 10 CFR Part event of a LOC 50 setj forth certain required and acceptable features that the 1 NRC's task is difficult because straightforward testing of evaluation models used to perform these calculations all transients in all plant design vanations would not be must contain. In many instances, these calculations result technically and economically feasible. On the other hand, in limits on reactor operation (c.

straightforward and exact theoretical analyses of a reac- order to comply with the 2200*h.'cbcak h> cal power dding temperature tor's fluid flows would take too long to compute because limit and other limits of 9 50.46.These limits may restrict of the complexity of heat exchange between reactor com- the total power output and optimal operation of many re-ponents, water,and steam,aswellasbecauseof the mov- actors (e.g., most Westinghouse plants)in terms of cffi-ing mechanical interfaces in pumps and the extensive baf- cient fuel utilization, maneuvering capability, and surveil-fle surfaces in the primary loops. lance requirements. Removing unnecessary restrictions on operation will allow increased U.S. electricity produc-As a result, the NRC must combine, in a complex under- tion, worth several h undred million dollars a year, without taking, limited experimental data, much of which is less loss of benefit to the public health and safety, than full scale, and limited calculational capability into a firm technical basis for evaluating oesign basis accidents, the safety implications of actual events in operating reac-tors, and hypothetical transient scenarios determmed t,o 2,1.2 Program Strategy be major contributors to nsk as a result of probabilistic risk assessment studies and these operating events.Two The NRC has a daal complementary approach toward examples of the need fora tect'nical basis to evaluate de- achieving a finn technical imderstanding of the thermal-sign basis accidents are g;vea below for the TMI. type of hydraulic behavier of the reactor. The dual approach is smail-break loss-of coolant accident (LOCA) and for the analytical and experimental with feedback to the analyti-revised ECCS rule. cal models.The NRC starts by simulating the act ual reac-tor's continuous flow of heat and fluids with a computer-The Three Mile lsland Unit 2 (TMI-2) accident in March ized model consisting of many small discrete cells 1979 was a small-break I OCA that led to core damage. exchanging b eat. fluid, va por, kinet ic e nergy, and momen-The reactor system at TMI-2 was designed by the tum at each small but finite time step. Physical laws are 11 NUREG-1266

2 Preventing Core Damage used when possible to calculate all these exchanges. Em- postulated LOCA could be much lower than those calcu-pirically denved formulas, obtained from experiments, lated using Appendix K methods.1hus a large body of re-are used as necessary to account for such complex effects search was mitiated to provide a method both to estimate as friction between vapor and liquid. The calculations are the degree of conservatism in Appendix K calculations made for each time step and for each cell, in a manner and to determine to a reasonable extent the uncertainty familiar to animated computer games, except that reactor associated with the estimate.

models have many more objects (cclls) and these objects interact in a tightly coupled manner at every time step, re- 2.1.3 Research Accomplishments in 1988 quirmg many more calculations.

2.1.3.1 Multiloop Integral System Test (MIST)

Our reliance on the computer codes to provide predic- Program tions of reactor response with acceptable uncertainties depends on three levels of experiments and comparisons The MIST facility, located in Alliance, Ohio, is a scaled 2 of experimental results with code predictions. First are by 4 (two hot legs and four cold legs) model of a H&W basie experiments to derive empirical formulas for deter. lowered-loop nuclear steam supply system stem.'Ihe ex-mining phenomena within each cell. Next are separate- perimental data for MISTare used to assess the capability effect experiments to test the code's predictions for a sin- of the NRC and industry thermal-hydraulic codes in -

gle, complex component such as a steam generator. Third dicting the behavior of H&W transients. Specifical the are mtegral system tests that are used to evaluate the code data are sufficient to validate ll&W small-break L CA predictions of a complete reactor. The results of these calculational models.1his small-break LOCA data base tests provide feedback to correct the code and our under. satisfied the condition imposed by NUREG-0737 (Clari-standmg of the transients.Two examples of this strategy fication of TMI Action Plan Requirements) Item are given below for the TMI-type small-break LOCA and II.K.3.30, which requires that small-break LOCA calcula-for the revised ECCS rule. tional models be compared to applicable data.

Since December of 1987,58 tests were conducted in the Since early 1980, discussions have been ongoing between MIST facility. These tests investigated the thermal-the NRC and hcensees of B&W plants relative to various hydraulic behavior in MIST for small-break LOCA tran-licensing issues. In September 1982 an mdustry/ govern- sients, steam generator tube rupture transients, feed and ment group known as the Test Advisory Group (1 AG) bleed recovery procedurcs, the effects of noncondensible was formed specifically to address these issues. The TA(; gas and reactor coolant puit p operation on transient pro-consisted of representatives from the NRC, H&W Own- grcssions, and strategics '; cope with station blackout.

ers G roup (H&WOG), Electric Powei Research Institute Toledo Edison Company 3 dependently funded three ad-(EPRI), and B&W. lhe TAG responsibility was to iden- ditional MIST tests to obtain data to verify the tify expenmental data needs, existmg data, and how well RELAP5/ MOD 2 code, a best-estimate thermal- hydrau-they addressed the technical issues related to small-break lic code used by the utility to support a design change at LOCA and natural circulation and to recommend future the Davis-Hesse (Ohio) nuclear power plant. The data programs. The TAG successfully completed the work in compared well with RELAP5/ MOD 2 analyses. This en-June 1983 and defined an Inte ral stem Iest(IST)pr ~

gram consisting of purchase o , an selected benchmark ables the utility to resolve NRC's concerns on the design change at Davis-Hesse. In 1988, the data from four MIST analyses of, GERDA test data,: a GERDA upgrade test tests were used to assess TRAC-PFl/ MODI code capa-(program known as the Once-Through Integral SystemOTIS) bilities. The TRAC-PFl/

program;MODI code is a best-estimate the development of a facility specifica' thermal-hydraulic code used by NRC to predict the be-tion, design, and associated test program of the 2 x 4 H&W havior of PWRs during transients. In 1989, data analyses loop known as the Multi-Imop Integral, System Test for the 58 MISE tests, as well as code analyses, will be (MIST) program. All these were to be carr ed out at the published as NRC reports.

H&W Alhance Research Center. Ihc culmination of the TAG cffort was the execution of a tri-party agreement in June 1983, including NRC, EPRI, and H&W to carry out 2.1.3.2 ECCS Rule the IST program. On September 16,1988, the NRC amended its regula-tions to allow the use of alternative methods to demon-In the 1973 Commission opinion published with the strate that the ECCS would protect the nuclear reactor ECCS rules, the staff was required to determine whether core during a postulated design basis loss-of-coolant these ECCS restrictions were more stringent than neces- accident (LOCA). The Commission took this action be-sary. So was begun an extensive experimental and cause research, performed since the former rule was writ-theoretical program of ECCS research by NRC, DOE ten in 1973/74, has shown that calculations performed (including AEL and ERDA), the U.S. nuclear industry, using the previous requirements resulted in estimates of and foreign institutions. The goal was to develop best- coolmg system performance tht are significan ly more estimate calculational models of LWRs during LOCAs, conservative than estimates based on the iniproved ~

providing a bese of knowledge that would penn;t the knowledge gained through this research.The old sppen-removal of the artificial, conservative rules of Appendix din K methods are conservative, but they do not resuh in K.The objective was to canfirm that the methods speci- accurate calculation of w hat would actually occur in a nu- '

ficd in Appendix K were highly conservative end that tne cleat power plant during a LOCA and may result in less actual cladding temperatures that would occur during a than optimal ECCS design and operating procedures. In NUREG-1266 12

2 Preventing Core Damage addition, the operation of some nuclear reactors was be- 2.2.3 Research Accomplishments in 1988 ing unnecessanly restricted by the rule, increasing the cost of electricity generation. This amendment, while 2.2.3.1 Iluman Factors continuing to allow the use of current methods and re-quirements, also allows the use of more recent informa- In May 1988, SECY-88-141 (Iluman Factors Initiatives tion and knowledge to demonstrate that the ECCS would and Plans)was submitted to the Commission. That paper protect the reactor during a LOCA. The amendment, described the NRC's human factors programs and imtia-which applies to all appliumts for, and holders of, con- tives and included an initial version of the Human l' actors struction permits or operating licenses for light-water re- Regulatory Research Program Plan (1IFRRPP). In Octo-actors, also relaxes requirements for certain reporting ber 1988, SECY-88-294 (11uman Factors Program) was and reanalyses that did not contribute to safety. submitted to address seven items related to the agency's human factors programs and initiatives and indicated that a revision of the IIFRRPP would be submitted in early 2.2 Human Performance 19g9, Also during 1988, research efforts were conducted that:

2.2.1 Statement of Problem (1) closed lluman Factors Generic Issue I.D.4, Control R m Design Standards, and (2) resuhed in coordinating A large fraction of all safety-related events re orted at research needs with other NRC offices and revismg the nuclear wer plants continue to be attributed to human human factors research program.

error. hods and data are needed to identify, system-atically set priorities for, and suggest solutions to human performance issues in the operation and maintenance of 2.2.3.2 Organizat,oni and Management nuclear, power plants during normal, transient, and emer- Research is directed toward credible modeling tech-gency situations. niques, data-gathering instruments, performance meas-urcs for systematically analyzing nuclear power plant or-2.2.2 Program Strategy g niz tion and emergency, andaccimanagent operatmg conditions.ement performan The human factors and reliability assessment research program has three objectives: (1) to broaden NRC s un- Accomplishments with other Federal during 1988 scientific, included (l) an regulator adreement military derstanding of human performance and to identify causes agencies to spmsor National Research (.y, rescarcn ouncil an of human error; (2) to accurately m,casure human per- on orgamzational effectiveness, (2) descriptive tech-formance for enhancmg safer operations and precludmg piques for modeling the operating characteristics of nu-critical errors, and (3) to develop the techmcal basis for clear power plant organizations during normal operating requirements, recommendations, and guidance related to conditions, (3) an imtial set of standardized instruments human performance. (survey, observational) for gathering organization and management information to support performance moni-The human factors regulatory research program is di.

vided into five interrelated program elements: (1) Per- toring and4)p(robabilistic plant level, publication of assessments NUREG/CR-5241 of report-personnel at the sonnel Performance Measurement,(2) Personnel Subsys- mg development and validation of programmatic per-tem, (3) Iluman-System Interface, (4) Organization and formance indicators in the area of plant maintenance and Management, and (5) Reliability Assessment. The pur- identification of candidate indicators in the area of per-pose of the Personnel Performance Measurement cle- sonnel training, and (5) a workshop from which emerged ment is to develop enhanced methods for collecting and recommendations for short-term and long-term research managing personnel performance data related to the ef- on organization and management performance in plant fect of personnel performance on the safety of nuclear and utility settings and, if necessary, on external entities operations and maintenance. Personnel Subsystem re- such as the NRC and public utility commissions that may search will broaden the understanding of such factors as have a significant mediating impact on plant personnel staffing, qualifications, and training that influence human performance.

performance in the nuclear system and will develop infor-mation necessary to reduce the negative impact of these 2.2.3.3 Reliability Assessment influences on nuclear safety. Research in the Human-System Interface element will provide the measures for Specifically reliability assessment research is directed to-evaluating the interface between the system and the hu- ward probabilistic data, data management systems, analy-man user from the perspective of safe operations and sis methods, and integrating procedures for (1) doing I maintenance. Organization and Management research quantitative and qualitative human and system reliability will result in the development of tools for evaluating or- analyses and integrating their rsults inte PRAs (2) mom-gaaization and management issues within the nuclear in- toring programmatic and risk. based performance trends, the Reliability Assessment element (3) establishing reliability -based and risk-based guidelines dustry. And, lastly,linary research that willintegratefor includes multidiscip hu-implementing and monitoring the effectiveness of man and hardware considerations for evaluating reliabil. plant reliability progrhms, and (4) sysicmatically empk>y-ity and risk in NRC licensing, inspection, and regulatory mg PRA processes to address safety issues of concern to decisions. regulators.

13 NUREG-1266

2 Preventing Core Damage Accomplishments durin3 1988 included (l)an agreement consequences of severe accidents can be greatly influ-with other Federal scientific, regulatory, and military enced by nuclear power plant operators and that many agencies to sponsor National Research Council research vulnerabilities to severe accidents can potentially be initial criteria eliminated by proper operator actions. The TMI-2 acci-on causal mechanisms for equating of human human tasks inside error,(2)de the nuclear and outsi dent and other abnormal occurrences in nuclear power industry to allow for use of error probability data from plants have shown that operators do not stand idle but ac-other industries as bounding or anchor values in nuclear tively intervene in attempts to control the event. If opera-l power plant risk studies, (3) publication of NURl!G/ tors are provided with proper guidance and training to CR-4639 (5 volumes) and dissemination of computer take beneficial actions when needed and, most impor-software (Version 1.0) to remote users of the Nuc! car tantly, refrain from actions that can have adverse effects, Computerized Library for Assessing Reactor Reliability the consequences of a severe accident can potentially be (NUCl.ARR), i.e., a data management system and data significantly reduced. Since many accident management store of human error probabilities and hardware compo- strategies do not involve significant plant design changes, nen t failure rates to support reliability and risk st udies,(4) substantial safety benefits can be quickly achieved by en-preparation of NUREG/CR-5323 reporting develop- suring proper operator actions.Thus, the initiation of ac-ment and partial validation of risk-based indicators in the cident management programs at operating plants is a logi-arca of saf ety system function trends,(5) partial validation cal result of the IPE process.

of an artificial intelligence-based Cogmtive linvironment l Simulation (CliS) and Cognitive Reliability Analysis 2.3.2 Program Strategy robabihstic analysts of Technique (CRiiATli)of for cognition,(6) publication NURIdomg p!G/CR-5200 reporting RIiS has been given the responsibility for the implemen-an evaluation of the risk impact of varying NRC require- tation of the IPli.This implementation has involved de-ments for surveillance testing and limitmg conditions of velopment of guidance for performance of the IPli, pre-operation, and (7) closed out TMI Action Item H.C.4 re- paring a generic letter to plant operators requesting the garding application of reliability engineering to maintain IPfi, and developing review plans and eventually review-operational safety throughout the plant life cycle, i.e., de- ing the results of the IPli submittals in cooperation with veloped and published as NURI!G/CR-4618 guidelines NRR. The requirement to correct any identified plant-for plant operational reliability programs. specific vulnerabilities not voluntarily corrected will be determined by the backfit rule. Accident management is not required as part of the IPII process but was high-2.3 Accident ManaEement and lighted in the IPh generic letter as a future requirement Individual Plant Examinations that will make use of the results of the IPli process. The consideration of severe accident vulnerabilities due to ex-t ernal hazards (earthquakes, flood, wind, etc.) will also be

- 3.1 Statement of Problem deferred until the staff considers how best to consider these hazards. Consideration of seismic hazards by the A severe accident in a nuclear power plant can be defm.ed IPli process must be coordinated with a number of other as an event m which the core is damaged and there is a seismic regulatory activities.

potential for reicase oflarge amounts of radioactive fis-l sion products. In the Commission policy statement on in support of the Severe Accident Policy Statement im-severe accidents in nuclear power plants issued on plementation and the IPlis, Ri!S is initiating an accident August 8,1985 (50 FR 32138), the Commission concluded management research program.This program has the ob-that existmg plants pose no undue risk to the public jective of providing an independent evaluation of selected health and safety and that ther.:is no immediate need for accident rnanagement strategies to provide a technical ba-generic rulemaking related to severe accidents. However, sis for staff review of licensee accident management pro-based on NRC and industry experience with plant-spe- grams. This program will also transfer the results of se-cific probabilistic risk assessments (PRAs), the Commis- vere accident research to the industry in a form that can a systematic examination be practically applied and will attempt to demonstrate the identi of carhto severe neratnhties existingacci plant to, dents.The pokey statement m-sionisconvinced of potential benefits of an accident management program.

dicated the intent of the Commission to take all reason-able steps to reduce the probability of a severe accident 2.3.3 Research AccomI)lishments in 1988 and, should a severe accident occur, to mitigate its conse-The Office of Nuclear Reactor Regulation (NRR) and quences to the extent possible. As part of the implemen- RIIS have prepared an accident management program tation ef the Comimssion's Severe Accident Policy State-plan, w hich is a key element in NRC's overall integration ment, the staff has requ red individual plant examinations plan for closure of severe accident issues. The accident

(!Plis) of all existing plants to identify any plant-spmfic management program will be a closely coordinated effort vulnerabi'ities to severe accidents with industry, in which industry and NRR will survey util-ity accident management capabilities. Based on this mfor-Much of the work performed to implement the Severe mation and on RI!S state-of-th<e-art survey of research in-Accident Policy Statement has focused on research into formation on accident management strategies NRR and phenomena that would occur during severe accidents and RI!S will define the scope and attributes acceptable for methods to systematically discever vulnerabilities for se- industry accident management plans. Using this guid-vere accidents. 'Ihis work has shown that the causes and ance, each utility will be expected to develop accident NURf!G-1266 14

2 Preventing Core Damage management plans and capabilities. Closure for indt:stry During 1988, the NRC staff completed plans and recom-is for each licensee to have an accident management pro- mendations for the IPEs, an integrated systematic ap-gram framework in place that can be expanded and modi- proach to examine each nuclear power plant now operat-fied to accommodate new information as it is developed. ing or under construction for possible significant risk The RES program will closely support this closure effort contributors that might otherwise be overlooked.This ef-and also develop comprehensive accident management fort included issuance of a generic letter to all licensees to strategy evaluation and insights based on the IPE process, initiate the IPE process and a staff evaluation of IDCOR the Severe Accident Research Plan, the Containment IPE methodology. A draft guidance document to assist li-Performance Initiative, the human factors program, and censees performing the IPE and indicating what the NRC cooperative programs bei.?- developed with industry and staff will expect in the IPE submittals was also issued.

foreign participants.

15 NUREG-1266

3 REACTOR CONTAINMENT PERFORMANCE AND PUBLIC PROTECTION FROM RADIATION The basic criteria for licensing nuclear power plants for ments, out-of-reactor experiments, examination of speci-construction and operation are judged to have provided a mens from TMI-2, and analytical model development. (2) considerable safety margin, affording the public protec- 'Ihe research on natural circulation in the reactor coolant tion from radiation even under severe accident conditions system includes scaled mock-up reactor ccx>lant system such as thcse that occurred in 1979 at Three Mile Island. experiments and detailed analytical studies performed The physical possibility of even more severe accidents with joint funding from the Electric Power Research In-than that at 'IMI is, however, recognized. Considerable stitute (EPRI) and NRC. (3)The steam explosion work is progress has been made in recent years in understanding currently limited to the development and validation of a the underlying physical and chemical phenomena that can model for fuel-coolant interactions for use in accident occur in a severe accident. Such information is essential as analysis. Some additional stcam explosion work is a basis for assessing potential safety improvements and planned because of the importance of this phenomenon for making decisions on whether or not particular im- to accident management. (4) The research on fission provements are warranted. As pointed out in the Com- prod uct behavior and chemical form is being cond ucted to mission's Severe Accident Policy Statement, such deci- determine fission product releases from fuel at high tem-sions should be based on a combination of engineering fission prod-judgment (i.e., a deterministic method of setting and as- peratures uct chemistrywhen and itscore effectgeometry on retentioniswit changing,hin the plant, sessing safety margins) and the application of probabilis- and the behavior of iodine fission products in the pres-tic risk assessment techniques based on up-to-date ex- ence of aqueous pools. Much of the work on the transport perimental information to evaluate the likelihood of the of fission products 4 acrosols has been completed, and occurrence of rare events. this program represents a smaller fraction of expendi-tures than in previous years.

In similar fashion, the same underlying science and deci-sion process can be applied to reevaluations of existing 3.1.3 Research Accomplishments in 1988 safety systems and regulatory requirements to determme if particular conservative ass,umptions have been war" 3.1.3.1 Core Melt Progression and liydrogen ranted in terms of risk reduction. Generation 3.1 Core Melt and Reactor Coolant In-vessel core melt progression research is concerned with the state of the reactor core in a severe reactor acci-System Failure dent from the time of core uncovery up to the time of re-actor vessel meltthrough.This research also includes the 3.1.1 Statement of Problem determination of the mode of vessel failure. Sensitivity studies have suggested that the uncertainties in the state Major uncertainties in estimating the probability of early of the core debris at the time of vessel failure produce the containment failure and radioactive releases of source greatest uncertainties m the ex-vessel phase of an acci-terms appear to be significantly related to uncertainties in dent, including core-concrete interactions and direct con-the in-vessel progression of the accident while the fuel tainment heating.The state of the core in core melt pro-material remams m the reactor pressure vessel. Accident pression is also the primary determinant of in-vessel management strategies will also be heavily dependent on nydrogen generati,on, fission pnx!uct and acrosol genera-the in-vessel phenomena covered in this program ele- tion and attenuation, explosive and non-explosive rapid ment. steam generation, and the potential for successful recov-cry actions in accident management.

This program addresses (1) the heatup and meltdown of the core, (2) hydrogen generation, (3) fission product re. Calculations and comparison of analytical models with lease and transport within the reactor coolant system,(4) Power Burst Facility (PHF) and Annular Core Research the natural circulation of hot gases that might cause early , Reactor (ACRR) data and,with what,we know of the failure of a pipe or steam generator, (5) energetic fuef. FMI-2 accident are contmumg. Analysis of the PWR re-coolant interactions that occur as molten debris falls into actor vessel failure by core melt attack with current mod-the water-filled lower head or as water is added to molten cls cannot determme which occurs first-failure of local vessel pene,trations or gross vessel failure by creep rup-debris, (6) the ture of debris composition, at the morp(hology, time of vessel or reactor coolant and sys. tempera.

ture. Expenments and continued model deve,lopment are tem) failure, and (7) the mode of vessel failure. under way to resolve this important question. Assess-mer" development, and code improvement are contmu-

! ing. The mechanistic PWR codes have been applied ex-l 3.1.2 Program Strategy tensively to the analysis of complex core damage accidents and experiments in the U.3 and abroad. HWR NRC's rescarch effort in this program consists of four versions have been developed and are being tested. Ana-main activitics. (1) The in-vessel core-melt progression lytical support was provided to the German COR A out-and hydrogen generation work includes in-reactor experi- of-pile fuel damage expcsent mai are providing much 17 NUREG-1266

3 Containment Performance and Radiation Protection high-quality information for code assessment and im- from the reactor coolant vessel,,and the extent of fission provement under international cooperative agreements. product revaporization from the reactor coolant system.

Another code is being developed to estimate the partition ofi dine between the aqueous phase and the gas phase in 3.1.3.2 Natural Circulation the containment, the production of organic todide spe-Natural circulation in severe accidents is the buoyancy- cies, BWR suppression pool chemistry, and the extent,of todme revaponzation and resuspension from contam-driven steam circulation between the reactor core and ment surfaces and sumps. Taken together, these codes upper-plenum region of a vessel (in-vessel circulation) can address a spectrum of questions related to fission with or without countercurrent flows in the hot legs and steam generators (ex-vessc1 circulation).This kind of mul- product release and transport within the reactor coolant tidimensional flow may exist during the core uncove and system and the contamment, meludmg important nsk core melt period of certain severe accidents in a P R. If questions related to revaporization of fission products such flow should occur,it will provide a means of transfer-and other questions relate to offsite consequences.

ring the decay heat from the core to the upper. plenum . . .

Work is contmuing on validation tests at the Pacific structures, hot leg piping, and steam generator tubes. As a result, the reactor coolant system pressure boundaries Northwest I;iboratory (PNL) as part of an evaluation of may be heated to high temperatures, which challenge light-water-reactor engmeered safety feature system fis-their structuralintegnty. sion product retention effectiveness dunng severe acci-dents for plants with ice condenser contamments. The code was developed to estimate the extent of acrosol par-MElfROGflR AC calculations were performed for ana- ticle retention m the ice compartments of PWR ice con-lyzing the in-vessel circulation in the Su plant during a station blackout accident with the loss 07 auxiliary feed- denser containment systems. The test program includ,es the mvest ation of article attenuation m a PNL facihty water (the TMLIP accident). Comparative calculations were also performed for Surry using countercurrent flow that,inclu es a full- ength (48-foot arrangement of four information. These calculations all su ggest that either the equivalent ice basket columns one full-size central column surrounded by four half-size columns and four surge line or the hot leg connection at the vessel may fail by creep rupture from the high temperatures and res- qarter corner columns . Valuable msights are being ob-sures before the vessellower head fails. As a result, kigh- tamed concerning flow icid,s as well as particle transport and dynamics under conditions myolvm, g low flow rates pressure melt e'ection may not occur during the TMLIP nd the mixmg of hot air and steam with the head of cold accident in a westinghouse PWR. However, uncertain- air developed by the columns.

ties in these calculations are yet to be estimated or bounded, and future work is needed to validate the codes against data and to estimate or bound the uncertainties in 3.2 Reactor Conta,inment Safety the results.

3.2.1 Statement of Problem 3.1.3.3 Fission Product Behavior .

This program element provides m. formation on the major Fission products deposited on the reactor coolant system phenomena that produce high pressures and tempera-structural surfaces during a severe reactor accident may tures that have the potential to,cause containment fail-subsequently heat up these surfaces when they decay. ure. It is known from previous nsk studies, and from the The increase m surface terr.perature may result in the experiences at Chernobyl and Three Mile Island, that revaporization of the deposited fission products.The con. containment survival or even delayed failure has ,an a,Il-sequence may be an increase in the overall source term important effect on mmimizmg the release of radioactiv-I leaving the plant in case of containment failure or bypass. ity to the environment in the event of a core melt acci-l One of the factors affecting the extent of fission product dent. The phenomena most likely to produce high revaporization is fission product chemical form. The pressure and temperatures are (1) the high-pressure ejec-tion from the pressure vessel of fmely divided particles of chemical form {s) of a spectfic fission product influences molten core debris,(2) the generation of noncondensible the volatility oi that fission product and therefore its po.

, tential for revaporization. The phenomenon may be par. and flammable gases from the decomposition of concrete l ticularly important for delayed' containment failure acci. by hot core debris, (3) the direct thermal and chemical dents where the source terms are otherwise small and the attack by molten core debns on structures and engineered quantity of the revriporized fission products may become safety features, and (4) the burning or detonation of hy-significant. drogen and other gases produced in the vessel or in the containment.

At present, research is being con 0 acted to develop thco-retically based fission product chemistry, rnodels to pre- 3.2.2 Program Strategy dict fission product chemical forms dunng transport in the reactor coolant system and the containment. A NRC's research effort in this program element consists of mechanistic code is being developed to provide the capa- phenomenological research in three areas plus the devel-bility to estimate the quantities of fission products and opment and use of computer codes that combine the ef-acrosols released from core, the extent of their transport fccts of all the separate phenomena.These four research through the reactor coolant system, the inventory of activities address: (1) the interaction of molten core de-radionuclides available for release once debris is expelled bris with structural concrete, including the ablation of NUREG-1266 18

l 3 Containment Performance and Radiation Protection I concrete structures, heat transfer to structuresin the con- mains tainment, the generation of flammable and nonconden- slump,and pressurized. Ieft unmitigated, collect at the bottom of the reactorthe core vessel. If will me sible gases that pressurize the containment, and the gen- molten core materialattacks the bottom head of the reac- ,

cration of acrosols, including fission products, (2) direct tor and a breach occurs, the core melt will be ejected un- j containment heating by molten debus particles ejected der pressure. If the material should be ejected from the j from the vessel at high pressure and hydrogen production reactor cavity into surrounding containment volumes as '

resulting from oxidation of metallic debris being ejected fine particles, thermal energy would be quickly trans-by steam, (3) the transport, mixing, and combustion of hy- ferred to the containment atmosphere.The metalliccom- i drogen in the containment, including the potential for ponents of the ejected core debris can further oxidize in i detonation, and (4) the development, validation, mainte- air or in steam to generate a large quantity of chemical en-  ;

f nance, and application of so-called integrated codes that ergy and further pressurize the containment. This is i are capable of describing multiple phenomena that occur called direct containment heating (DCil).  !

in severe accident sequences ofinterest. While the scien-tific understanding of severe accident behavior is gener- A rogram was developed at Sandia to investi ate core ated in the phenomenological, program activities, the em- de ris dispersed at various scales. Tests at if20th and bodiment of that understandmg in the integrated codes often provides NRC with its most useful tools. All of the 1/10th linear scale have been completed, and in 1988 two experimental programs were in progress-one at Sandia baseline calculations for the NUREG-1150 risk study, for example, were performed with integrated codes devel- and the other at Brookhaven National I aboratory. Four oped in this program. tests were completed in the Sandia facility, and experi-mental results confirmed substantial pressurization and acrosol generation. Because of the complexity of the 3.2.3 Research Accomplishments in 1988 DCH problem coupled with the high cost of running large-scale tests in the Sandia facihty, separate-effect 3.2.3.1 Core. Concrete Interactions tests are bein,g performed at Brookhaven to address core debris dispersal. Transparent plexiglass models, In those severe accident scenarios in which the reactor 1/42-scale, of Zion, Surry, and Watts llar reactor cavities vessel fails, high-temperature core debris may fall into were constructed. Both water and Wood's metal were the reactor cavity where it interacts with structural con. used to simulate core debris. Experiments and analyses crete.The consequences of these thermal and chemical were initiated to determine whether there exists some core-concrete interactions may significantly impact con. reactor coolant system pressure below which ejection of tainment loading, the modes of containment failure, and molten core from the failed reactor vessel, will not pres-the radiological source terms. To characterize the threat surize the containment and challenge its integrity. Data to containment integrity and the nature of the ex-vessel are now being used to develop models for lmth lumped-releases, experiments are being performed, and mathe. parameter and finite-difference codes. DCII models and matical models are being developed and assessed. correlations have been developed and incorporated into a parametric containment performance code. DCH-A code has been developed that models the physical and specific models were incorporated into a mechanistic fi-chemical processes that occur when gas bubbles gener- mte-difference code to provide a detailed description of ated by the decomposition of concrete pass through the particle behavior to guide the selection of parameters for molten debris pool and break at the surface. Tests of the parametric calculations.

aerosol generation by mechanical processes and tests of aerosol generation by vapor-condensation were initiated, 3.2.3.3 flydrogen Combustion and data are being used to assess the code.The degree to which refractory radionuclides are sparged from molten

'Ihe hydrogen combustion program assesses both the con-debns depends m part upon the relative vapor pressures of the pool constituents. A refined model based on recent sequences and methods used to control or mitigate deflagrations, diffusion flames, accelerated flames, tran-hi h; temperature measurements of chemscal act ef icients is being prepared for incorporanon to them,ivitys tion c - from deflagration to detonations (DDT), and deto-code. nations that might be causcd by hydrogen burns in a severe reactor accident. A lumped-parameter computer code was developed at Sandia National 12boratories and A number of transitn! phenomena that may, occur in the has been used in the analysis of nuclear reactor accidents reactor cavity dunng, or closely following, pnmary vessel involving the transpon and combustion of hydrogen. A fai1ure fire now bemg mvestigated.11xpenments to s,tudy flame propagation model has been incorporated into this the hydrodynamic behavior of core debris have been amts- code. A three-dimensional finite-element analysis tool ated to determtne the manner m which it may spread and developed at los Alamos is used to benchmark th'e Sandia l relocate in the reactor cavity. Whether the,BWR Mark I code and to pmvide more detailed hydrogen transport steel drywell shell survives a core melt accident may de- and mixing calculations. The Sandia models have been as-pend upon such debris behavior.

l sessed using EPRI I;trpe-Scale Hydrogen Combustion Nevada Test Site expenments. The assessment of both 3.2.3.2 Direct Containment Ileating the Sandia and los Alamos codes continues with the use of the data generated from the large-scale hydrogen In certain reactor accidents, degradation of the reactor transport experiments performed at the IIDR facility in core can take place while the reactor coolant system re- the 1 ederal Republic of Germany.

I l

19 NUREG-1266

3 Containment Performance and Itadiation Protection Flame acceleration and deflagration-to-detonation tran- 3.3.3 Research Accomplishments in 1988 sition and detonation experiments were analyzed and documented. A review continues of the effect of elevated Activity has continued on a set of programs whose objec-l temperature and high steam concentration on the various tives are to provide the data base required for the qualifi-modes of combustion. A detonation propagation model cation of methods for predicting the response of LWit was assessed against limited data. Newly developed fbme containment buildings during severe accidents (those be-acceleration and DDT correlations were assessed against yond design basis events) and extreme earthquakes. This German, Canadian, and United States data. set of programs is examining the modes of containment failure that would result in the release of radioactive ma-terials beyond the containment boundary. These modes 3.3 Conta.inment Structural Integr.ty i include structural failure of the containment building, leakage through or past the penetrations (electrical or mechanical), failure of containment isolation systems, or 3.3.1 Statement of Problern failure of the basemat by the molten reactor core.

The major source of risk to the public from the operation 'Ihc preponderance of effort was devoted to developing a of nuclear power plants stems from accidents that lead to complete understanding of the results from a 1/6-scale a containment failure. The regulatory concern is that the model of a reinforced concrete containment that was failure modes and associated load levels for contamment tested to failure in July 1987. Post-test analyses centered structures cannot be predicted with any real confid,ence on the measurements of strain and displacement taken at by the methods used for design. This is especially so if the cach discrete pressure step to evaluate the accuracy of contemplated failure mode is localized leakage. Iloth as- pre-test predictions made using different analytical tech-sessments of the risk posed by loads outside the design niques. Nine organizations, including three from the basis and estimates of the effectiveness of proposed United States, three from the United Kingdom, and one mitigative steps require an ability to predict the way in each from France, Italy, and the Federal llepublic of Ger-which a contamment will fail. many, made pre-test predictions and participated in the post-test evaluation. An initial comparison of results took place in connection with the Fourth International Work-3.3.2 Program Strategy shop on Containment Integrity, held in Arlington, Va.,in June 1988. 'the proceedings of this workshop were pub-Research on containment failure modes is based on the lished in November 1988 as NUllEG/CP-0095. A joint observation that excessive leakage can occur, basically, report, highlighting lessons learned by comparing predic-from four sources: tions with results, will be completed early in 1989.

1. Failure of the shell, either the containment shell it. A full-size personnel airkick, obtained from a cancelled self in the case of steel containments, or the liner in nuclear ower plant, was tested at Chicago llridge & Iron scam and Development Gn!cr in I lainfield, Ill. Ihc the case of concrete containments'. objective of the tests was to obtam structural data on the behavior of an airlock, especially the scaling surfaces, un-

2. Leakage at large penetrations as a result of the in- der severe accident conditions. In the tests, several load clastic deforrnations and/or degradation of seals and cycles were applied to the inner door of the airh>ck. The gaskets; two most important h>ad cycles were; (1) temperature held at approximately 400 F with pressurization up to 300 psig and (2) temperature held at approximately

3. I.cakage at c!cctrical penetrations due to degrada- 800 F with pressurization up to 300 psig. No significant tion of materials under the high temperatures asso- leakage was observed past gaskets in the inner or outer ciated with accident scenarios; and doors of the airkick for the test environment of 400"F and 300 psig.

4. Leakage through valves due to pressure and tem- F r the test environment of 800 F and 300 psig, a large perature effects. portion of the gasket on the inner door was ejected from its groove at 150 psig and from thir point on the inner door Research Iclated to shell failure or deformations of pene- measurabic leakage was recorded. Ttic pressurization trations rests on analyses of and experiments on model continued to 300 psig, t.it no leakage past the outer door tests of actual containment designs. These tests involve was detected. Although leakage past the inner door may pressurization up to failure levels ur. der ambier.t tem- occur underihese conditions, the redundancy of the outer peratures. Since seal and gasket materials are adversely door precluded leakage to the outside environment be-affected by the temperatures associated with severe acci- cause the temperature of the outer door remained quite dents, separate tests focusing on the development oficak- moderate for all tests.

age are performed under pressure and temperature con-ditions, usually at full scale. Experiments to examine the In addition to the above tests, tests were Mso performed possibuity of developing leakage through electrical pene- on inflatable seals under severe accident conditions. In-tratien assemblics and valves also require experiments flatable seals are used to prevent leakage around the pe-under temperature and pressure conditions at full scale. rimeter of airkicks and are fastened to the outer edge of NUREG-1266 20

3 Containment Performance and Radiation Protection the airlock doors.The seals are pressurized with air to scal Westinghouse SP-90.This PRA was submitted as part of the gap between the door and the airlock bulkhead. In- an application for a Preliminary Design Approval for the flatable seals are either currently installed or planned for Westinghouse SP-90 standard plant desion.Two separate use in 11 commercial power, plant containment struc- reviews were conducted to cover both the calculation of tures. Tests performed so far mvolved both aged (radia- the frequency of accidents involving significant damage to tion and thermal)and unaged seals at room temperature the core and the calculation of the consequences should and at elevated temperatures representative of severe such accidents occur.

accident conditions.

Point lleach.This PR A was submitted in rebuttal to some

. staff calculations done in support of the decay heat re-3.4 Reactor Acc.i dent Risk Analysis moval issue.'lhis review concentrated on the differences in assumptions, methodology, and details of plant design between the staff and industry calculations.

3.4.1 Statement of Problem 1)iablo Canyon. In order to comply with a license condi-Probabilistic risk analysis (PRA) has been shown to be a tion, the hcensee for Diablo Lanyon (Cal.) has developed systematic and comprehensive method for identifying and ro a long-term seismic pro, gram. As a part of this program, evaluating the effectiveness of safety posed to reduce the likelihood and consequences 0improvements -

7nu-the hgensee is performmg a level 11 RA. llecause the wisnue port,on i of this work mvolves the development of clear power plant accidents. PRA is used by the NRC staff s m new I RA methodology, the staff reviewis proceed-for evaluating the level of ing as the van,ous stages of the i RA are being done. Ihis in a number of ways, includinhmts; for assessinf the mar-safety at selected operating p gins of safety in current requirements in light o the Com-review is still ongomg.

mission's Safet G,oal Pohey; for monitoring plant per- Ihowm Ferry. As part of NRC's review of the licensee's formance; and or identifyin otential improvements m application for restart, the e,taff audited the methods and equipment or operator rel 2 ity. results of the licensee's PRA.

3.4.2 Program Strategy 3A.3.2 Assessment of Severe Accident Risks In February 1987, the NRC issued the draft version of The reactor accident risk analysis research effort is ap. NURIiG-1150, Reactor Risk Reference Document, as plied in four ways.These include (1) providing xpert re- well as a series of supporting contractor repons, for pub-view of severe accident PRAs to assess the ns implica- I'c comment. lhe dralt report assessed the nsks from pos-tions of accident management strategics in order to .

minimize the release of radioactive material to the envi- sible core da, mage accidents m five U.S. nuclear power ronment during severe reactor accidents; (2) developing, [p ntsme fyn plants studied are hurry (Va.), Zion (111.),

verifying, demonstrating, and maintaining methods for (Miss. . lhe report discussed the imph) cations of the fiveSequo analyzmg the consequences of in-plant and offsite severe risk assessments on regulatory issues such as the techni, cal accident physical 3rocesses for use in risk assessment,and gulations and im-developing and demonstrating methods for quantifying bawsfor p esent eme ,ency,plannm p menta, tion of the ommission s ety Goal and Se-in risk estimat es and the relative contribu- vere Accident Policy Statements. I,wo NRC-funded re-the uncertaintfic tions of speci issue uncertainty to the overall uncer-tainty;(3) reassessing periodically the frequencies, conse-quences, nd risks of severe accidents m nuclear power Mhc t tw ta ed publis the American Nuclear Society sponsored and published a plants and performing eer review of review of the results obtamed; and ( ) developmg n, methods sk-based manage- used anddraft report.

ment tools capable of determimng the incremental risk While the review process was under way, the NRC staff reduction associated with proposed plant design and op- and sup orting contractors have been updating the five crational modifications and assistmg m the pnontization risk ana yses. These updates are intended to reflect the of efforts in inspection and hcensmg activities. present plant design and operating characteristics, im-prove the methods used, and incorporate new experimen.

tal calculational data on severe accidents resulting frem 3.4.3 Research Accomplishments in 1988 the research programs of NRC and others. At the present time, the analyses of core damage frequen"y have t>ecn l 3.4.3.1 Review of PRAs completed, with documentation of results iri progrms.

Analyses of containment performance and overall risk l

Probabilistic risk assessment is now used by the staff to are at praent still under way. Compiction of this work, support the resolution of a wide spectrum of regulatory related documentation, and the staff's summary r: port is issues. For licelised plants, PR As are sometimes volun- scheduled for the spring of 1989.

tarily submitted by licensees to support their specific means for resolving such issues. For advanced plants, aP' 3.4.3.3 New StalT Computer Tools plicants are required to perform and submit PR As as part of their overall license application. Reviews performed in In regulatory decisionraaking, it is often necessary to ask 1988 included the fallowmp: what impact a propased moWfication to plant hardwarc or 21 NURl!G-1266

3 Containment Performance and Radiation Protection procedures will have in terms of risk. Generally, the most radiation protection measures, and economic impact esti-appropriate way in which to answer such a question is to mates.

examme existing PRAs, change the affected parameters, recalculate the analysis, and observe the resulting change In July 1988, modifications to the MACCS model were in core damage frequency and public risk. Such calcula- suspended to permit use of this model in the second ver-tions are currently being done for prioritization of agency sion of NUREG-1150. An independent code verification resources and for regulatory analyses of generic safety is- exercise was performed on this version of the code.

sues and unresolved safety issues. Still other uses, such as targeting inspection acti,vities and prioritizing the imposi- 3.5 Application of Severe Accident '

tion of Multi-Plant Actions, are also emergmg.

Research The System Analysis and Risk Assessment,(SARA) sys- 3.5.1 Statement of Problem tem was conceived to address the rulemakmg needs de-scribed above and also to provide the NRC with reliability A severe accident in a nuclear power plant (as defined in data that are currently available only on large mainfram,e Section 2.3)is an event in which the core is damaged and computers. The development of high-performance mi- there is a potential for release of large amounts of fission cr,0 computers has provided greater capacities to interact products. Significant research has been performed on the with extensive data bases for a large number of users. likelihood, progression, and consequences of a severe ae.

During 1988, a, draft users' manual and executable code cident as discussed earlier in this chapter. Much of this module were given limited distribution, and a course was work has concentrated on the performance of the con-held to train staff personnelin the use of the code. SARA tainment during a severe accident, including potential was o,ne of the tools used by the generic issues program; enntainment failure mechanisms, and the ability of the and, m still another program, many outstandmg Multi- containment to mitigate the consequences of a severe ac-Plant Actions (i.e., MPAs that have been imposed but not cident.

yet implemented by the licensee) were analyzed and documented. SARA was also extensively appl,ied to per- 'Ihis program element provides for the application of the form sensitivity studies on the safety s,igmficance of results of severe accident research directly to the regula-changes in motor-operated-valve and circuit-breaker fail- tory process. Modification of the Commission's rules or ure rates. policies regarding siting, emergency planning, and con-tainment design are examples of areas m which the results in support of the NRC staff performance and review of of severe accident research may affect future changes.

PR As, a new, fast-running computer model for in-plant severe accident ana1y' sis has been developed. This 3.5.2 Program Strategy model-MELCOR (Version 1.7)-analyzes such acci-dents from initiating event (e.g., a pipe break) through Severe accident research has identified a number of in-core degradation and welding and containment failure sights concerning containment performance during a se-(i.e., if all core and containment protection systems have vere accident. These insights have included both failed). This code makes use of simplified versions of strengths and weaknesses of existing containment de-more comprehensive codes (e.g., CONTAIN), permitting signs. In some cases, identified contamment weaknesses analysis of a larger number of accident sequences ofim- or uncertainties in containment performance have raised portance in PR As. MELCOR Version 1.7 has seen use in concerns about severe accidents, particularly for BWR the staff's NUREG-1150 effort (described above) and in Mark I containments.

the staff ongoing PRA of the laSalle BWR. In parallel with use, validation exercises are being performed, com- The Containment Performance Improvement (CPI) pro-paring code calculations with results of experiments and gram systematically examines insights gained from severe the TMI accident. accident research to identify containment vulnerabilities and to identify potential improvements to correct vul-nerabilities. Because of concerns about Mark I contain-In coordination with the NRC staff wcrk on ments, the CPI program has imtially studied these con-NUREG-1150, a new model for asscssing the conce- tainments. However, studies of all types of containments quences of radioactive releases hw been developed.'Ihe ar e also plar.ned. lf potential improvements are identified taodel-MACCS (Version 1.5)-has the capabilhy to that provide sigmficar t enhancements to safet and are treat radionuchde ' releases leng for a short U.me or a shown to be cost-effective pursuant to 10 CF 150.109, prolor.ged period, including the effect cf change in the this program will recommend specific regulatory require-v.ind direction at the reactor dering the release, and to rnents.

sample the variability of prec:pitation intensity from the reactor site's meteorological data.

The CPI program is closely related and complementary to the individual plant examinations (IPEs) and accident MACCS incorporates newer or more realistic models for management programs (see Section 2.3). The CPI pro-health effect projections developed for NRC after publi- gram exammes containments for vulnerabilities on a cation of WASII-1400 (1975) anu BEIR-III (1980), long- generic basis so that utilities do not have to deal with ccm-term (chronic) radiation exposure from continued use of plex and highly uncertain severe accident phenomena on coritaminated environment, emergency response and an individual basis. The IPE, on the other hand, deals with NUREG-1266 22

3 Containment Performance and Radiation Protection plant-specific containment vulnerabilities unique to a On July 15,1988, the staff provided interim recommenda-

~ particular plant, which are not treated under the generic tions for Mark I's to the Commission as " Status of Mark I

. CPI program. Utilities have been requested not to make Containment Performance Evaluation," SECY-88-206.

changes to their containments under the IPE progra:n un-L til results of the CPI program have been considered and Although staff assessments are not yet complete, the fol-factored into their IPEs. lowing safety enhancements tentatively appear attractive in terms of their potential risk reduction capability as well '

The results of severe accident research in the area of as implementation costs: ( ) accelerated implementation source terms may have implications for other regulatory of existing ATWS and stat on blackout requirements,(2) areas, including siting, emergen planning, and a num- assurance of a backup water supply to the residual heat ber of generic and unresolved sa tyissues.This p ram removal and other contamment systems (e.g., drywell element also includes work to examme these areas or po- sprays) with normal and emer ency ac inde endent tential resolution of issues or changes to existin regula, pum tions as a result of severe accident research fin ings. the a t capability,(3)a,to open and reclose hardene i ,t independentiventing f ac power, improved reliability of the automatic epres-surizatio(n system, and (5) improved emergency opera 3.5.3 Research Accomplishments in 1988 procedures. The staff expects to complete its Mark I as-sessment and to make recommendations to the Commis-3.5.3.1 Emergency Preparedness On April 20,1987, the NRC published in the FederalReg. 3.6 Radiation Protection and Health ister(52 FR 12921 a proposed rule on emergency prepar- Effects edness for fuel cy e and other radioactive matenal heen-sees.The rule would apply to about 301arge faciliti,es.The 3.6.1 Statement of Problem facilities that would be required to comply with this regu-lation are those for which a release large,enough to re- The NRC must provide a radiation rotection program quirethesu portofoffsite res onseor anizationstopro- that ensures that workers and mem ers of the general  !

tect the pu ic was considere credibt . 'Ihe rule,would public are adequately protected from the adverse conse-r utre, among other, things, prompt notification of

, quences of ex sure to ionizin radiation from licensed o ite response orgamzations in case of a serious acci- activities. RE activities neede to support this gram dent, procedures and equi ment,for coping with the include developing radiation protection stand r s and emergency, and traming an exercises for response per- guidelinesforimplementingthemand lanning, develop- l sonnel. A final rule was submitted for Commission con- mg, and directing safety research stu ies to provide the  !

sideration on July 15,1988. imormation necessary for licensing decisions, inspection and enforcement activities, and the standards develop-On November 3,1987, a final rule dealing with nonpar- ment process. This includes analyzing available scientific ticipation of State or local governments in emergency evidence to evaluate the relationship between human ex-planning was published in the Federal Register (52 FR posure to ionizing radiation and ra ioactive material and 42078). In September 1988, the criteria for utility offsite the potential occurrence of both late and early radiogenic  ;

plannmg and eparedness (NUREG-0654; FEMA. health effects, including the radiation risk to workers and )

REP-1, Rev.1, upp.1) was published as a final report. the public, and estimates of th,e probability of increased I meidence of cancer and genetic effects. These analyses are used to provide base,s for severe accident consequence ]

On Ma 9,1988, the Commission published in the Fed- an lys s, robabilistic nsk assessment (PRA), the devel. .

eral er(53 FR 16435 a notice of proposed rulemak-ing that would establish)more clearly what erner cy opment o safety goals and, emerge cy plans, the identifi-cation of radiation protection prob ems, the alloca, tion cf lannin and pr redness requirements are neede or pn ntiesforregulatoryaction,andenvironmentalimpact fuel loading anbow blic plants. Approximately f70b r testing of nuclear comment letters were power assessments. Recommendations of such organizations as the International Commission o,n Rsdiolo cal Protection received and evaluated. 'It final rule was ublished on SJPtember 23,1988 (53 FR 36955). (ICRI and the National Council on Radi ion Protection and measurements (NCRP), Presidential guidance to Federal agencies, consensus standards, licensee perfonn-3.5.3.2 Mark I Containment Improvement Program ance indicators, cost and feasibility data, and tvailable techmcal infonnation also previde bases for develo ing As part of this program, a pub!ic workshop was held on regulatory arad technical documems related to rad ion February 24-26,1988, to discuss a number of issues asso- Protection for workers and the pubhc.

ciated with Mark I containment challenges, failure modes, and potential containment improvements with re- 3.6.2 Program Strategy searchers, industry representatives, and interested mem-bers of Ihe public. A major topic at the workshop was the The Commission's regulatory process requires that phenomenon associated with containment shell melt- changes to rules and guidance be systeraatiadly screened through, means to reduce its like'ihood, and methods to to ensure the t there is a substantia'l mcrease in public pro-mitigate its amsequences. tection and that based on analysis the costs are justified.

1 23 NUREG-1266

i 3 Containment Performance and Radiation Protection Realistic values of the dollar-per-pers(m-rem criterion sure. Specific health effects research areas are identified are needed for analysis to justify changes, but technology below.

gaps in knowledge associated with radiation health effects cause uncertainties in these analyses. The strategies of Five health effects studies are beiag used to support the this program are to identify and compensate for uncer- NRC reactor risk assessment study (NUREG-1150). The tainties m radiation risk coefficients used for health effect subjects of these studies include the early effects of in-estimates in PR As and regulatory decisions. (A feasibility haled beta-emitting radionuclides (NUREG/CR-5025),

study is under way on whether cellular and molecular ef- completed in November 1987; carly and continuing ef-fects data can reduce the range of uncertainty in health fccts of combined alpha and beta irradiation of the lung risk effects.) (NUREG/CR-5067) and early mortality and morbidity from high-level internal and external exposures-alpha

'Ihe Commission approved the whole body dosimetry ac. emitters (NUREG/CR-5198), completed in March and crepitation rule. At the same time, they directed the NRC August 1988, respectively; and early mortality and mor-staff to extend the rulemaking to include extremity bidity from high-level internal and external exposure-dosimetry. Therefore, the strategies of this program are and models for pul-to (l) improve regulatory performance for radiation pro, rnonaryemitters beta lethality and (NUREG/CR-5353)fter morbidity a irradiati tection by establishing measurement performance crite. mternal and external sources (NUREG/CR-5351), to be ria and accreditation programs in the areas of extremity completed in early 1989. The results of these studies, to-dosimetry, bioassay, and air sampling; (2) investigate el gether with information from other sources, were used to fcctive new measurement techniques for these areas:(3) develop the health effects models presented in NUREG/

establish the data base required for regulations; and (4) CR-4214, " Health Effects Models for Nuclear Power monitor specific indicators to detect improving and de. Plant Accident Consequence Analysis," which has been clining licensee performance. received in draft form and is currently being reviewed.

These models are used in NUREG-1150.

Federal guidance was approved by the President on occu-pational radiation protection. As a result of this new guid- NRC-funded studies published previousi indicated that ance, NRC regulations and regulatory guides wdl have to limits on exposure to soluble uranium sh uld be reduced be revised. Ihe strategies,of this program are to (1) mod- with a by majora factor sectionofoffive. TwoR-IV the HE )ublished reports,t alonhe basis report, provide ify radiation protection utdance and standards to be con-sistent with Presidential guidance on radiation protection for the ultimate decision. "Nephrotoxicity of Uranium" requirements and (2) contmue to monitor heensee per- (NUREG/CR-4951) was published in 1987, and in April formance indicators b using the Radiation Exposure In- 1988, NUREG/CP-0093, " Proceedings of the Meetin formation Reporting ystem (REIRS) program. on Ultrasensitive Techniques for Measurement o Uranium in Biological Sampics and the Nephrotoxicity of Uranium," was published.

3.6.3 Research Accomplishments in 1988 3.6.33 IIcalth Physics Technology 3.6.3.1 A1 ARA Center Performance in bioassay and extremity dose measure-The Brookhaven National Laboratory (BNL) ALARA ments is known from NRC-funded testing to be unaccept-Center, funded by NRC, continued its work on surveil- ably inaccurate. Accreditation programs are necessary, lance of DOE and industry dose reduction and ALARA Consensus performance standards were brought to an ad-research. In 1988 DNL published Volume 1 of NUREG/ vanced stage of com letion, and performance testmg CR-5158," Worldwide Activities on the Reduction of Oc- agamst these standar is m progress.

cupational Exposure at Nuclear Power Plants." The work re orts on a wide range of activities m plant chemistry, In the past, a troublesome aspect of the NRC ins cction c( alt reduction decontamination, remote tools, robotics, rogram arose wnen high bioassay results were o tained and adsanced designs. Ihe report concludes that world-for wort ers and it was necessary to determine whether th wide efforts to reduce doses are woikmg but regmre radionurJide intake exceeded regulatory limits. The in-take must be calculated, but the calculations v cre often

.continuin encouragement and further development.

Specifical y, it is necessary to focus on dose reduction for arameters, controversial ar.d methods used because of the varying bV the licensees. NURE models,b/CR-4884 nuclear power plant workers who receive the highest doses. " Interpretation of Bioassay Measurements," published ia July 1987, standardizerj this process. In 1988 a regulatory guide to endorse the results of this report is being devel.  ;

3.6.3.2 llealth Effects Research oped. j Significant research (which spans 1987 and 1988 and con- The NRC found that many nuclear power plants were be~ k tinues into 1989 and beyond) on to ics that relate to ginning to discover discrete radioactive articles attached health effects have contributed to N (C's ability to im- to the skin and clothing of workers. N REG /CR-4418, fi prove the assessment of risk associated with normal op- " Dose Calculation from Skin Contamination, Varskin }

crations and potential accidents in licensee facilities. This Code," published in August 1987, provided a standard. ]

research also ircluded studies on early effects and latent iyed, technically sound war of determining the dose from )

effects (cancer, gcnetic) resulting from radiation expa- f hese l' articles. In 1988, ttie NRC funded and reviewed a j l

NUREG 1266 24 l

3 Containment Performance and Radiation Protection draft NCRP report on the same subject matter. 'lhis re- guide provides acceptable criteria for bioassay programs port is still under review and will provide the basis and at uranium mills, based on current technology and recom-support for a future regulation or guidance or both for ex- mendations.

posure to discrete radioactive particles attached to the skin. Proposed rulemaking on 10 CFR Part 34, " Safety Re-quirements for Industrial Radiographic Exposure De-3.6.3.4 Dose Reduction and Standards Development vices," was published in the Federal Register on March 15, 1988. Most occupational radiation overexposure occur

'Ihe final comprehensive revision to 10 CFR Part 20, among industrial radiographer; the cause is often attrib-

" Standards for Protection Against Radiation,"was sent to uted to the design of radiography devices.This rule would the Commission m November 1988.This revision bnngs establish safety design requirements.

the NRC worker protection regulations into conformance with ICRP recommendations and the recent Presidential guidance. In addition, revisions to regulatory guides that Proposed rulemaking on 10 CFR Part 35, " Medical Use were necessitated by the rule change were begun in 1988 Of Ilyproduct Material," was published in the FederalRg-and will continue into 1989. ister m December 1987. This rule applies to radioactive aerosols and gases that may enter the atmosphere m NUREG-0713, Vol. 7, " Occupational Radiation Ex medical institutions; it provides for the maintenance of safe workmg conditions in a practical manner. ihe final sure at Commercial Nuclear Power Reactors and Ot r rule was published in 1988.

Facilities-1985," was published in April 1988. This report is used extensively by the NRC staff, by other government agencies, and by industry. Revision 1 to Regulatory Guide 8.22,"Ilioassay at Ura-nium Mills," was issued m August 1988. This guide pro-Regulatory Guide 8.32, " Criteria for Establishing a Trit- vides acceptable criteria for bioassay programs at ura-ium Ilicassay Program," was issued in June 1988. This nium mills based on current technology.

l l

i 25 NURI!G-1266

4 CONFIRMING SAFE'IT OF NUCLEAR WASTE DISPOSAL The NRC 's waste management research seeks to develop several other disciplines related to the management of and verify methods for predicting and assessing the per- high-level waste. The research combines theoretical formance of waste disposal facilities; evaluate and con- study with laboratory and field experiments to identify firm the data bases used in such performance assess- and quantify the physical processes that determine re-ments; provide technical support to the licensing staff in pository performance and quantify the uncertainties asso-their interactions with the Department of Energy (DOE) ciated with characterization and measurement of these and the States; and develop regulatory standards to sup- processes. All this work is integrated into an independent port the licensing of facilities and methods for the dis- systems performance assessment methodology. Effort is posal and management of high-level and low-level radio- also required to validate many of the models that underlic active wastes, the methodology. 'lhe ultimate goal of the NRC's waste management research is to provide the technical basis to 4.1 IligildMel Waste support the licensing staff's independent udgment as to the appropnateness and adeguacy of DO . s demonstra-tion of compliance with 10 LI R Part 60 and the EPA's 4.1.1 Statement of Problem HLW standard. In addition. NRC's waste management research seeks to provide technical support to the licens-The waste management issue involves both high-level ing staffin their interactions with DOh, the State of Ne-waste and low-level waste. The high-level waste (HLW) vada, and other participants and interested parties and to disposal policy for the United States is defined by the develop regulatory standards to support the licensing of Atomic Energy Act, the Energy Reorganization Act, the the disposal and management of high-level radioactive Nuclear Waste Policy Act, and the Nuclear Waste Policy wastes.

Amendment Act (NWPAA).'Ihe latter, signed into law in 1987, provides for the development of a geologic reposi- 4.1.3 Research Accomplishments in 1988 tory f,or the permanent disposal of high-level radioactive waste in the State of Nevada at Yucca Mountain and as- 4.13.1 Waste Package Performance signs responsibility for repository development to the DOE. IILW environmental standards development is the Investigating the performance that can oc expected from responsibility of the Environmental Protection Agency the waste form and waste package is essential to the (EPA), and the Energy Reorganization Act assigns the NRC's ability to independently evaluate DOE's demon-regulation of IILW disposal to protect public health and stration that the waste form and waste package comply safety and the environment to the NRC. with the containment and controlled release require-ments of 10 CFR Part 60. During 1988, the NRC spon-An HLW repository poses regulatory considerations and sored research on the integrated testing of HLW over-uncertainties related to waste emplacement, monitoring, pack materials in simulated repository environments.

and performance assessment that are unique in the his- NRC eompleted a 5-year study on the long-term perform-tory of the NRC. Much of this uniqueness stems from the ance of HLW packages.'Ihis work made definitive contri-type of facility, first-of-its-kind geologic disposal installa- butions to the technical base of NRC's HLW regulatory tion, and the fact that it will be placed m low permeability / program.

Iow flow geologic systems that have not been investigated previously because of their low economic value. NRC 'Ihe Japan Atomic Energy Research Institute (JAERI),

must have an independent capability to evaluate the under a cooperative research agreement with the NRC, DOE safety analyses and confirm whether long-term re- continued a series of experiments on the stability of HLW leases predicted by DOE will be within established limits, when it is in the form of glass and on the durability of The NRC research program objective is to provide the HLW containers in high-radiation environments. This technical capability necessary to evaluate DOE's site work complements the laboratory research studies being characterization activities as required by the NWPAA supported by the NRC of radioactive waste containe:s and to assesa DOE's license application when it is submit- and of the various forms of radioactive waste.

ted.

! 4.131 Geochemistry 4.1.2 Program Streteg 7he h RC has an active r escarch program m the vital field of geochemistry t related to the mmapmen; of HLW.

The yesearch proprun has been guided t'y the need to Work cominues t.t the University of California at provide the techmcal foundation for NRC development Berkeley or, the geochemistry of radioactive wastes m re-cf a set of regul?tions for the review and incensmg cf the posito:y environmems. In 1988, chemical reactions in tuff HLW repository. This framework for NRC review wdl al- and f,round water in the thermally affected area of an knv the formal hcensmg activities and the supperting re- HLW repository were investigated m the laboratory. 'Ihe search to be focused on the significimt techmcal issues. NRC is participating in an intemhtional field study at an ore body in Austraha to examine actual movement o~

At present, the NRC has active research programs in by- radionuclides. The first year of the study has been om-drology, geology, materials, science, geochemistry, and pleted successfully, with hydrologic tests well under way.

27 NUREG-1266

l 4 Waste Disposal Safety Work was completed at Oak Ridge National laboratory NRC licensing but also those States that regulate LLW on the chemistry of technetium in brines that could occur disposal under the Agreement State programs. This di-in a salt repository.'Ihe chloride ion was found not to af- verse user community makes the coordination and defini-fect technetium significantly.This work concluded all ef- tion of 11W research and the dissemination of products a fort for a salt repository. Research Information Letter much more complicated undertaking than that for the (RIL) 152 us issued reporting and summarizing the re- IILW program. l<urther, the fact that many States will be suits of a 5-year investigation of isotopic and geochemical the licensors and are looking to the NitC for technical methods of dating ground water to complement dynamic support in their licensing and regulatory programs drives hydrologic testing f or characterizing the hydrology of a re- the LLW research to be more prescr ptive and develop-pository site. Specific attention was given to applying the mental than is the HLW research program.

technique to the Yucca Mountain site.

4.2.3 Research Accomplishments in 1988 4.2 Low-Level Waste In May 1988, the NRC published a notice o.s proposed rulemaking on required geologic repository disposal of 4.2.1 Statement of Problem above Class C waste unless an alternative has been ap-proved by the Commission. Comments were being ana-Disposal of low-level waste (LLW) involves issues at the lyzed at the close of the comment period with a final rule forefront of technology, e.g., waste form and waste pack- expected in mid-1989. A proposed rule on criteria and age integrity and long-term retention of radionuclides in procedures for evaluating requests for emergency access the disposal facility environment. Research is required to to low-level waste disposal sites was issued in 1988, with establish regulatory criteria to permit sound evaluation of the final rule to be issued in early 1989.

proposals for disposal facilities and to ensure that all regulatory requirements, particularly those on radio- 'lhere is great interest on the part of States and State nuclide release limits, will be met.*lhe NRC will employ compacts in alternatives to shallow land burial, as cur-performance assessment methods for such evaluations. rently practiced. In 1988, the Idaho National Engineering Establishing these criteria in a timely rnanner is made laboratory completed research on the reliability of the more urgent and complex by two factors. First, the low- engineered components for alternatives to shallow land Level Radioactive Waste Policy Amendments Act of 1985 burial of LLW. Ihis research indicated that tbc cover (P.L 99-230) set a very tight time schedule for establish- component was most responsible for the reliability of the ing facilities within the various States. Second, the Str tes engineered alternative designs. Concrete is expected to and compas ts of States have chosen to consider alterna- play an important role in the engineered alternatives to tive disposa' methods to shallow land burial. Certain of shallow land burial, these alterndives must be critically examined by tightly focused rescirch to determine their acceptability and to An NRC-sponsored cooperative project with Atomic En-give guidanct to the States. crpy of Canada Ltd. (AECL) and the Battelle-Pacific Northwest l2boratories used data collected from 40 years The direction of the llW research program has re. of LLW waste disposal at AECL's Chalk River facility to

, sponded to h gislative mandates that resulted from an assess the capability of existing modeling techniques for predicting future LLW site performance. The work was I carlier histon af shallow land burial of wastes at a rumber done at two Chalk River sites with well-characterized l of sites for sev e ral decades. Vagu e and differing eriteria as to site suitab! ity, waste package design, etc., were used.

plumes containing measurable quantities of radio-Disposal criteria for LLW have evolved as experience, nuclides. Modelmg was done m two stages. Ihe first stage knowledge, public awareness, and political controversy used a lirnited data set (20 wells) typical of a site charac-nave grown. In particular, through the Low-Level Radio- terization program. The second stage used the complete active Waste Policy Amendments Act of 1985, the Con. data set for each site (over 120 wells per site). The results of the site characterization data set were then compared grcss has required the NRC to quickly complete the de.

velopment of sound technical bases for regulatory with the complete data set. Modeling results from the two decisionmaking regarding engineered LLW disposal stages were m reasonably close agreement, thus giving methods, so-called alternative 11W disposal. This confidence to the use of site characterization data for pre-change has broadened the scope of NRC 11W research. dieting future s,ite performance.The greatest uncertainty in this project mvolved estimating a source term and de-terminmg a realistic retardation coefficient for the Chalk 4.2.2 Program Strategy River soils.

At present NRC research in support of licensing activities The Brookhaven National Laboratory is continuing the for 11W disposal facilities is focused on addressing (1) NRC research project to study the use 'of high-density water entry into disposal units, (2) performance of waste polyethylene (HDPE) for LLW containers. Representa-forms and waste packages, (3) characterization of the tive samples of the material were subjected to the various LLW source term, (4) mechanisms for transport of environments expected in the waste forms and the radionuclides from the disposal units, and (5) the safety surroundings, e.g., culfates, acids, and gamma fields, in and performance of em;ineercJ enhancements and aber- order to study their failure and degradation mechanisms natives to conventional shallowland burial for f LW dis- and, if possible, to develop methods for predic;ing the posal. This research is intended to support not only the performance of the material over a period of 300 to 500 NUREG-1266 28

4 Waste Disposal Safety i

years. Results thus far indicate that HDPE can be either - management," under investigation promises to be highly ,

beneficially or adversely affected by gamma radiation, de- effective at sites subject to subsidence, or " bath tubbing,  ;

pending on the dose rate. because the disposal units have liners or are sited in low permeability sediment. liioengineering manageraent The University of California at Ilerkeley in cooperation uses impermeable pancis to enhance surface runoff, and  !

witti the Umversity of Maryland is field testing at vegetation is planted in narrow openings between the lleltsville, Md., a variety of covers designed to inhibit panels to remove through evapotranspiration the small 1 water percolation into waste disposal units. Covers under amount of water that passes through the panels. Such a  !

investigation include types being considered for future system is lowering the wate: levels in two large lysimeters l LLW dis at Ileltsville while mounded grass-covered lysir :ters ad- .!

cover,(2)posal sitesclay a compacted and include layer beneath(1) an aerosion compacted pro. clay jacent to these are experiencmg rising water sevels be-  !

tection layer (np-rap), and (3) a compacted clay layer cause of water percolation through the vegetative cover. .i above a conductive layer barrier (flow layer above a capil. The results of the lleltsville work can be applicable to any  ;

lary break). An additional design, " bioengineering disposal scheme employing carthen covers. l i

I I

l I

29 NURl!G-1266 l

h-5 RESOINING SAFETY ISSUES AND DEVELOPING REGULATIONS

. This program is directed toward the development of the During 1988, the NRC identified three new GSis and two techmcal basis and related regulatory requirements GSIs to be reprioritized, established priorities for 27 is-needed to protect the health and safety of the public from sues, and resolved 14 GSIs (see Table 5.1) and five USIs the risk resulting from the generation of electricity and (described below). In addition, six GSIs scheduled for the manufacture, use, transport, and storage of nuclear resolution were integmted into the action plans for reso-fuel and other radioactive materials. This program also lution as part of other GSIs.

supports efforts to ensure that proposed Commission regulations are adequate and that they are developed in The methodology and criteria needed to set prioritics for an efficient and timely manner. research activit es were developed.

5.1 Generic and Unresolved Safety 5.1.3.1 USI A-44-Station Blackout ISSLles The ioss of aii alternating curreni (ac) ciectric power (from both normal offsite and emergency onsite sources) 5.1.1 Statement of Problem is referred to as statien blackout. In the event of a station blackout, the capability to cool the reactor core would be dependent on the availability of systems that do not re-lheCommissiondirectedtheNRCstaff ority list of all generic safety issues, includingto f[ reapri-related gutre these ac power sources and on the abihty to restore issues, based on the potential safety significance and cost ac power m a timely manner.

of implementation of each issue. In December 1M3, the ,

listing was approved by the Commission.This guidance is The Commission amended its regulations on June 21, reflected in the NRC Policy and Planning Guidance, the 1988 (53 FR 23203) to require that light-water nuclear NRC Strategic Plan, and the NRC Five-Year Plan, power plants be capable of withstanding a total loss oia; power to the essential and non-essential switchgear buses for a specified duration of time. Regulatory Guide 1.155, 5.1.2 Program Strategy " Station illackout," which provides guidance on how to evaluate plant copmg capability for a specified duration.

A gencric safety issue is an issue that involves a safety con. was also issued with the station blackout rule (10 CFR cern that may affect the design, construction, or operation 9 50.63).

of all, several, or a class of reactors or facilitics and may have a potential for safety improvements and issuance of 1he station blackout rule and supporting regulatory guide new or revised requirements or guidance. Timely resolu. were developed in response to the staff's study of USI tion of these issues is a major NRC concern. A prioritiza. A-44, " Station Blackout." The technical findings are re-tion and management process has been established for ported in NUREG-1032 " Evaluation of Station Black-identifying important issues for immediate action, for out Accidents at Nuclear Power Plants." A regulatory eliminatmg non-cost-effective and duplicate issues from analysis was prepared and is reported in NUREG-1109, further consideration, and for keeping the Commission " Regulatory /Hackfit Analysis for the Resolution of Unre-and the publicinformed of the resolution of these issues. solved Safety Issue A-44, Station Blackout." Implemen-Currently, a 'oacklog of approximately 40 proposed ge. tation of the rule is estimated to limit the contribution to neric issues is awaitmg pnoritization. Strategies for this core damage frequency from station blackout initiated

. program are to provide timely prioritization of proposed events to approximately 1 in 100,000 reactor-years. The new genene issues, climinate the backlog of proposed is- estimated total cost for industry comphance with the rule sues (as resources permit), and issue penodic updates on is $60 million. This results in an overall cost / benefit ratio l the status and progress toward resolution of generic of about 2,400 person-rems per million dollars.The rule safety issues. will provide further assurance that a loss of both offsite and emergency onsite electric ac power systems will not adversely affect the public health and safety.

5.1.3 Research Accomplishments in 1988 The station blackout rule requires that all nuclear plants The NRC continued to use the methodology set out in the be capable of copmg with a station blackout for some 1982 NRC Annual Reportfordeterminingthepriorityof specified period of time after which, experience has generic safety issues (GS!s). In December 1983, a com. shown, there is a high probability of restoring offsite prehensive list of the issues subjected to this method was power. The period of time for a specific plant will be published in "A Prioritization of Generic Safety Issues" determined based on a comparison of the individual (NUREG-0933), which is updated semi-annually (sup- plant's design with factors that have been identified as the piements in June and December). The list of issues main contnbutors to risk of core damage resulting from includes TMI Action Plan (NUREG-0660) items and station blackout. These factors, which vary significantly unresolved safety issues (USIs).The results of the NRC's from plant to plant because of considerable differences in continuing effort to identify significant unresolved GSIs design of plant electric power systems as well as site-will be included in future supplements to NUREG-0933. specific considerations, include (1) redundancy of onsite 31 NUREG-1266

j

[ 5 SafetyIssues and Regulations Table 5.1 Generic Safety Issues Resolved in 1988 issd:e . Resolution - Date i , Number ' Title Product Resolved 86 Long-Range Plan for Dealing with Stress Corrosion NUREG-0313, 01/88 Cracking in BWR Piping Rev. 2; GL 88-01 93 Steam Binding of Auxiliary Feedwater Pumps GL 88-03 02/88 d I.D.4 Control Room Design Standard No Req. 03/88 II.E.4.3 (Containment) Integrity Check NUREG-1273 03/88 B-5 Ductility of Two-Way Slabs and Shells and Buckling No Req. 04/88 i i

Behavior of Stect Containments HF8 Maintenance and Surveillance Program Policy 05/88 '

Statement I.A.4.2(4) Review Simulators for Conformance Ru!e 05/88 -

43 Reliability of Air Systems GL 88-14 09/88 66 Steam Generator Requirements NUREG-0844 09/88 102 Human Error in Events Involving Wrong Unit or NUREG-1192 09/88 Wrong Train

125.11.7 Reevaluate Provision to Automatically Isolate Feed - NUREG-1332 09/88 I water from Steam Generator During Line Break  ;

II.C.4 Reliability Engineering . No. Req. 10/88 HF4.1 Inspection Procedure for Upgraded Emergency Inspectio'n 10/88 Operating Procedures Procedure 86-64 99- RCS/RHR Suction Line Interlocks on PWRs GL 88-17 11/88 l

emergency ac power sources (i.e., number of sources 5.1.3.2 USI A-45-Shutdown Decay Heat Removal minus the number needed for decay heat removal);- Requirements

2) reliability of onsite emerge ac power sources usually diesel generators); and ) site-specific design The capability to cool a reactor core must be continuously provisionsforoffsitepower,includmgvulnerabilityof the maintamed in order to ensure the removal of decay heat site to hurricanes, tornados, and ice storms. generated by fission products.The risk resulting from loss function in operating of this decay nuclear heat was power plants removal invest(DHR)igated in USI A-45, Application of the methods in Regulatory Guide 1.155 " Shutdown Decay Heat Removal Requirements." Th,e would determine the station blackout duration . ., 2, 4, staff has resolved this issue requinng plant-specific ,

8, or 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />) for which coping capability must hown analyses under the Individual ant Exammation (IPE) at that plant. However, applicants and licensees could program.

. propose alternative methods to that specified in the regu-latory guide to justify other durations for station blackout The technical findings for this resolution are summarized capability. Licensees may use an alternative ac power in NUREG/CR-5230 and include important insights source to cope with a station blackout if that source meets gained from decay heat removal failure-related risk as-specific criteria for independence and capacity and can be sessments for six operating plants.These studies included shown to be available within an hour. The rule calls for assessment of the reliability of decay heat removal sys-submittal of ant. specific station blackout coping evalu- tems, thermal-hydraulic analyses, emergency operatmg ationsin Ap I1989.The schedule forimplementation of procedures, system engincenng feasibihty studies, and any equipment and associated procedure modifications evaluation of the vulnerability of these systems to fire, I

necessary to meet the requirements of the rule will be es- flood, earthquake, and sabotage. A regulatory analysis tablished by the NRC staff in coordination 7 with the licen- that evaluates six alternative resolutions is reported in sees, but generally within about 2 years. NUREG-1289.

NUREG-1266 32

5 Safety issues and Regulations These studies, together with the operating history of " reflect the important strategies of the agency in protect-DHR failures, led to the conclusions that (1) the risk due ing public health and safety; aids in the planning of future to loss of the DilR function could be unduly high for research activities needed to fulfill agency goals; and some plants: (2) DIIR failure vulnerabilities, and the op- identifies the relative importance among ongoing re-timum corrective actions for those vulnerabilities, are scarch activities to the goals of the agency. Penodically, strongly plant specific; (3) a new dedicated D1IR system this report will be revised to include prioritization of new is neither cost beneficial nor necessary and therefore research activitics, deletion of completed activitics, and should not be required on a generic (all plants) basis; and changes in budget allocations ano projections.

(4) detailed plant-specific analyses under the IPE pro-gram, as part of the Commission's severe accident policy. 5.2 Standardized and Advanced will be the most effective means of determming DHR vul-nerabilities and the most appropriate corrective actions Reactors for each plant.

5.2.1 Statement of Problem 5.1.3.3 USIs A-3, A 4, and A-5-Steam Generator Tube Integn,ty 'Ihe Commission has issued a policy statement on the regulation of advanced m! clear power plants (51 FR 24643), which states that the NRC will review and As part of tlye steam generator USl program, the staff is-sued Genenc Letter 85-02 to all PWR heensees and ap- comment on new design concepts. It also encourages plicants to inform them of the staff recommended actions carly interaction with applictmts. As part of this program, and to request a description of their overall programs for the NRC will develop, review, and implement advanced reactor safety and policy issues in the ongoing NRC re-ensunng steam generator tube mtegnty and steam gen- view of the Department of Energy's advanced reactor crator tube ru ture (SG FR mitigation. 'lhe staff's assessment of t{ic licensee an)d apphcant responses concepts.

to In-depth independent analysis will be per-Generic Letter 85-02 was provided to the Commission in formed as necessary to verify that advanced reactor de-SECY-86 ,97 in March 1986. T he staff concluded on the s gns have the potential for enhanced margins of safety, basis of this assessment that the large majority of the and appropriate means will be used to accomplish their safety function.

licensees and apphcants were followmg programs, prac-tices, and/or procedures that were partially to fully con-sistent with, or equivalent to, the staff-recommended .lhe Commission has issued a policy statement on actions. ,

standardization (52 FR 34884).The purpose of the policy statement is to encourage standardization and to provide for certification of plant designs that are essentially com-Following the North Anna 1 SGTR event on plete in both scope and level of detail. The policy state-  ;

July 15,1987, NRC llulletin No. 88-02 was issued re-ment also reflects the applicable provisions of the Severe questing that licensees and operating license applicants Accident Policy Statement (501 R 32138) and the pro-perform specified inspections and analyses to determme posed Nuclear Power Plant Standardization and 1.icens-whether their plants are susceptible to the failure mecha- mg Act of 1987. As part of this policy, the NRC willimple-nism that led to the North Anna event and that they im- ment the standardization policy on future standard plant plement corrective actions if necessary. applications.

'Ihe Commission's current regulations (10 CFR Part 50, Appendices A and B; 10 CFR 50.55a; 10 CFR 50.109; 5.2.2 I,rogram Strategy  !

J and 10 CFR Part 100) provide the staff with sufficient 'the Department of Ener has submitted tbrec advanced authority to ensure that licensees are implementmg pro- l design concepts for NRL review. 'lhese are the Sodium grams relating to steam generator tube mtegnty that pro- Advanced Fast Reactor (SAFR), Power Reactor l vide adequate protection to public health and safety. the inherently Safe Module (PRISM), and Modular iligh- I staff will contmue to momtor steam generator expenence . The as an indicator of the effectiveness of heensee programs Temperature for ensuring steam generator tube integrity. As exernph- strategies for thisGas-Cooled program are: (1) Reactor conceptual (MiiTGR) design review of these three plants,(2)ider.tification of major fied by Bulletin 88-02, the staff may impose additional issues that need to be addressed prior to licensmg, requirements (pursuant to applicable regulations) to con- (3) identification of design features that should be ven-tinue to ensure that hcensees are implementmg fied, and (4) issuance of safety evaluation reports (SERs).

adequately effective programs where such action is deter-mined to be necessary on the basis of operating experi-ence or as a result of ongomg staff studies. Ihus, as stated With regard to standardization, as part of the implemen-m SECY-88-272, USIs A-3, A-4, and A-5 were resolved tation of the Commission's licensmg reform proposals, the legal staff is presently working on rulc changes to clar-and requirements were established, i ify agency policy on design certification by rulemaking l permitted under Appendix 0. RES will provide support in l S.1.3.4 Priorization of Research Activities connection with this rulemaking. I l

j NUREG-1319,"A Prioritization of Research Activities," Parallel with the policy development effort, NRR is j was issued in December 1988. This report provides a basis reviewing three standard designs (AHWR, the i for management decisions on allocating funding so as to Westinghouse SP-90, and the CE LESSAR) and design 33 NUREG-1266

5 Safety Issues and Regulations requirements prepared for standard plants by EPRI. 1988. A public workshop was held on this subject in De- l

. Generic issues pertaining to standardization requiring cember 1988 to solicit comments on technical issues and further attention are expected to emerge from these on the general approach to rulemaking. The need for se- .

reviews. vere accident rulemaking will be reexamined later in 1989 l after the Commission has approved design features for The strategies of this program are: (1) support of the each of the standardized plant designs.

legal staff's rulemaking activities on standardization and (2) resolution of genene issues pertaining to standardize- In March 1988, the Commission issued a Policy State-tion ansmg from the ongoing standard plant reviews. ment on the Maintenance of Nuclear Power Pfants. In this policy statement, the Commission indicated its inten-5.2.3 Research Accomplishments in 1988 tion to pursue a rulemaking on maintenance. In develop- )

ing this proposed rulemakmg on maintenance, the staff I 5.2.3.1 Advanced Reactur Concepts had extensive interactions with U.S. industry (airline and I nuclear) and foreign nuclear mamtenance programs and The staff continucd to review three advanced reactor con. practices. In addition, a 3-day public workshop was held in cepts that were submitted by the I)cpartment of Energy. July 1988 to solicit feedback on rulemakmg options. In-The purpose of these reviews is to determine the licen. formation gathered m these mteractions and from the sabihty of these unique designs /Ihe conceptual designs workshop was used in formulating the proposed rule and consist of two advanced liquid metal reactors and one ad, its supporting regulatory guide. Ihe proposed rule was vanced modular high-temperature gas-cooled reactor. published for comment m the Fedeml Repster in The draft SERs for these advanced reactors (MHTGR, November 1988. The Commission intends to issue the PRISM, and S AFR) have been reviewed and commented final maintenance rule and a draft regulatory guide, pro-upon by the Advisory Committee on Reactor Safeguards viding guidance for complying with the rule, m early 1989.

(ACRS).The Commission made the decision to issue the SER on the MHTGR to the Department of Energy for In 1988, the final report on Sandia's Fire Risk S:oping comments.The staff plans toissue SERs on the two hquid Study was completed. Based on the findings of the study, metal advanced reactors in early 1989. Commission pa- the staff intends to issue a comminion paper (in early pers (SECY-88-202 and SECb88-203) identified key 1989) to recommend that a fire research policy issues associated with the advanced reactor de- tinued in specific areas. In addition, three program potentialbe con-signs. In addition NUREG-1226, " Development and generic issues have been identified. These will l'e Utilization of the NRC Policy Statement on the Regula- addressed through the generic issac procesa, tion of Advanced Nuclear Power Plants," was issued in June 1988 to provide further guidance on the staff's ad.

vanced reactor review p ans. The final re ort on Sandia's Environmental Qualification '

of Electrica Equipment Risk Scoping Study (NUREG/

CR-5313) was completed in May 1988. This report con-5.2.3.2 Standardization tains conclusions and recommendations based on the n ngs o tk M s4 The NRC believes that standardization of nuclear power plant designs is an important initiative that can signifi-cantly enhance the safety, reliability, and availabihty of 5.3 Fuel Cycle, Transportation, and nuclear plants. The Commission intends to imptove the hcensmg process for standardized nuclear power phmts sareEuards and to reduce complexity and uncertainty in the regula-tory process. In this regard, the Commission issued a 5.3.1 Statement of Problem revised Standardization Policy Statement on Septem-ber 15,1987, which stated the Commission's intention to Effective regulation of fuel cycle, safeguards, transporta-develop a rule, codifying the process for approvmg stan- tion, and material safety activities involves the task of dard plant designs. plannag, developing, and issuing appropriate regulatory positions. Using mformation generated internally or Accordingly, in August 1988 the Commission issued for through narrowly directed research, new positions are public comment proposed regulations (10 CFR Part 52) developed or existing positions are modified. These posi-to implement the revised standardization policy.The pro- tions can take the form of regulatory requirements,polig posed Part 52 will provide a regulatory framework for cer- statements, guidance, or criteria for activities withm this tification of reference designs by means of rulemaking to element. Specific activities include: decontaminating and alleviate the need to reconsider design issues in individual decommissmning licensed nuclear facilities; transporting licensing proceedings on future license applications that radioactive materials; disposing of low-level radioactive reference the certihed designs- waste streams; and safeguarding facilities and special nu-clear materials. Setting of prionties for regulatory needs 5.2.3.3 Policies and Standards Development or deficiencies are undertaken to ensure that the prob-lems of greatest significance to the public health and A commission paper (SECY-88-248) recommending safety or the common defense and security are addressed rulemaking to implement the Severe Accident Policy for in an expeditious manner through properly defined Future 1.ight Water Reactors was issued in September regulatory and supporting research programs.

NUREG-1266 34

5 Safety Issues and Regulations 1

l t

- 5.3.2 Program Strategy 5.3.3 Research Accomplishments in 1988 5.33.1 Fuel Cycle The program strategy for each of the activities in this . . .

clerient are as follows: An,m, terim policy statement presenting existing NRC -

policies on residual radioactivity has been developed.

This interim policy would allow lands and structures to be released for unrestricted public use following license ter.

The NRC needs to develop a regulatory, appr,oach to minations. Also described are current NRC activities di-evaluate future requests myolvmg decommissiomngs and l rected at policy expansion and modification. '

' license terminations. This regulatory approach should define acceptable alternatives, requirements, and criteria for decommissionmg before stich a request is received. 533.2 Transportation The strategy has two parts: (1) to develop or modify regu- A proposed rule modif latory requirements and guidance to protect workers and tions has been issued. Theying NRC'speriod comment transportation for this pro-regula-the public from radiation risks associated with fuel cycle posed rule has been extended to allow the Department of and material operations involving the decommissioning of Transportation to issue a companion rule.The rule would

. licensed nuclear facilities and (2) to establish tafe radio- maximize compatibility between NRC and IAEA regula-logical criteria for residual radioactivity that avoids un- tions and proposes limitations on the shipment of low-necessary expenditures of funds to protect against trivial specific-activity materials.

risk.

5333 Material Safety In the area of transportation of nuclear material, the U.S. A C"**I88I " P" er proposing an exemption policy was Trade Agreements Act of 1979 directs Federal agencies completed."The [ommission modification of !this, pol cy to develop standards that are internationally consistent, w s published as an advanced notice of Commission m-whenever appropriate. A proposed rule has been Issued tent to develop such a policy, and public comments were to revise the transportation regulations for low-s ceific- solicited. An mternational workshop to di i was held mg is scheduled activity material and ensure that the properties o[the ra- fpr early m,in October 1988; a pubh 1989. A proposed rule to allow onsite incinera-dioactive in shipmentmaterials adequately being shipped protect and the public thet epackah'es and occu- used tion of slightly contammated waste oil generated at nu-pational health and safety. The final rule will achieve clear power plants was published for public comment, comments were received and evaluated, and a final rule is maximum compatibility between NRC regulations and the transportation regulations of the International under development.

Atomic Energy Agency (IAEA).

533.4 Safeguards During 1988 efforts were directed toward the develop-The purpose of the safeguards program is to issue ment of a fMal policy statement or rule regarding require-changes or additions to safeguards policy, regulations, or ments foi unescorted access to nuclear power plants. In guidance that meet the needs of the Offices of Nuclear addition, minor rules proposing to amend and reflect cur-Reactor Regulation and Nuclear Material Safety and rent administrative procedures were developed. A final Safeguards. The strategies are to (1) determine that rule improving safeguards requirements at fuel facilities physical security and accountability of strategic special possessmg weapons-grade nuclear material was issued.

nuclear material (SSNM) remain adequate; (2) ensure that the value of security, physical protection, and mate-

, 3A Developing and Improving rial control and accountability are balanced against imple-mentation costs; and (3) develop or modify safeguards Regulations regulatory requirements and guidance to be internally consistent. 5.4.1 Statement of Problem I

l RES has the primary responsibility to manage, coordinate j reviews, and control all NRC rulemaking activities and to In the area of material safety, the low-Level Radioactive monitor scheduling of such rulemakings to ensure that Waste policy Amendments Act of 1985 requires NRC to rules are developed in a timely manner. In addition, RES establish standards and procedures for expedited action provides support for preparation of regulatory impact on "below regulatory concern" (HRC) waste disposal pe- analyses (RIAs) that accompany all rulemaking through titions. Federal agencies, includmg NRC, are currently in development of generic methodology and guidance.

the process of establishing BRC or exemption levels for Techmcal reviews of all RIAs are performed upon re-radioactive waste disposal. The strategy of this program is quest. The NRC Regulatory Agenda Report and other to carry out Commission directives to develop a broad management information systems associated with rule-Commission exemption policy.This policy would serve as making activities are maintamed.

the framework for specific exemption decisions involving disposal of low-level waste streams, as well as other activa. Needed regulatory products, e.g., regulations and regula-l- ties concerning the re! case or use of radioactive material. tory guides, are developed. Rulemaking is proposed or l

35 NURl!G-1266

5 Safety Issues and Regulations initiated, as appropriate, and complex rulemakings that ity of shift operating crews to effectively respond to off-span the techmcal or organizational responsibilities of norm:3 situations and could also add operating several groups or that involve novel or complex questions expu.ence to plant management by opening for senior of regulatory policy are managed. Petitions for rulemak- operators a career path into plant management.

mg are mvestigated.

'Ihe NRC is proposing to amend its regulations to rees-5.4.3 Program Strategy onsite dis-tablish posal ofits regulatory low-level authority radioactive foratapprovinkC-licensed waste N power reactor sites located in Agreement States. Also, for i

,Ihe purpose of the NRC nuclear regulatory program is to facilities licenses for special nuclear material activities, l ensure that nuclear facilities are designed, constructed, NRC believes it is prudent to clarify and to establish in the and operated m a safe manner. Therefore, there exists a  !

regulations that the onsite disposal of small quantities of contmumg need to revise rules and guides and to develop special nuclear material waste remains an NRC licensing new ones. The strategies of this program are: (1) review function in order to retain control over the decommis-the effectiveness of 1.WR regulatory requirements and sioning process. The NRC believes that these amend-guidance and make recommendations for revisions; ments are necessary in order to avoid unnecessary dupli-(2) develop screening methodology to systematically cation of effort on the part of both the Agrectnent States review requirements and guidance: (3) coordinate and and the Federal Government. Sole NRC jurisdiction review proposed changes to the IAEA safety standards; would allow for uniform standards of approval and (4) develop or assist the development of rules and regula-recordkeeping for onsite disposal, which would provide tory guides; and (5)contmue to develop and, mamtain greater assurance that the radioactive material is disposed management information systems for rulemakmg. ofin a manner that would not present a health hazard at a later date after the site is decommissioned.

5.4.3 Research Accomplishments in 1988 A rulemaking is being developed to amend the 10 CFR Part 35 regulations that apply to the medical use 5.4.3.1 Develop or Modify Regulat. ions of byproduct material. The amendments would require medical use licensees to implement quality assurance In a program initiated m. 1985 and ccmtinued through (QA,) programs and would revise misadministration re-1988, the NRC staff undertook to evaluate existmg regu- portmg requirements. Implementation of the new re-latory requirements m terms of their nsk effectiveness quirements would be supported by issuing a regulatory and to chminate or modify requirements without compro- guide that would include specific cnteria for medical QA mismg safety. A three-volume research report (NURI:Gi programs. The feasibility of this will be evaluated during a CR-4330) provided detailed technical assessments of re- pilot study involving several medical use licensees.

qutrements associated with a number of topics. Based on these and continuing studies, the NRC staff will recom-mend whether to eliminate or modify related require- The NRC, m. August 1988, published an active regulatory ments without compromising safety. 8.uide to provide guidance on screenmg areas to identify a site or sites for near-surface disposal of low-level radioac-

. .. tive waste. In order to expedite the site screening and se-It is anticipated that commercial nuclear power reactors lection process, the regulatory guide sugpsts that the li-will have a major need in the near future for additional censee conduct a geographic mformation system analysis spent foci storage space to supplement existing reactor of relevant geophysical and land-use data. An overview of spent fuel storage pools. After manyyears of commercial the methodology for conducting the site screening analy-power operation, these spent fuel storage pools are near- sis is provided m the regulatoiy guide.

mg full capacity. In response to this need, the Nuclear Waste Policy Act of 1982 directed the Secretary of Energy to establish a dry spent fuel storage demonstration pro- 5.4.3.2 Independent Review and Control of gram with the objective of establishing one or more tech- Rulemaking nologies that the NRC may approve for use at civilian nu-l clear power reactor sites without, to the maximum RES has the lead responsibility in the NRC for rulemak-i practicable extent, the need for additional site-specific ing. The control of rulemaking responsibilities includes l approvals. A proposed rule is being developed that would coordinating rulemaking activities with the requesting of-amend 10 Cl R Part 72 to allow dry storage of spent fuel fices and making recommendations for assignment of re-in NRC-approved casks. Holders of nuclear power reac- quested rulemakings within RES.

tor operatmg licenses would be allowed to store spent fuel in NRC-approved casks at reactor sites under a gen- During 1988, there were 40 rulemakings in various stages cral license. of completion, e.g., ongoing, completed, or terminated.

Of these,15 new rulemakings were initiated during the

'Ihe Commission is considering a proposed amendment year. In addition to the ongoing rulemaking efforts, two to its regulations regarding enhanced professional or advance notices of proposed rulemaking were published; educational credentials for senior nuclear power plant ten proposed rules were published for public comment; operating personnel.The proposed requirements are m- ten rules were published as final actions; and three tended to contribute to the goal ofimproving the capabil- rulemaking actions were terminated.

NUREG-1266 36

APPENDIX 1988 REGULATORY PRODUCTS FROM Tile OFFICE OF NUCLEAR REGUIATORY RESEARCil Date Regulatory Product Description Integrity of Reactor Components e January 1988 Standard Review Plan Plant design for protection against postulated piping (SRP 3.6.1) failures in fluid systems outside containment March 1988 Research Information Results of Meers Fault investigations Letter (RIL) 151 May 1988 Regulatory Guide Radiation embrittlement of reactor vessel materials 1.99, Rev. 2 May 1988 Final rule Update of 10 CFR Part 50.55(a), Codes and Standards, to incorporate 1986 editions of ASME Code Sections 111 and XI May 1988 Regulatory Guide Design and fabrication Code Case acceptability 1.84, Rev. 25 May 1988 Regulatory Guide Materials Code Case acceptability 1.85, Rev. 25 May 1988 Regulatory Guide Inservice inspection Code Case acceptability 1.147, Rev 6 May 1988 Regulatory Guide Criticality accident alarm system 8.12, Rev. 2, for comment June 1988 Regulatory Guide Scismic qualification of safety-related equipment for nuclear (

1,100, Rev, 2 power plants  ;

June 1988 Final rule General requirements for decommissioning nuclear facilities July 1988 Final rule Requirements for storage of spent fuel and high-level waste in monitored retrievable storage facility November 1988 Final rule Permits use of alternative method for containment leak rate testing Preventing Damage to Reactor Cores i September 1988 Final rule ECCS Rule (10 CFR 50.46 and Appendix K) on emergency core l cooling systems; revisions to acceptance critena l

1- October 1988 RIL 158 Final resolution of TMI Action Item II.C.4 on reliability NUREG/CR-4618 engineering November 1988 Generic Letter Individual plant examinations for severe accident vulnerabilities 88-20 November 1988 Staff Evaluation IDCOR methodology for individual plant examinations Reactor Containment Performance and Public Protection from Radiation January 1988 Regulatory Guide Instruction concerning prenatal radiation exposure. Revised as 8.13, Revision 2 a result of Presidential guidance and recommendations from the International Commission on Radiation Protection (ICRP) and the National Council on Radiation Protection and Measurements (NCRP).

37 NUREG-1266

Appendix A - l l

1 l- Date Regulatory Product Description Reactor Containment Performance and Public Protection from Radiation (Continued)

July 1988 ' Final rule The NRC regulation (10 CFR Part 35) was revised in response to a petition for rulemaking."Ihe petitioners requested that the Commission remove the requirement that radioactive aerosols be administered in rooms having a negative pressure relative to surrounding rooms. 'lhe Commission agreed that negative room pressures could have an adverse impact on the delivery of health care to patients with pulmonary diseases and that the requirement is not necessary to protect workers and public health and safety.

July 1988 Regulatory Guide Criteria for establishing a tritium bioassay program. This guide 832 provides criteria for developing and implementing a bioassay program for any licensee handling or processing tritium. The NRC determined that, because of the lack of such guidance, many licensees appeared to be unnecessarily concerned with tritium at the expense of other radiological hazards.

August 1988 Regulatory Guide Bioassay at uranium mills. This revision reflects the results of 8.22, Revision 1 more recent techniques identified and the results of the studies documented in NUREG-0874, " Internal Dosimetry Model for Applications to Bioassay at U;anium Mills."

September 1988 Final rule Emergency planning and preparedness requirements were estabhshed for fuel loading and low-power testing on nuclear power plants.

Confirming Safety of Nuclear Waste Disposal April 1988 RIL 152 Results of research on dating groundwater for high-level-waste repository site characterization.

April 1988 RIL 153 Methodology to assess safety during preclosure period of high-level. waste repository.

May 1988 Proposed rule Definition of high level waste.

August 1988 Final rule Licensing requirements for independent storage of spent fuel and high-level radioactive waste.

Resolving Safety Issues and Developing Regulations January 1988 Generic Safety Issues resolved in 1988, see page 32.

December 1988 January 1988 Final rule The NRC regulation (10 CFR Part 73) was revised in response to a petition for rulemaking. The petitioners requested several changes m the qualifications for armed security personnel. The final rule incorporated part of the petition that deleted a scheduling link between the time a medical examination and the physical fitness test  ;

l be given to armed security personnel.

August 1988 NUREG-1317 Regulatory options for nuclear plant license renewal.

August 1988 Guidance for selecting sites for near-surface disposal of low-level 4.1p)Re ulatory Guide radioactive waste. This guide provides guidance for conducting a site {

a NUREG-1266 38 ]

Appendix A Date - Regulatory Product - Description R solving Safety issues and Developing Regulations (Continued) screening investigation to identify a site or sites that have a high potential for meeting site suitability requirements mandated by the lew-Level Radioactive Waste Pohey Act (LLRWPA) and the IlllWPA as amended in 1985.

November 1988 Regulatory Guide Selection design qualification testing and reliability of diesel 1.9, Proposed generator elements used as onsite electric power systems at Revision 3 nuclear power plants.

November 1988 Final rule ne NRCsafeguards improved regulations (10 CFRbase requirements Part 73)d on the findings of awer research review team,which compared DOE and NRC safeguards programs. Improved safeguard areas included security system performance evaluations; night firing qualifications for guards; 100 percent entrance searches, armed guard at material access area control points; two protected area fences; and revision of the design basis threat.

39 NUREG-1266

NRC FOAM 335 U.S. NUCLE AR REGUL ATOR Y COMMIS$10H 1. REPORT NUMBE R E 'c b m. %KOlo"n"utm^$,7fh4"I#"'

nm. no BIBLIOGRAPHIC DATA SHEET (See instructions on the reverse) 2, TRLE AND SUBTMLt NtJREG-1266, Vol . 3 NRC SAFETY RESEARCH IN SlJPPORT OF REGllLATION - 1988 ,[' "'" "* *'[', "/,

May 1988

4. FIN OR GR ANT NUMBE R

b. AUlHOR(S) 6. TYPE OF REPORT Regulatory  !

7. PE R100 COVi k E D lanclume Datest 1988

8. RF RMi ANIZ AT LON - N AM E AND ADDR ESS HI NRC. provre Dwunon, OHwe or Regnon, U.S Nuctuar Regulatory Commossmn. and madme addren. It contractor, provoor Office of Nuclear Regulatory Research lj. S. Nuclear Regulatory Commission Washington, DC 20555

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10. SUPPLEMENTARY NOTES 11, ABSTRACT (200 words or seus This report, the fourth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used.

It summarizes the accomplishments of the Office of Nuclear Regulatory Research during'1988.

The goal of this office is to ensure that safety-related research provides the technical bases for rulemaking and for related decisions in support of NRC licensing and inspection activities. This research is necessary to make certain that the regulations that are imposed on licensees provide an adequate margin of safety so as to protect the health and safety of the public. This report describes both the direct contributions to scientific and technical knowledge with regard to nuclear safety and their regulatory applications.

12. K E Y WORDS/DESCR:P TORS tirar words or phrases ther win euest researchers m iocerme rhe roport.J 13 AV AILAbitsi v s1 A r tMENT lJnlimited i Nuclear regulatory research '"*"'""^""^"~ ,

(This Vegel Safety Research lJnclassified lThis Neporti lJnclassified 1b. NUMBER OF PAuE S 16 PRICE NRC FORM 3J6 Q-89)

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NUCLEAR REGULATORY COMMISSION 'oS'^ j30!'8 <

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PENALTY FOR PRIVATE USE. 4300 g(51

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