ML20086K577

NRC Safety Research in Support of Regulation - Fy 1994

ML20086K577
Person / Time
Issue date:	06/30/1995
From:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-1266, NUREG-1266-V09, NUREG-1266-V9, NUDOCS 9507200206
Download: ML20086K577 (100)

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NUREG-1266 Vol. 9 NRC Safety Researca in Sungort of Regu ation - FY 199L r

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AVAILABILITY NOTICE i

Availability of Reference Matenals Cited in NRC Pubbcations Most documents cited in NRC pubhcations will be available from one of the following sources:

1.

The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2.

The Superintendent of Documents. U.S. Government Printing Office. P. O. Box 37082, Washington, DC 20402-9328 3.

The National Technical information Service, Springfield, VA 22161-0002 Although the listing 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 frcm the NRC Public f

Document Room include NRC correspondence und internal NRC memoranda; NRC bulletins, I

circulars, information notices, inspection and investigation notices; hcensee event reports; vendor reoorts and correspondence: Commission papers; and applicant and licensee docu-ments and correspondence.

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The following documents in the NUREG series are avadable f or purchaso from the Government l

Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro-ceedings, international agreement reports, grantee reports. and NRC booklets and bro-chures. Also available are regulatory guides, NRC regulations in the Code of Federal Regula-tions, and Nuclear Regulatory Commission Issuances.

4 Documents available from the National Technical Information Service include NUREG-series

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reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commisston.

Documents available from pubhc and special technical huraries include all open literature items, such as books, journal articles, and transactons. Federal Register notices, Federal and State legislation, and congressional reports can usually be obtained from these hbranes.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC con.

ference proceedings are avadable for purchase from the organization sponsonog the pubhca-tion cited.

Single copies of NRC draft reports are avautre 'ree. to the extent of supply, upon written request to the Office of Administration. D ' obuton and Mad Services Section. U.S. Nuclear l

Regulatory Commission. Washin;; ton DC ~ U 5-0001.

Copies of industry codes and standards used in a substantwe manner in the NRC regulatory process are maintained at the NRC Library, Two Wn:te Fhnt North.11545 Rockvdle Pike. Rack-j vi!!e, MD 20852-2738. for use by the pubhC. Codes av standards are usually copynghted and may be purchased from the onginating organizaton or, if they are Amer Can National Standards, from the American National Standarcs institute.1430 Broadway. New York. NY i

10: 18-3308.

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NUREG-1266 Vol. 9 l

NRC Safety Research in I

Support of Regulation - FY 1994 1

l Manuscript Completed: April 1995 Date Published: June 1995 i

Olrice of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Wasliington, DC 20555-0001

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ABSTRACT a This report, the tenth in a series of annual related decisions in support of NRC regulatory /

reports, was prepared in response to licensing / inspection activities. RES also has congressional inquiries concerning how nuclear responsibilities related to the resolution of generic regulatory research is used. It summarizes the safety issues and to the review of licensee accomplishments of the Office of Nuclear submittals regarding individual plant examina-Regulatory Research durmg FY 1994, tions. It is the responsibility of RES to conduct The goal of the Office of Nuclear Regulatory the NRC's rulemaking process, including the Research (RES)is to ensure the availability of issuance of regulatory guides and rules that sound technical bases for timely rulemaking and govern NRC licensed activities, iii NUREG-1266

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Contents Page Abstract.....................................................................

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H igh ligh t s...............................................................................

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Part 1-Nuclear Safety Research-Reactor Licensing Support 1.

Reactor Aging an d Re n ewal............................................................ 1-1 1.1 Reactor Vessel Safety and Piping Integrity.................................

1-1 1.2 Aging of Reactor Components..........

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Stan dard Reactor Designs............................................................. 2-1 l

2.1 Systems Performance of Advanced Reactors..................................... 2-1 2.2 Engineering Issues for Advanced Reactor Designs...........

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2.3 Regulatory Application of New Source Terms..................................... 2-4 Part 2-Nuclear Safety Research-Reactor Regulation Support 3.

Plant Performance.......................

3-1 3.1 S ta t em e nt of Proble m............................................................. 3-1 l

3.2 Program Strategy.........................

3-1 3.3 Research Accomplishments in FY 1994.............................................. 3-1 i

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H u m an Reli abili ty................................................................ 4-1

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4.1 S t at e me n t of Proble m............................................................. 4-1 4.2 Program Strategy.........................................

4-1 4.3 Research Accomplishments in FY 1994............................................. 4-1 5.

Reactor Accident Analysis............................................................. 5-1 5.1 Reactor Risk Analysis...............

5-1 5.2 Containment Performance.......................................................... 5-3

'5.3 Severe Accident Phenomenology.................................................. 5-7 5.4 Reactor Containment Structural Integrity......................................... 5-11 5.5 Severe Accident Policy Implementation........................................... 5-13 6.

Safety Issue Resolution and Regulation Improvements................................. 6-1 6.1 Earth Sciences..............

6-1 6.2 Plant Response to Seismic and Other External Events............................... 6-4 6.3 Generic Safety Issue Resolution................................................. 6-6 6.4 Reactor Regulatory Standards..................................................

6-8 6.5 Radiation Protection and Health Effects...................................... 6-10 l

l 6.6 Small Business Innovation Research.............................................. 6-13 l

l Part 3-Safeguards Regulation Program l

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Nuclear Materials Research...............................................

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7.1 St etemen t of Problem........................................................... 7-1 7.2 Program S trategy................................................................ 7-1 7.3 Research Accomplishments in FY 1994........................................... 7-2 v

NUREG-1266

Contents (continued)

Page 8.

Low. Level Waste Disposal.................

8-1 8.1 Stat ement of Problem......................................................... 8-1 8.2 Pr ogra m S t ra t egy............................................................... 8-1 83 Research Accomplishments in FY 1994........................................

8-1 Part 4-Assessing the Safety of IUgh-Level Waste Disposal 9

H igh. Level Waste Resea rch...................................................... 9-1 9.1 Statement of Problem 9-1 9.2 Program Strategy..............

9-1 93 Research Accomplishments in FY 1994.....

9-2 Appendix-FY 1994 Regulatog Products from the Office of Nr. clear Regulatory Research......... A-1 Tables 6.1 Generic Safety Issues Prioritized in FY 1994..

6-7 6.2 Generic Safety issues Resolved in FY 1994.

6-7 6-7 63 Generic Safety Issue., Scheduled for Resolution...........

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IllGIILIGIITS Part 1--NUCLEAR SAFETY in performing audit inspections of piping at RESEARCII--REACTOR an operating nuclear power plant.

LICENSING SUPPORT A collaborative research agreement was e

reached between the Japan Atomic Energy Reactor Aging and License Renewal Research Institute (JAERI) and the Oak Ridge National Laboratory (ORNL) to Pressure Vessel Safety and Piping Integrity exchange information gained from ORNL a min tion of material from the Based on public comments, modifications o

were made to the draft regulatory guide for decommissioned Japan Power Demonstration evaluating pressure vessels whose beltline re ctor. h informaton exchange wd1 result 1

materials have Charpy upper-shelf energy in a better understandmg of changes in (the indication of reactor vessel toughness) reactor pressure vessel matenal propert,es i

that has fallen below the 50 ft-lb regulatory caused by long-term irradiation.

limit prescribed in Appendix G to 10 CFR An improved method for performing e

Part 50. Generic analyses using the draft inservice inspections of steam generator regulatory guide methodology demonstrated tubing was successfully demonstrated at two that adequate pressure vessel integrity exists operat!ng nuclear power plants. The method, for pressure vessels with Charpy upper-shelf which emph>ys multiple " pancake" type coils energy values well below the 50 ft-lb value and multifrequency data analysis for better specified in Appendix G. It is expected that sensitivity, produced significantly higher the final guide will be published in FY 1995.

signal levels with a greatly enhanced inspec-i tion speed over other currently produced Based on public comments, modifications improved techniques.

o were made to the draft regulatory guide for calculational and dosimetry methods for A draft regulation and draft regulatory guide e

determining pressure vessel neutron fluence addressing the engineering and metallurgical for power reactors. In support of this draft aspects of thermal annealing for U.S. plants guide, new neutron cross-section libraries were developed and published for public that apply the latest evaluated nuclear data comment. The proposed regulation provides files were published.

the administrative and technical basis for performing thermal annealing of reactor Interim fatigue design curves were revised to pressure vessels in U.S. commercial nuclear o

more accurately describe fatigue life of power plants. The draft regulatery guide primary pressure boundary components elaborates the information licensees should exposed to the high-temperature coolant in develop and provide as part of the thermal light-water reactor (LWR) systems. Fatigue annealing plan development, including tests are in progress to validate and/or detailed information concerning the pressure update the proposed design curves.

vessel and other components that could be affected by the high-temperature annealing, A Synthetic Aperture Focusing Technique for measurements that are to be made before, o

Ultrasonic Testing (SAFT-UT) system, which during, and after the annealing, and the provides a method for more reliable detection method for determining the post-anneal and sizing of flaws, was fabricated for the material properties for the pressure vessel NRC's nondestructive examination mobile beltline materials.

i laboratory. Specialized training was provided e

Amendments to 10 CFR 50.61 and to for the mobile laboratory personnel. The 4

i system was successfully used by the mobile Appendices G and H of 10 CFR Part 50 were laboratory personnel for the first time in 1994 proposed and published for public comment.

vii NUREG-1266

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Highlights The proposed amendments would clarify the if the MOVs comply with GL 89-10. Friction i

pressurized thermal shock requirements, the experiments were conducted on samples of fracture toughness requirements, and the corroded materials typical of certain valves.

reactor vessel materials surveillance program Additional experiments are continuing to requirements.

determine the effects of corrosion on friction and thrust.

Electrical and Mechanical Components Standard Reactor Designs Research continued into the aging-related e

degradation of performance of safety-Systems Performance of Advanced Reactors significant components and systems. The hfodification of the ROSA facility to validate e

research also addresses methods for mitigat-the applicability of NRC codes to the AP-600 ing and managing the aging process in these design and provide independent confirmation components and systems. Draft reports were of the predicted performance of AP-600 issued addressing the chemical and volume plant safety systems was completed. Twelve of control systems for pressurized water the 14 Phase 1 tests were successfully carried reactors, containment cooling systems, out. Results thus far suggest that this reactor reactor core isolation cooling systems, can be successfully cooled under a variety of accumulators, air-operated valves, and postulated accident scenarios. Two potential isolation condenser systems.

safety issues were identified during testing (i.e., a large thermal gradient occurred in Results from work completed on the risk-piping where cold water from the passive e

based methodology for assessing aging effects heat removal systems enters and the possi-in nuclear power plants have been used for bility of a water hammer resulting from identifying safety-related motor-operated contact between subcooled water and steam) valves (MOVs) having the most impact on and are Nng carefully evaluated for their plant risk. The valves being considered are safety implications.

those covered by Generic Letter 89-10 Construction of an integral test facility is

" Safety-Related Motor-Operated Valve Testing and Surveillance." Dynamic tests and nearing completion at Purdue University to surveillance tests, in accordance with GL carry out tests on a broad spectrum of 89-10, could then be performed on those loss-of-coolant accidents and transients MOVs with the largest risk impact. Relative postulated for the General Electric simplified risk importance of single MOVs and the boiling water reactor. Testing is scheduled to interactions of MOVs with other components, begin in the summer of 1995 and continue including other MOVs, can be analyzed using into 1996.

this approach. A draft report documenting The program to demonstrate a method for e

the results of this work was issued for review estimating the reliability of passive safety in FY 1994. The results have provided the systems m advanced reactor designs such as technical basis for evaluating licensees, the Westinghouse AP-600 and the General submittals for ranking their respective MOVs for tests in accordance with GL 89-10.

Electric SBWR was contmued. A peer review meeting to discuss the approach and its status was held. The demonstration project Research continued into the factors affecting e

will be completed in late 1995.

the performance of motor-operated valves (MOVs), specifically addressing whether Engineering Issues for Advanced Reactor corrosion m internal valve parts can sig-Designs mficantly affect the toique and thrust The Advanced Light-Water Reactor (ALWR) required for MOVs to perform their func-tions when called upon to do so. Additionally, Equipment Qualification Panel has resolved this information will be used in determining all issues pertaining to the experience-based NUREG-1266 viii

i Highlights approach for seismic qualification of Group 1 burnups and to revise, as necessary, the items. Group 1 equipment has a mature licensing fuel damage criteria.

design with little design variability and, in Developmental assessment, peer review, and general, has demonstrated characteristics of inherent seismic ruggedness.The ALWR documentation of the revised RELAP5 code Equipment Qualification Panel met with have been completed with NRC participation vendors of batteries and transformers to in the International Code Assessment and discuss design variability and failure modes Maintenance Program continuing. The for ALWR designs. In addition, the panel RELAP5 code is being used for the certifi-toured a select group of the candidate cation of the AP-600 and SBWR as well as to facilities that had experienced carthquake support licensing activities related to damage and would be used in the EPRI/ ARC operating plants, e.g., pressurized thermal data base.

shock.

Regulatory Applications of New Source Terms Iluman Reliability A method has been developed for assessing A revised accident source term document e

e (NUREG-1465) reflecting severe accident the effectiveness of training programs at research insights gained over the last 30 years nuclcar power plants and efforts are con-was issued. The new source term provides tmum, g to establish a techmcal base for more realistic estimates of fission product settmg mmimum staffing levels for both releases into containment in terms of timing, control rooms and operatmg support staff.

nuclide types, quantities, and chemical form than the current source term which dates A handbook detailing the effects on human o

from 1962.

performance of environmental factors such as light, heat, etc., has been published for use by Public comments on a proposed revision to NRC inspectors.

e NRC's siting criteria (10 CFR Part 100) are being analyzed, and a final rule is being The proceedings of the NRC/NIST workshop o

on digital system reliability have been issued developed reflecting comments received. This i

and the first phase of a study by the National rule will incorporate bas,c reactor sitmg Academy of Sciences to define an effective criteria and the continued use of accident approach to the regulation of computer-based source terms and dose calculations for siting (digital) technology in nuclear safety and

. of certain custom plants.

control systems was initiated.

W rk continued on the project cooperatively e

Part 2-NUCLEAR SAFETY RESEARCH--REACTOR sponsored with EPRI to develop guidehnes REGULATION SUPPORT for the verification, validation, and quality assurance for certification of high-integrity software for use in plant safety systems.

Plant Performance Similar studies have been initiated to develop Changes in fuel pellets and cladding that ases fo@ enWronmental quaMeadon o

of advanced instrumentation and control appear to reduce the fuel's resistance to damage occur at high burnups. These systems.

changes affect fuel behavior computer codes Reliability and risk analysis tools have been that are used in licensing safety analyses and developed to permit ready evaluation of the affect a number of fuel damage criteria that risk impact of changes to plant technical appear in regulations, regulatory guides, and specifications.

the standard review plan. Research programs are now in place to update the NRC's fuel

" Advanced Human-System Interface Design behavior codes for application at high Review Guidelines" (NUREG/CR-5908), in ix NUREG-1266

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Highlights support of the standard review plan, was the Zion reactor, determined that the issued setting forth guidance for staff pressure loads produced by DCH are consideration of proposals for control room significantly kiwer than earlier estimates and

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designs or modifications.

present a negligible threat to reactor con-tainment integrity. Peer-reviewed reports Reactor Accident Analysis (NUREG/CR-6075 and Supplement 1 to NUREG/CR-6075) documenting these Reactor Risk Analysis findings have been issued.

In cooperation with the Japanese Ministry of Results of an analysis of the risks of potential e

o occurrences during low-power and shutdown International Trade and Indust y, con-t operations have been completed. These struction has been completed on the High-analyses suggest that traditional technical Temperature liydrogen Combustion facility, specifications may not always be adequate to This facility will study the modes of high temperature H combustion (deflagration /

accommodate potential accidents occurring 2

during such operations.

detonation) to assess the possible threat of hydrogen detonations on containment An analysis of the South Texas nuclear integrity. Lower temperature testing to o

characterize H combustion under conditions t

project to support its request for modifi-2 cations of plant technical specifications, simulating accident environments (steam)

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based in part on risk, was completed. The suggest that mixtures initially nonflammable analysis determined that the applicant's (because of steam concentrations) became assessment of core damage frequency was flammable as steam was condensed and that within the range of estimates for other similar glow plugs were successful in igniting the gas 3

facilities, and this information has been used with no detonation or accelerated flame by NRR to support their regulatory decision.

propagation observed.

Under an agreement between the Commis-The cooperative program with the European e

o Communities and the Organization for sion of European Communities Joint Economic Cooperation and Development to Research Center (JRC) and the NRC in the carry out an intercomparison exercise on six field of severe accident research related to accident consequence codes, including the molten fuel-coolant interactions (FCIs), four i

NRC-developed MACCS code, was success-successful tests have been performed to date fully completed. The study indicated substan.

in the large-scale (150-kg reactor prototypic tial agreement in the reactor accident melts) FARO test facility. These tests consequence predictions made by these investigated non-explosive melt breakup and codes.

quenching. structure heating, and the limits of melt-coolant mixing at high pressure.

o Probabilistic risk assessment (PRA) data Additional tests have been performed in the i

from four more licensed power plants were small-scale KROTOS facility (one-added to the SAPillRE (Systems Analysis dimensional shock tube geometry) to investi-l Programs for Hands-on Integrated Reliability gate steam explosion phenomenology at low Evaluation) data base to bring the total to 17.

pressure.

SAPHIRE is a set of codes used in The OECD RASPLAV project is investigat-performing PRAs and allows the staff to e

create, quantify, and evaluate accident risks.

ing melt-vessel interactions and providing data on internal natural convection flow and Containment Performance the local heat flux distribution inside the

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lower head of the reactor pressure vessel for The culmination of extensive experimental various melt compositions. This project o

and analytical research on the issue of direct involves large-scale integral experiments using containment heating (DCH), principally for UO in representative lower head reactor 2

NUREG-1266 x

Highlights pressure vessel geometries (i.e., slice Safety Issue Resolution and Regulation geometry), analytical studies, and a number Improvements of small-scale separate effects experiments.

Recently, a small-scale experiment that Earth Sciences showed the feasibility of using slice geometry for large-scale experiments was carried out.

Archaeological and geological evidence identified during investigations in the New l

Madrid seismic zone, along with geological Reactor Containment Structural Integrity evidence from other studies, demonstrate that prehistoric earthquakes (at least two)large In order to improve the state of practice in enough to cause liquefaction (magnitude 5.5 o

inspection of containments to reduce the to 6.0 or larger) have occurred in the New i

chances of having significant undetected Madrid seismic zone in the late Holocene.

degradation due to corrosion, work continued These findings contribute toward providing a in 1994 on the rulemaking to incorporate by basis for estimating the seismic hazard in the reference Subsections IWE and IWL of south central United States.

Section XI, Division 1, of the ASME Boiler The panel of experts assembled for thej.. t e

and Pressure Vessel Code into 10 CFR om 50.55a. Subsection IWE provides rules for the NRC/ DOE /EPRI study of methods for inservice inspection of metal containments pr babilistic seismic hazard analys,s (PSHA) i and the liners of concrete containments.

has cornpleted the development of new Subsection IWL provides rules for the guideh,nes for performmg PSHA. The panel inservice inspection of the reinforced has issued a draft report describmg the concrete and the post-tensioning systems of guidelines, which emphasize methods of concrete containments. The proposed rule clicitmg expert opmions. A peer review by the does not address inspection ofinaccessible National Academy of Sciences / National areas; the state of practice for inspection of Research Council is in progress.

inaccessible areas will have to be improved before a resolution of this issue is achieved.

Plant Response to Seismic and Other External Events The Commission approved the staff's rec-e Severe Acc. dent Policy Implementat. ion i

ommendation to issue a second proposed revision of Appendix A," Seismic and o

Seventy-six IPE (internal-event) submittals Geologic Siting Criteria for Nuclear Power have been received to date, with one due in Plants," to 10 CFR Part 100, " Reactor Site CY 1995. Staff evaluation reports (SERs)

Criteria," for public comment. This revision have been issued for 27 submittals. It is reflects new information and research results expected that all IPE submittals will be available since the first proposed revision of reviewed and SERs issued by the end of CY the regulations was issued and comments 1996.

received from the public on that proposed revision of the regulations. Draft regulatory Sixteen complete and four partial IPEEE guides and standard review plan sections o

(external-event) submittals have been received providing methods acceptable to the NRC of which four are being evaluated.

staff for implementing the proposed regula-tions were issued for pubhc comment in Studies began of the IPE results to gain more o

generic insights. Issues such as the plant to-On January 17,1994, a magnitude 6.7 earth-e plant variability in estimated core damage quake occurred in the San Fernando Valley frequency results and the reasons for this near the town of Northridge, California. This variability are being studied.

is the same general area affected by the xi NUREG-1266

I Highlights A proposed rule,10 CFR Part 20, on fre-magnitude 6.5 San Fernando earthquake of e

1971. Representatives from the NRC Offices quency of medical examinations for use of l

of Nuclear Regulatory Research (RES) and respiratory protection equipment was issued Nuclear Reactor Regulation, and an RES in September 1994. The proposed amendment contractor, Lawrence Livermore National would remove the requirement for an annual Laboratory, toured the damaged area. In medical examination and allow for alternative general, well-engineered structures and equip-timeframes.

t ment that may have experienced ground A proposed rule,10 CFR Part 21, on pro-motion far in excess of their design remained e

functional. However, extensive damage to curement of commercial grade items by l

several modern mid-rise and low-rise steel.

reactor liccuees was issued in October 1994.

moment frame buildings was observed.

The proposed rule responds to a petition for Components made of brittle materials, such rulemaking (PRM-21-02) submitted by the as ceramic insulators and cast iron com.

Nuclear Management and Resources Council ponents, received significant damage (NUMARC), which is now incorporated into consistent with that observed after other the Nuclear Energy Institute (NEI). The pro-earthquakes.

posed amendment would clarify and add flexibility for procuring items for t

s ety4clated sent Generic Safety Issue Resolution A proposed rule,10 CFR Parts 50,55, and e

During FY 1994, the NRC identified no new 73, on reduction of reporting requirements e

generic issues, established pnonties for three imposed on NRC licensees was issued in I

issues (see Table 6.1), and resolved five issues November 1994. The proposed amendments (see lhble 6.2). Table 6.3 contams the would reduce reporting requirements on i

schedules for resolution of all unresolved power reactors, research and test reactors,

issues, and nuclear material licensees.

Reactor Regulatory Standards Part 3--NUCLEAR MATERIALS LICENSING AND j

A final rule,10 CFR Part 55, on requalifi-e cation requirements for licensed operators for REGULATION SUPPORT j

renewal of licenses was issued in February 1994. The rule deletes the requirement that Nuclear Materials each licensed operator pass a comprehensive A proposed and a final rule,10 CFR Part 40, e

requalification written examination and an Appendix A, on uranium mill tailings were operatmg test during the 6-year license term issued in November 1993 and June 1994, as a prerequisite for license renewal.

respectively. The final rule conforms NRC regulations to Environmental Protection An advance notice of proposed rulemaking, Agency (EPA) regulations under the Clean i

e 10 CFR Part 52, concerning standard design Air Act and supports recision of certain EPA certification for evolutionary LWRs was Clean Air Act requirements.

issued in November 1993. The contemplated rulemaking would define the form and e

A final rule,10 CFR Parts 30,40,50,70, and content of the rules that would certify the 72, to allow self-guarantee as an additional design.

mechanism for financial assurance for decommissioning was issued in December A proposed rule,10 CFR Parts 19 and 20, 1993. This rulemaking is in response to a e

regarding the use of " controlled areas," the petition for rulemaking (PRM-30-59) definition of occupational and public submitted by the General Electric Company J

exposure, and training requirements was and Westinghouse Electric Corporation. The issued in February 1994.

final rule applies to certain financially strong, NUREG-1266 xii

l Highlights non-electric utility licensees and allows the formula quantities of strategic special nuclear use of self-guarantee as financial assurance material.

for decommissioning funding. It does not apply to electric utility licensees.

Uranium Enrichment An advance notice of proposed rulemaking A proposed and a final rule,10 CFR Parts o

e (ANPR),10 CFR Part 20, on disposal of 19, 20, 21, 26, 51, 70, 71, 73, 74, 76, and 95, for radioactive material by release in sanitary certification of the operations of gaseous sewer systems was issued in February 1994.

diffusion enrichment facilities were issued in The ANPR requested comments on the February 1993 and September 1994, re-appropriateness of current NRC regulations spectively. The rule covers both the certifi-and solicited comments on possible cation process and the standards to be used alternative approaches.

to judge acceptable performance for certifi-cation of the operations of the gaseous o

A proposed rule,10 CFR Part 34, on the diffusion enrichment facilities leased by the conduct of radiography using sealed sources U.S. Enrichment Corporation from the was issued in February 1994. He proposed Department of Energy.

rule responds to a petition for rulemaking (PRM-34-04) submitted by the International Low-Level Waste Disposal Union of Operating Engineers, Iocal No. 2.

A roposed rule,10 CFR Parts 30,40,70, The proposed rule represents a complete P

revision to this part of the Commission's and 73, on clarification of decommissioning regulations, including certification of funding requirements was issued in June radiographers and implementation of a 1994. The amendments would clarify when 1

two-person rule with radioactive sources.

decommissioning funding assurance was required and provide that assurance would o

A proposed rule,10 CFR 72.214, adding a be available after operations were terminated standardized HUHOMS cask to the list of and decommissiomng mitiated.

approved spent fuel storage casks was issued A final rule,10 CFR Parts 2,30,40,70, and e

m June 1994. The rule will merease the number of NRC-certified spent fuel storage 72, on timeliness in decommissioning of materials facilities was issued in July 1994.

casks available under a general heense.

The rule establishes timeliness criteria for decommissioning nuclear sites or separate o

A proposed rule,10 CFR 35.75, on criteria buildings or areas following permanent for release of patients adm,,stered rad *,

cessation of licensed activities.

mi active material was issued in June 1994. The proposed rule addresses the requests of three A proposed rule,10 CFR Parts 20,30,40,50, e

petitions for rulemaking: PRM-20-20 from 51,70, and 72, on radiological criteria for Dr. Carol S. Marcus and PRM-35-10/10a decommissioning was issued in August 1994.

from the American College of Nuclear The proposed rule is based on comments Medicine. The proposed amendment would received from seven workshops and a draft of specify a dose limit of 5 mSv (0.5 rem) rather the rulemaking published in February 1994.

than the limit of 30 mci currently specified.

e hyam o

A final rule,10 CFR Part 73, on physical fitness programs for security personnel at The NRC supports the Small Business e

Category I fuel cycle facilities was issued in Innovation Research (SBIR) program to July 1994. The amendment requires physical stimulate technological innovation by small fitness training programs as well as annual businesses, strengthen the role of small performance testing for specific security force business in meeting Federal research and personnel at facilities authorized to possess development needs, increase the commercial xiii NUREG-1266

liighlights application of NRC-supported research Participation in this program has continued results, and improve the return on investment since the program was established in FY from Federally funded research for economic 1982. In FY 1994, the NRC was supporting 17 and social benefits to the nation.

SBIR projects-in-progress.

t NUREG-1266 xiv

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PART 1--NUCLEAR SAFETY RESEARCH--REACTOR LICENSING SUPPORT l

1. REACTOR AGING AND RENEWAL This program is conducted to ensure that reactor components. The regulations, codes, or guides plant systems and related components perform as that pertain to the structuralintegrity of LWRs designed during normal operation and transient were written to ensure that possible combinations and accident conditions and ensure that their of material properties, loads, and flaws will yield functional integrity and operability can be adequate margins against failure of primary maintained over the operating life of the plant.

system components. The goal of the reactor vessel The program includes the reactor system pressure safety and piping integrity element is to ensure boundary. Failure to maintain pressure boundary that appropriate analytical procedures and integrity could compromise the ability to cool the inspection methods exist for assessing the safety reactor core and could lead to a loss-of-coolant of components during normal operation and accident accompanied by the release of hazardous transient and accident conditions and that fission products.

sufficient critical experiments are conducted to validate those procedures and methods.

1.1 Reactor Vessel Safety and Piping Ensuring the structural integrity of the pressure Integrity boundary has been at the center of several recent well-publicized regulatory issues-for example, the 1.1.1 Statement of Problem 1984 decision to require an accelerated schedule of five BWR inspections due to cracking in the The reactor system pressure boundary of a light-coolant pipes; the 1991 review of the Yankee Rowe water reactor (LWR)is the principal boundary plant; and the 1992 review of the Trojan plant enclosing the nuclear fuel core and the water steam generators. Additionally, incidents of coolant used to maintain suitably low tempera-cracks and leaks in piping and steam generator tures of the fuel cladding and to conduct the heat tubes have highlighted the need for materials from the fission reaction and convert the water data, analysis methods, and inspection techniques coolant into steam for electricity generation. The for these components.

primary system includes the reactor pressure vessel, primary coolant piping, primary pumps, Much of the work is completed and has been put and steam generators for pressurized water in practice through several regulations, regulatory reactors (PWRs). For boiling water reactors guides, and parts of the standard review plan, as (f WRs), the primary system includes the steam well as through national codes and standards. The line piping out to the first isolation valve. His remaining work is providing the basis for both boundary must be kept intact and fully service-confirming and revising some of the earlier able at all times to ensure that water coolant is regulatory positions, with the overall aim of always available to cover the fuel core so that the providing a stable, fully validated regulatory heat generated during power operation or from framework for ensuring the integrity of the decay following shutdown can always be safely primary pressure boundary for the foreseeable conducted away, thus precluding a core meltdown future.

accident. The principles of ensuring the structural integrity of the primary system components are embodied in the elements of fracture mechames 1.1.2 Program Strategy used to predict conditions for failure. nese The approach used for this element is to develop clements are (1) knowledge of the material analytical procedures for predicting continuing properties (strength, toughness, embrittlement, integrity or conditions-for-failure and to ensure etc.), especially the changes in those properties that an adequate experimental basis exists to that can occur as a consequence of nuclear validate those procedures. The most critical facet operations;(2) knowledge of the pressure and of pressure vesselintegrity is embrittlement of the other loadings that can be applied to the pressure vcssel steel as a result of exposure to components either from normal operations or neutrons escaping from the fuel core during from accidents: and (3) knowledge of the presence normal service. Experiments are conducted to and size of cracks or other flaws in the develop a base ofinformation on all the factors 1-1 NUREG-1266

1. Reactor Aging that will cause embrittlement to increase during review of the Yankee Rowe (Mass.) plant, and the service life. Much work is done to establish 1992 review of the Trojan (Ore.) plant steam correlations between small-specimen behavior and generators. The underlying concern in ensuring thick-section behavior to ensure that the analyses the integrity of the pressure boundary is that performed to assess structural integrity are valid.

failure to do so could compromise the operator's Similarly, the ability to predict integrity in piping ability to cool the reactor core and possibly bring has required testing of full-sized sections of pipe about a loss-of-coolant accident (LOCA) that having a variety of cracks to determine if such could be accompanied by a release of hazardous cracks could cause failure during either normal fission products.

service or accident conditions. For both vessels and piping, knowledge of the rate at which cracks Research in this area is a broad-based program grow is very important to ensure that a com-initiated in 1967. The original program was ponent will not fail during its forthcoming focused solely on the properties and fracture operational period. Experiments are conducted on behavior of the reactor pressure vessel-the large, a wide variety of pertinent materials under a thick-walled steel cylinder that houses and range of typical and expected exposure conditions supports the reactor core. As the full challenge of to determine the maximum bounding rates of ensuring the integrity of this critical component crack growth. Detection and sizing of flaws and was realized, the scope of the research program cracks in all primary system components are was expanded to include irradiation damage, conducted by the industiy through periodic service-induced cracking mechanisms, and inservice inspections at shutdowns. To ensure that methods for periodically inspecting the pressure the inspections reliably detect and accurately size vessel. Incidents of cracking and leaking in pipes the flaws, extensive tests are conducted with and steam generator tubes have accentuated the inspection teams drawn from the industry using need for materials data, analysis methods, and typical equipment and techniques on samples inspection techmques relevant to these whose flaw conditions are known. From the components.

results, it is possible to determine which tech-niques are effective and the magnitude of the The research program on pressure vessel safety error bands for flaw detection and sizing.

has expanded to meet these additional challenges.

Improvements in methods are proposed and flowever, much of the work is completed and has qualification procedures developed that can been put in practice through several regulations, provide better assurance of flaw detection in regulatory guides, and parts of the standard future inspections and for sizing flaws more review plan, as well as through national codes and accurately. Materials and components removed standards. The remaining work is providing the from actual service are used to measure material bases for both confirming and revising some of properties after years of service, to evaluate the the earlier regulatory positions, with the overall extent of corrosion, and to validate the existence aim of providing a stable, fully validated regula-of flaws that have previously been identified and tory environment for ensuring the integrity of the had their size estimated.

primary pressure boundary for the foreseeable future. The technical efforts in the research program-fracture evaluation and radiation 1.1.3 Research Accomplishments in FY 1994 embrittlement-are central to sound regulatory positions addressing the safe operation of the 1.1.3.1 Pressure Vessel Safety pressure vessel. For example, efforts to revise the basis for determining the allowable operating This area of NRC research focuses on ensuring pressure and temperature to preclude brittle the structuralintegrity of the reactor system failure of the pressure vessel drew on research pressure boundary, i.e., keeping it free from results from the pressure vessel safety program.

damage and leaktight. Ensuring the structural integrity of the pressure boundary has been at the

1. Fracture Evaluation. Addressing fracture center of several recent well-publicized regulatory analysis methods assumed a particularly large role issues-for example, the 1984 decision to require in the overall program during FY 1994. Fracture an accelerated schedule of five BWR inspections analysis work involves an ongoing program to because of cracking in the coolant pipes; the 1991 develop and reduce to practice advanced analysis NUREG-1266 1-2

1. Reactor Aging methods that will improve the ability to predict did not address complete details of all the the allowable pressures and temperatures for the potential loading conditions for reactor pressure pressure vessel and the ability to evaluate the vessels, nor did it include guidance on deter-integrity of the pressure vessels under design basis mining appropriate material properties for use in and hypothetical accident conditions. Basic work the evaluation method.

is being performed by researchers at Oak Ridge National Laboratory (ORNL), augmented by The RES staff published a draft regulatory guide research being performed at Brown University, at the end of September 1993 expanding the the University of Illinois, Texas A&M University.

ASME Code's guidance to include evaluation and the U.S. Navy's Naval Surface Warfare Center methods pertinent to all service loading con-(NSWC). The researchers are developing state-of-ditions, guidance on selection of transients for the-art analysis methods and evaluating them consideration at various service load levels, and against test data developed at ORNL, the specific guidance on estimating material proper-National Institute for Standards and Technology ties. During FY 1994, public comments on the (NIST), and the NSWC. The work performed draft guide were received, analyzed, and applied during FY 1994 has been very promising, and the to revising the draft guide for final reviews for programs have been vigorously pursued to permit publication during FY 1995.

evaluation of test geometries and loadings that are more typical of reactor pressure vessels in the The RES staff worked with researchers at ORNL ductile-to-brittle transition region of the material's to initiate development of technical bases in fracture toughness versus operating temperature p obabilistic fracture mechanics for revising behavior in the beltline region of the aged reactor Regulatory Guide 1.154 on plant-specific evalu-vessels. De researchers are also coordinating ation of pressurized thermal shock in PWR their work with international efforts through a pressure vessels. Additional research is being cooperative project on fracture analysis of large.

coordinated with efforts of the staff in thermal-scale experiments, under the auspices of the hydraulics and probabilistic risk assessment to Committee on Safety of Nuclear Installations.

revise Regulatory Guide 1.154, in accordance with Collaborative efforts with another European the SECY-92-283 document on the lessons Community program are well under way and are learned from the Yankee-Rowe reactor pressure expected to provide results from a large-scale test vessel integrity evaluation. It is planned that the that will closely simulate a reactor pressure vessel development of technical bases will be completed subjected to accident loads. This will provide a in FY 1995, and draft revisions to the guide will more realistic validation of the revised analysis be completed and published in 1996.

methods'

2. Radiation Embrittlement. Of special concern in ensuring the integrity of the reactor pressure During FY 1994, the results of several efforts were vessel is the embrittlement of the pressure vessel put to use in performing generic analyses of steel caused by neutrons escaping from the reactor pressure vessels fabricated from materials reactor core during normal operation. These with a low resistance to a " ductile tearing" failure neutrons impinge on the pressure vessel wall and, mode. In the early 1970s, the NRC recognized that through a complex process, reduce the ability of some pressure vessels were fabricated using steel the steel to resist fracture. The embrittlement plates and some weld types that did not provide increases with continued operation. To ensure the the high resistance to this failure mode exhibited continued safe operation of pressure vessels, the by most of the plates, forgings, and welds used in research program includes a significant effort to reactor pressure vessels. The NRC issued quantify the effects of neutron radiation em-Appendix G to 10 CFR Part 50 in 1973 to provide brittlement, to understand the mechanisms that explicit requirements on the Charpy upper-shelf control this process, and to find methods to energy-a measure of the ductile tearing re-mitigate the embrittlement and restore the sistance of these materials-for both new original fracture toughness.

construction and for operating plants. The American Society of Mechanical Engineers During FY 1994, the radiation embrittlement (ASME) published a Code Case N-512 (Section research efforts moved forward on several fronts.

XI. Division 1. February 1993) on this issue, but it Test reactor irradiations were completed by 1-3 NUREG-1266

1. Reactor Aging i

ORNL, using the University of Michigan test continuing work is generating important data reactor, to evaluate the effects of neution radia-relevant to the NRC's program to validate neutron tion on weld materials removed from the canceled fluence calculation methods and is providing Midland Unit 1 (Mich.) reactor pressure vessel.

technology transfer and validation of the methods The materials are representative of the so-called being used by the different laboratories. This

" limiting" material in several operating nuclear work contributed to the staff's effort to evaluate power plants. The materials are also being public comments and to revise a draft regulatory irradiated in the surveillance programs of an guide on calculational and dosimety methods for operating power plant as part of an NRC-industry determining pressure vessel neutron fluence.

c(x>rdinated research effort. When the results from each of these programs are available in the Work continued in FY 1994 to compile and late 1990s, they will provide important informa-evaluate embrittlement trends using the power tion about the embrittlement trends for these reactor pressure vessel material surveillance data.

materials and equally important information These data are reported to the NRC in accord-about the differences between test reactor and ance with Appendix H to 10 CFR Part 50 and power reactor irradiation conditions, as well as reflect embrittlement trends for reactor pressure about the mechanisms controlling embrittlement vessels irradiated under typical power reactor of these materials. Radiation embrittlement conditions. The work by ORNL to compile these research also includes a study of the effects of data into a comprehensive data base has provided thermal annealing and reirradiation on fracture the basis for work by Modeling and Computing toughness properties. Annealing tests of irradi.

Services to develop statistically based models for ated test specimens of typical reactor vessel stects predicting radiation embrittlement. The ORNL have demonstrated the degree of recovery of data base has also been used by the regulatory fracture properties using conditions generally staff in both plant-specific and generic evalua-proposed for annealing. Studies on the effects of tions. The ORNL work is a continuing effort while reitradiation are in progress.

the Modeling and Computing Services work is expected to be completed in 1995. This work will enable the NRC to evaluate the need for further During FY 1994, a collaborative research agree-ment was reached between the Japan Atomic revision to Regulatory Guide 1.99, which provides the methods for estimating radiation embrittle-Energy Research Institute (JAERI) and ORNL that provides for ORNL to examine pieces of the ment and is a fundamental part of the NRC's l

pressure vessel from the decommissioned Japan approach to ensunng pressure vessel safety.

Power Demonstration reactor. This examination Research to better understand the mechanisms of will focus on the changes m microstructure and radiation embrittlement continued in FY 1994, fracture properties caused by long-term exposure with significant advances being made by ORNL to irradiation and provides an opportunity t and the University of California at Santa Barbara, examme m depth a reactor vessel from an actual in conjunction with researchers in the United nuclear power reactor. The ORN,L studies will Kingdom, in modeling the complex interactions complement ongomg research being conducted by among the impinging neutrons and the atoms in JAERI at their Tokat research establishment.

the pressure vessel steel. This work is closely integrated with the experimental work being done During FY 1994, ORNL published neutron cross-in Europe. Understanding the controlling section libraries, BUGLE-93 and VITAMIN-B6, mechanisms is essential to confidently extrapo-that can be used in evaluating the neutron fluence lating empirical models of radiation embrittle-for power reactors, which is an essential mput m ment to unique operating circumstances. The estimatmg the level of radiation embrittlement for progress in the mechanisms research is providing reactor pressure vessels. In addition to the cross-assurance that the empirical models are section libraq work, researchers at ORNL have conservative and is helping to define the limits of worked with researchers in the Czech Republic, extrapolation for those models.

and other East European researchers, in performing calcu'ations to predict the results of In 1994, a proposed amendment to the NRC carefully controlled " benchmark" experiments regulations for nuclear power plants was issued to conducted by the Czech researchers. This clarify several items related to the fracture NUREG-1266 1-4

3

1. Reactor Aging toughness requirements for reactor pressure internationally funded research project, the vessels. The proposed amendment would clarify Second International Piping Integrity Research the pressurized thermal shock requirements for Group program, also being conducted by Battelle.

thermal annealing of a reactor pressure vessel. In hat work is examining the effects of simulated addition, a draft regulatory guide was issued to seismic loading on the fracture behavior of provide the information and criteria needed to cracked pipe and piping fittings. It is anticipated evaluate an application to perform a thermal that all the NRC's large-scale pipe fracture annealing treatment for a reactor vessel when research will be completed by early FY 1996.

j neutron radiation has reduced the fracture toughness of the vessel material. The thermal 1.133 Core Internal Components annealing treatment is expected to restore the fracture properties to acceptable levels.

Irradiation-assisted stress corrosion cracking (IASCC) of core internal components of both BWRs and PWRs has been observed and is 1.13.2 Piping Integrity becoming a more common problem as reactors During the 1980s, increased concern with inter-age and core materials accumulate higher fluence.

granular stress corrosion cracking in BWR piping Although many of the affected components can be systems and increased needs for research on other replaced, others are difficult or impractical to aspects of environmentally assisted cracking and replace. He susceptibility of materials to IASCC pipe fracture behavior led to increased research seems to be strongly dependent on mmor on piping integrity as part of an overall pressure vanations m matenal composition and micro-boundary integrity research program.

structure. Hus, nommally identical materials show large differences in resistance to IASCC.

Ongoing research by the NRC and by others is In FY 1994, work on large-scale pipe fracture and attempting to identify those characteristics that fracture characteristics of cast stainless steels make materials susceptible to IASCC. In wound down. However, other concerns in the p rticular, the NRC is sponsonng the irradiation piping system are still being investigated. The of a 1 rge gr up f m terials m the Halden potential for fatigue damage in reactor systems has long been recognized, but data from Japan a ctor in Wrwas manunat, n of thes,e matenals m

should help to clarify the role of matenal compo-indicate that the effects of the water coolant on sition, fluence, and the operatm, g environment.

the expected fatigue lives may not be adequately accommodated by the present ASME design 1.13.4 Inspection Procedures and Technologies rules. The Argonne National Laboratory (ANL)is collecting additional data on coolant effects on The NRC's approach to ensuring the integrity of fatigue in BWR and PWR chemistries and the reactor pressure boundary builds on the analyzing these data, and data obtained from overall " defense-in-depth" concept. The research other sources, as part of a program to develop program parallels this fundamental approach and better characterizations of fatigue behavior for includes programs geared to each of the major NRC use in evaluating remaining service life in considerations in providing structural integrity-aging plants. During FY 1994, new interim design analysis methods, material properties, and curves that incorporate the effects of reactor inspection techniques. The research program environments were published and are being addressing inspection procedures and tech-updated using newly developed ANL data.

nologies provides an independent basis for evaluating the efficacy and reliability of industry Pipe fracture research continued during FY 1994, inspection programs. The program includes with one of the major programs at Battelle studies of improved methods for selecting Memorial Institute drawing to a close. His components for inspection and strategies for research has provided the technical basis for the setting the required capability for the inspection flaw evaluation methodologies for piping that are method and inspection periods to provide a contained in the ASME Code Section XI and the reliable overall inspection. The program also deals technical basis for the NRC's leak-before-break with the inspection technologies and methods evaluation methodology. The balance of the necessary to ensure reliable detection and research is being conducted as part of an accurate sizing of flaws. Finally, the program 1-5 NUREG-1266

1. Reactor Aging includes a focused effort to transfer this tech-SAFT-UT system was fabricated for the NRC's nology to practitioners in the NRC regional and nondestructive examination mobile laboratory and headquarters offices.

operational training was provided to the NRC personnel who conduct independent field audits International Studies. The NRC is an active of ISI results. This system was successfully used participant and a leader in the Program for the for the first time in 1994 by the NRC staff for Inspection of Steel Components, Phase 111 (PISC performing audit inspections of piping at the III). This international program, organized in Peach Bottom (Pa.) nuclear power plant.

1986, is assessing the effectiveness of non-destructive testing technologies and procedures Field Trials for improved Eddy Current Inspection for the inservice inspection (ISI) of nuclear power ofSteam Generator 7hbing. Researchers from plant components. The participants in this pro.

ORNL participated in two inservice inspections of gram have invested an estimated $40 million in steam generator tubes at the Prairie Island 1 and the program, including contributions of materials, 2 (Mirm.) plants to test new eddy current probes inspection services, and manpower. The products anu signal analyses techmques and instrumen-from this program will assist regulators and code tation under field conditions. The new probes use bodies in establishing technical bases for multiple " pancake" type coils for better improving ISI requirements, sensitivity. The design incorporates several coils around the circumference of the probe so that The focus of the PISC 111 program is on the rotation is not needed, and the probes can be nondestructive testing of realistic LWR primary translated along the tube at high speed. Thus, the circuit components containing realistic flaws, new array probes offer the best features of the During FY 1994, results were reported for a flaw two currently used probes: high speed and sizing study in a reactor pressure vessel, detection sensitivity. In the field tests, the new probes and sizing of flaws in dissimilar metal weldments, produced signal levels from flaws that were 5 to and the detection and sizing of flaws in stainless 10 times greater than the current-practice rotatmg steel piping. This work shows that some inspec-pancake coil (RPC) inspection. The ORNL tors were effective and had a high flaw detection inspect, ion was 75 times faster than the current rate with a corresponding low false call rate.

RIy mspection and nearly as fast as the bobbm, flowever, other inspectors demonstrated an e il mspection. The high sens,tivity and high i

i ineffective performance with a low flaw detection mspection speeds that can be achieved with these rate and high false call rate. For flaw depth sizing, Probes would permit inspection of the entire,

there were a few inspectors and conditions in length of tubes in generators that are experienemg which performance was acceptable. Ilowever, the considerable degradation in a fraction of the time overall performance was poor with low correlation currently required for m, spectmg just short lengths f tub, g with the RPC probes. Ihe new array m

and large errors between the depth estimates and the true depth size.

Probes are also sensitive to axial cracks, cir-cumferential cracks, and volumetric defects, which is a significant improvement over current bobbin Improved Ultrasonic Dctccrion and Sizing of Flaws, An improved method for more reliably detecting cml probes.

flaws and sizing them with greater accuracy in.

1.1.3.5 United States-Russian Federation / Ukraine j

L.WR primary circuit components is the Synthetic Cooperative Agreement Aperture Focusmg Techmque for Ultrasome i

Testing (SAFT-UT). The SAFT-UT technology is The NRC staff and researchers from ORNL and based on physical principles of ultrasonic wave the University of California at Santa Barbara propagation and uses computers to process the participated in September 1994 workshops and data to produce high-resolution, three-meetings in Kiev, Ukraine, and Moscow, Russia, l

dimensional images of flaws to aid the inspector as part of the Joint Coordinating Committee on in locating and sizing them. The SAFT-UT Civilian Nuclear Reactor Safety (JCCCNRS).

technology has been developed through extensive Working Group 3 on " Radiation Embrittlement" laboratory testing and validated through blind held a 2-day workshop in Kiev to discuss pressure trials. The technique performed well in the PISC vessel integrity issues, followed by a 4-day working 111 pressure vessel flaw sizing studies. A group meeting in Moscow. A total of 16 papers NUREG-1266 1-6

1. Reactor Aging were presented during the workshop (eight from of Operating Licenses for Nuclear Power Plants,"

the United States and eight from the Ukrainian was issued in final form in December 1991. The participants), with a total of 24 papers presented initial form of draft Regulatory Guide DG-1009, during the working group meeting (eight from the

" Standard Format and Content for Applications United States and 12 from the Russian to Renew Nuclear Power Plant Operating participants).

Ljcenses," was issued for comment in 1991. Since publication of the final license renewal rule, a NRC staff members and representatives of the number of significant policy issues have been Department of Energy and the national identified. As a result, the Commission is in the laboratories participated in a September 1994 process of amending the license renewal rule.

meeting in Moscow of the JCCCNRS Working Proposed revisions were published in September Group 12 to discuss issues related to nuclear 1994.

power plant aging and plant life extension. The United States delegation presented 10 papers 1.2.2 Program Strategy dunng the workmg group meetmg. Subsequent Working Group 12 activities have included the NRC staff effort in aging is being pursued in exchange of more information on special topics several areas, including technical and scientific and preparations for the sixth Working Group 12 research to identify the effects of aging on the key meeting to be held in the United States in the safety-related components of the plant and to summer of 1995.

examine methods for mitigating such effects.

Specifically, the strategy is to achieve, relative to e ch cmnponent, the foHowing resuks:

1.2 Aging of Reactor Components 1.

Identify and characterize aging and service 1.2.1 Statement of Problem wear effects that, if unmitigated, could cause degradation of structures, systems, and Aging affects all reactor structures, systems, and components and thereby impair plant safety.

components in various degrees and has the potential to increase risk to public health and 2.

Develop methods of inspection, surveillance, ensure cont,cffects are not controlled. In order to and monitoring and of evaluating residual life safety ifits muous safe operation, measures must of structures, systems, and components that be taken to monitor key structures, systems, and will permit compensatory action to counter components and interfaces to detect agmg significant aging effects prior to loss of safety degradation and to mitigate its effects through function.

mamtenance, repair, or replacement. For an older plant approaching the end of its design life and 3.

Evaluate the effectiveness of maintenance, 1or which extended operatmn beyond its ongmal repair, and replacement practices, current b,eense penod of 40 years is contemplated, agmg and proposed, in mitigating the effects and becomes a critical concern and will clearly be diminishing the rate and the extent of crucial to any assessment of the safety degradation caused by aging.

implicatmns of heense renewal.

He NRC and the nuclear industry have initiated 1.2.3 Research Accomplishments in FY 1994 a significant effort aimed at renewing plant 1.2.3.1 Aging Research licenses beyond their ongmal term of 40 years.

According to an early Department of Energy Aging affects all nuclear reactor structures, study, the pmjected net benefit to the United systems, and components. If aging degradation is States economy can be on the order of $230 not detected and corrected, it can increase risks billion through the year 2030, assuming a 20-year to public health and safety. Failures of safety-period of extended operation for current plants.

related components have occurred in the past ne benefit reflects both the lower fuel cost of the because of such age-related degradation processes nuclear plants and reduced outlays for replace-as corrosion, embrittlement, wear, and fatigue.

ment of generating capacity. The license renewal The objective of aging research is to develop the rule,10 CFR Part 54," Requirements for Renewal technical bases for continuous safe operation of 1-7 NUREG-1266

1. Reactor Aging nuclear power plants as they progress through torque and thrust requirements for operating the their design life; to define the operative aging MOVs when necessary, particularly to mitigate j

mechanisms; and to confirm existing and/or accident conditions. Friction experiments were developing recommendations for new detection conducted on samples of corroded materials and mitigation methods in order to prevent or typical of certain valves. Although the test results mitigate the deleterious effects of the aging indicated that the friction due to corrosion does process.

increase the thrust requirements, many new questions about the need for simulating actual The Nuclear Plant Aging Research (NPAR) loadings, temperature, and other parameters must program continued to study the aging-related be answered before the magnitudes of the degradation of performance of safety-significant increases in friction can be determined and components and systems and methods for validated. Subsequent investigations to answer mitigating and managing the aging of these these questions were made, and a better con-components and systems in commercial nuclear trolled series of friction experiments will start in power plants. During FY 1994, preliminary or late 1994 and will be completed in FY 1995.

compreh:nsive aging assessments were completed or final reports were issued for the following The information will be used in determining if the safety-related components, systems, and MOVs comply with Generic Letter 89-10, associated special topics:

" Safety-Related Motor-Operated Valve Testing and Surveillance."

Chemical and Volume Control System for o

Pressurized Water Reactors Air-Opcm/cd 11dres. An evaluation of aging and (NUREG/CR-5954) service wear of air-operated valves was completed and reported in NUREG/CR-6016, " Aging and Containment Cooling Systems Service Wear of Air-Operated Valves Used in o

(NUREG/CR-5939)

Safety-Related Systems at Nuclear Power Plants."

The evaluation was based on data taken from the Reactor Core Isolation Cooling System Nuclear Plant Reliability Data System (NPRDS) o (NUREG/CR-5692) for the period January 1,1988, to December 31, 1990, which, after removal of inconclusive data, o

Selected Fault Testing of Electronic Devices involved reports of 1503 failures of varying de-(NUREG/CR-6086) grees. The data were processed in ways to reveal trends and the effectiveness of testing, and it was o

Managing Aging in Nuclear Power Plants-found that neither were there trends nor was Insights from NRC Maintenance Team testing especially effective in detecting degrada-inspection Reports (NUREG/CR-6016) tion that led to the failures. The results showed that failures involving complete loss of function o

Accumulators were usually the result of failures in the controls or the valve actuator vs. some failure of the valves o

Isolation Condenser Systems themselves. While many of the controls and actuator components are known to have high Air-Operated Valves (NUREG/CR-6016) failure rates wherever they are applied, they have o

not usually exhibited signs of degradation prior to o

Characterization of Check Valve Degradation complete failure. It was concluded from this that and Failure Experience (NUREG/CR-5944, a basis could be developed for replacing certain Vol. 2) components on the basis of aging environment alone in any cases where failures are Aging Effects on Motor-Operated lidre Perform-unacceptable.

ance. In 1994, initial research efforts were completed to identify motor-operated valves Check Mdres. The check valve degradation and (MOVs)in typical PWR and 11WR plants that are failure study completed in 1993, covering failures i

most susceptible to internal environmental occurring in 1984 to 1990, was expanded to corrosion. He NRC concern is whether corrosion examine and process NPRDS records on failures of internal valve parts can significantly affect the of check valve internals occurring in 1991. After NUREG-1266 1-8

-- -~-----

L Reactor Aging screening the data base to eliminate unsuitable published related to check valve testing and records,401 failures remained to be analyzed. As condition monitoring. They are ORNL/NRC/

in the past study, a primary goal was to identify LTR-93/6, " Review of Monitoring and Diagnostic any correlations of valve failure rates with plant Methods for Check Valves," and ORNUNRC/

age, valve size, system of service, manufacturer, LTR-94/04, " Utility Survey PWR Safety Injection etc. A further goal of the study was to identify any Accumulator 'Ihnk Dischnge Check Valve apparent trends in failure rates, failure detection, Testing."

severity of failures, etc. With the cooperation and assistance of the Nuclear Industry Check Valve Aging Assessment and Mitigation ofMajor LWR

' Group, additional information was obtained on Components. Of intrinsic importance to reactor most of the valves regarding specific valve types, aging research is the assessment and mitigation of specific design features, valve configuration, valve aging damage to major components and struc-application, and applicable inspection programs.

tures. The objective of this aging assessment task, "the latter information was vital to a new activity an element of the NPAR program, is to identify, that will provide the independent source of data develop, and evaluate various aging management to allow NRC to respond to expected industry techniques for the major LWR components and requests for extension o check valve test and structures. The approach is to gauge the degra-r inspection intervals. The results reported in dation of the major LWR components and NUREG/CR-5944, Vo'ume 2, "A Characterization structures by the synergistic influences of the of Check Valve Degredation and Failure Experi-various aging mechanisms affecting the per-ence in the Nuclear Power Industry-1991 Fail-formance of these components and structures.

ures," showed some positive trends from those reported in Volum; 1. For example, failures Research completed in this area in FY 1994 detected by abne. mal occurrences declined from focused on completing the development of 19% to 5%, an$ the percentage of significant msights for aging management of selected LWR failures decreased from 53% to 36% Also, the components and structures m order to ensure most effecave means of detecting failures continued safe operation. The studies also continued to be by programmatic inspections-included an evaluation of advanced inspection 77%, up from 59% in the earlier study. Thus, it and monitoring methods for characterizmg the appears that degradation is being detected earlier agmg damage. The results should prove useful to in the failure process.

the NRC in its task of resolving safety issues associated with LWR aging degradation and developing policies and guidelines for operating In response to the improvement in effectiveness of license renewal. The major components completed programmatic tests and inspections and condition in 1994 were the LWR metal containments and monitoring techniques, as well as the improve-the LWR reinforced and prestressed concrete ment in availability and reliability of information containments. Results are documented in a on failures, a new project has been initiated to multivolume report, NUREG/CR-5314. A draft help NRC respond to anticipated requests to report (NUREG/CR-5824) discussing the extend test and inspection intervals for those identification of advanced monitoring methods for check valves that have exhibited low failure rates.

estimating stresses causing fatigue damage has Independent evaluations of the available infor-also been completed and is finishing internal NRC mation and inspection and monitoring techniques review. Publication is scheduled for 1995.

will be produced to provide the basis for approv-ing the requests. In addition, national consensus PRA-Based Methodologyfor Aging Assessments and standards groups will be supported, and industry Priority Assignments. The risk-based methodology groups will be contacted to ensure that the for assessment of aging in nuclear power plants direction and focus of the evaluations are con-and for defining priorities among risk sistent with industry activities. Further, partici-contributions and maintenance activities pation in the standards activities may contribute (published in previous years as NUREG/CR-5587 to the acceptability of revised standards expected and NUREG/CR-5510)is subject to uncertainties to be produced in response to improved data and because of limited available aging data and also condition monitoring. In addition to NUREG/

because of certain modeling assumptions.

CR-5944, Volume 2, two other reports have been Research has focused on developing sensitivity 1-9 NUREG-1266

1. Reactor Aging and uncertainty analyses to address data and Work continued in FY 1994 to set priorities for j

modeling uncertainties and to validate risk-based environmental stressors associated with advanced j

methods. This work was documented in draft digital instrumentation and control (I&C) systems NUREG/CR-6045 in 1994 and submitted for in nuclear power plants, based on their risk NRC review, significance. Analog I&C systems in nuclear power plants are being replaced by digital sys-tems. Digital I&C systems are vulnerable to The application of age-dependent risk method-common environmental stressors, such as ology requires age-dependent component failure moisture / humidity and temperature, and the rates. But age-dependent component failure rates effects of such stressors are being identified and are not generally available and need to be esti-measures developed to rank them. The risk-based mated from limited recorded plant failure data approaches are being tested for the I&C systems and plant mamtenance logs. A maj,or h,mitation of using plant-specific PRAs. A draft report was the age-dependent methodology has been the lack issued in FY 1994 identifying the approaches that of recorded component aging data and can be used to accomplish this work. This effort approaches to develop aging failure rates based has required more time than expected because of on the available information. To address this the lack of failure data for these components in limitation, an approach was developed to nuclear plant applications.

incorporate age dependence in probabilistic risk assessments (PRAs) that does not require Aging of Passive Components. In earlier research absolute age-dependent component failure rates.

efforts, a methodology was developed for Instead, the aging of a component is expressed in including the effects of aging on passive com-terms of relative aging rates that are found to be ponents (pipes, structures, and supports) in a fairly constant across different components and PRA model to determine the resulting impact on different plants. A draft report (NUREG/CR-plant risk. The methodology is based on prob-6067) was completed on the aging data assessment abilistic structural analysis for calculating the methodology. Because of the importance of the failure probability of these components. The role of PRAs in future risk-based regulations, methodology was documented in a final draft NUREG/CR-6067 was extensively reviewed by the report (NUREGiCR-5730) and submitted for i

NRC staff in FY 1994. Many comments were NRC review in FY 1994. Although the method provided, and it is expected that they will be meets the condition of the contract, the method is incorporated in FY 1995.

not compatible with the method for including aging of active components in a PRA. Since a total capability for determining the effects of In a previous year, an important application of aging on plant risk is the ultimate goal of this the risk-based methods resulted in the develop-project, both active components (e.g., pumps, ment of PRA-based approaches for identifyin8 valves) and passive components should be in-safety-related MOVs having the most impact on cluded in the same model. Because of this result, plant risk covered under Generic Letter (GL) a more practical and simple approach, which 89-10. " Safety-Related MOV Testing and Sur-considers the results of NUREG/CR-5730, is veillance." Dynamic tests and surveillance tests, in being completed to meet the goals of the project.

accordance with GL 89-10, cotild then be The documentation of this integrated method will performed on those MOVs with the largest risk be completed in FY 1995.

impact. Relative risk importance of single MOVa and the interaction of multiple MOVs can be Equipment Operability. For the past 5 years, analyzed using this approach. A draft report significant progress has been made by the NRC in documenting the results of this work was issued advancing the state of the art of MOV technology.

for NRC review in FY 1994. This work has Full-flow experiments were conducted in prior provided the technical basis for evaluating two years that led to the determination that MOVs are different submittals by licensees for ranking their not being calibrated properly to ensure their respective MOVs for tests in accordance with GL operation when required. Since that time, evalua-89-10. These NRC evaluations resulted in identify-tions of the vast amount of data from those earlier ing many weaknesses with the submittals, which experiments and from additional smaller-scale the licensees are now resolving.

tests have provided breakthroughs in NUREG-1266 1-10

1. Reactor Aging understanding this complicated technology.The Environmental Guahfication Research. Questions MOV engineers from nuclear power plants in the concerning the environmental qualification (EQ)

United States and Europe consider the NRC of electrical equipment used in commercial MOV researchers as leaders in this field.

nuclear power plants have recently become the subject of significant regulatmy interest. Initial The timely and effective transfer of research questions centered on whether compliance with results to the NRC regulatory staffs has been the EQ requirements for older plants is adequate useful for determining the integrity of MOVs and to support plant operation beyond 40 years. After is a high-priority objective of this project. Spe-subsequent investigation, the NRC staff concluded cifically, the results that were transferred con.

that questions related to the differences m EQ sisted of technicalinformation for determining requirements between older and newer plants whether licensees' MOVs are in compliance with c nstitute a potential generic issue that should be regulatory document GL 89-10. This document evaluated for backfit, mdependent of license requests licensees to develop MOV programs that renewal activities.

will ensure MOV operability throughout the lives of the respective nuclear power plants. Since there EQ testing of electric cables was performed by are an average of 150 safety-related MOVs in each Sandia National Laboratories (SNL) under con-nuclear power plant in the United States, it is tract to the NRC in support of license renewal imperative that these MOVs are accurately activities. Results showed that some of the calibrated to ensure their performance as environmentally qualified cables either failed or

required, exhibited marginal insulation resistance after a simulated plant life of 20 years during accident s mula n s inhated Gat Om W process for 1he results of research completed in prior years as well as more recent findin8s in 1993 and 1994 smne e ctric capes may not k consewatn have also provided the techmcal basis for issumg These results raised questions regarding the EQ several other NRC regulatory documents identi-process, including the bases for conclusions about fymg potential MOV problems of which the the qualified life of components based on artificial heensees must be aware. In addition, numerous aging prior to testing.

supplements to GL 89-10 have been issued for licensee compliance as a result of the NRC As the first step m. developing a research pro-research findings as well as other experiences 8f"S* RES held a pubhc workshop to obtam from industry MOV testing.

techm. cal input from industry representatives, as well as experts in the field. The Environmental Qualification Workshop was held on November During FY 1994, efforts continued on developin8 15-16,1993. The workshop proceedings were the technical basis for evaluating efficiencies for issued in May 1994 as NUREG/CP-0135.

AC and DC motor-operators. When completed, this information will also be provided to the NRC regulatory staff for their use m determmmg The workshop provided a unique opportunity for whether these devices, which supply the power to the open exchange of ideas and information valves, are achievmg the outputs as claimed by the among industry personnel, researchers, equipment manufacturer, particularly under degraded voltage manufacturers, and regulators involving EQ and elevated temperature conditions.

issues, descriptions of state-of-the-art activities in condition monitoring and research techniques.

The discussions included several recent equip-All the research results obtained over the past 5 ment failures and their causes at operating years are being used by NRC in evaluating the facilities, and presentations describing current Electric Power Research Institute (EPRI) topical licensee actions related to monitoring normal report on MOVs.The topical report contains the service conditions, such as on-line temperature information obtained from the EPRI MOV monitoring in specific plant locations. Additional research program that the licensees will use in discussions centered on the limitations of con-complying with GL 89-10. The NRC evaluation of dition monitoring techniques currently available, the topical report started late in FY 1994 and will qualification testing techniques, and pre-aging be completed in FY 1995.

techniques. Several participants expressed a 1-11 NUREG-1266

1. Reactor Aging concern that any testing could lead to additional existing supplementary requirements for IST regulatory burden on licensees.

of containment isolation valves; 3.

Allow alternatives to the ASME Codes, which 1.2.3.2 Engineering Standards Support would permit licensees to use later editions or addenda of the ASME Codes as alternatives; The national standards program is coordinated by the American National Standards Institute (ANSI). ANSI provides procedural guidelines to 4.

Identify safety-significant code changes that help ensure that participation in the private sector the staff has determined are necessary for standards development process is sufficiently imposition on licensees, specifically Appendix broad based and that input from individual VIII, " Performance Demonstration for interests are fairly considered. NRC participation Ultrasonic Examination Systems," of Section in this process is compatible with Office of XI of the ASME Code, which provides rules Management and Budget Circular A-119, dated for qualification of personnel and equipment October 26,1993, which provides policies for used to perform inservice nondestructive Federal participation in the development and use eraminations on nuclear power plant of voluntary standards.

components; 5.

Establish a new regulatory guide that would ne NRC staff is very active on the ASME codes endorse alternatives to the baselm, e ASME and standards writing committees because Code rules, meluding the use of later editions portions of the ASME Boiler and Pressure Vessel and addenda of the ASME Code; and (BPV) Code have, since 1971, been incorporated by reference into 10 CFR 50.55a, " Codes and 6.

Establish a new regulatory gm.de that would Standards," of the NRC regulations in order to document NRC review and acceptance of establish requirements for the construction, OM Code Cases.

inservice inspection, and inservice testing of nuclear power plant components. Section 50.55a These act. ions would support substantial reduc-has periodically been amer.ded to update the references to include more recent versions of the tions m the current regulatory burden on hcensees ASME BPV Code. In 1994, work continued on as determmed by the staff durm, g the review of a changes to the regulations that the NRC staff is CBLA request. Work also contmued on a rule-considering. These changes, which include actions makmg that would, for the first time, mcorporate to address a cost-beneficial licensing action by reference Subsection IWE and Subsection (CBLA) request, would:

IWL,Section XI, ASME BPV Code. Subsection IWE provides rules for the inservice inspection of metal containments and the liners of concrete 1.

Eliminate 1he 120-month update require-containments. Subsection IWL provides rules for ments for h,eensees m, service mspection (ISI) the inservice inspection of concrete containments i

and mservice testmg (IST) programs; and their post tensioning systems. The proposed rule was published for public comments in the 2.

Establish baseline regulatory requirements Fedemi Register on January 7,1994 (59 FR 979).

for ISI and IST programs, maintain a Comments were received from 25 separate specified ASME BPV Code edition for ISI sources. The comments have been addressed, and and incorporate, for the first time, by the draft final rule is under development.

reference the 1990 Edition of the ASME Code for Operation and Maintenance of ASME Code Cases provide alternatives to the Nuclear Power Plants (OM Code) for IST, rules specified in the ASME BPV Code.

expand the scope of 9 50.55a to include Regulatory Guides 1.84,1.85, and 1.147 identify inservice testing and examination of safety-those Code Cases for design and fabrication, related snubbers to allow licensees an option materials, and inservice inspection, respectively, to delete existing technical specification that the NRC has found to be acceptable. These snubber test requirements and use the ASME regulatory guides, which are updated on a regular OM Code for IST of snubbers, and delete the basis, were revised and issued in 1994.

NUREG-1266 1-12

1. Reactor Aging i

1.23.3 StructuralIntegrity Regulatory applications of this research include:

Concrete structures play a vital role in the safe (1)impr ved predictions oflong-term material operation of all light-water reactor plants. In and structural performance and available safety m rgms at future times;(2) establishment of general, the performance of concrete structures in nuclear power plants has been good. However, Hmits on exposure to environmental stressors; l

(3) the ability of NRC to reduce its total reliance there have been several instances where the f licensing on inspection and surveillance capability of concrete structures to meet future functional and performance requirements has thmugh development of a methodology that will been challenged because of problems arising from e t e mtegnty of stm tures to k assessed en either improper material selection, construction (either pre-or post-accident); and (4) improve-and design deficiencies, or environmental effects.

ments in damage inspedm, n metpology Gmugh j

l Potential incorporation of results mto national ExamP es of some of the Potentiall more serious Y

standards that could be referenced by standard i

meidents m. elude post-tens.ionmg anchor head review plans.

failures, leachmg of concrete m tendon gallenes, voids under vertical tendon-bearing plates, con-tainment dome delaminations, corrosion of steel 1.2.3.4 License Renewal Regulatory Standards tendons and rebars, water intrusion through basemat cracks, and leakage of corrosion A final rule (10 CFR Part 51) concerning the inhibitor from tendon sheaths. Such incidents environmental review for renewal of a nuclear indicate that there is a need for improved power plant operating license is under develop-surveillance, inspection and testing, and ment. The proposed rule was published for public maintenance to enhance the technical bases for comment in September 1991. Over 120 comments assurance of continued safe operation of nuclear were received on the technical analyses and power plants.

certain procedural aspects of the proposed rule.

Concern was expressed that the proposed rule The structural aging (SAG) program is addressing would constrain public comment on environ-the aging management of safety-related concrete mental issues at the time of license renewal review structures in nuclear power plants for the purpose for an individual nuclear power plant. In FY 1994, of providing improved technical bases for their four public workshops were held to discuss continued service. lo accomplish program approaches to resolve specific concerns expressed objectives, the SAG program has conducted by the Agreement States over the treatment of activities under four major technical task areas:

need for generating capacity and alternative (1) program management (2) materials property energy sources. All comments are being data base,(3) structural component assessment /

considered in developing the final rule, the generic repair technologies, and (4) quantitative method-environmental impact statement, and other ology for cominued service determinations. The supporting documents. It is expected that the final final program report will be completed in mid-rule and supporting documents will be published 1995.

in FY 1995.

1-13 NUREG-1266

l

2. STANDARD REACTOR DESIGNS 2.1 Systems Performance of Advanced little additional research is needed to support Reactors their certification or licensing: the four non-com cutional reactor design concepts have not 2.1.1 Statement of Problem ped the point where certification is expected m the near term. Consequently, the current The Commission has issued a polics statement on emphasis of the advanced reactor research the regulation of advanced nuclear pawer plants pr gr m is on developmg the mformation needed (51 FR 24643) that states that the N KC will review to support the certification of the AP-600 and and comment on applications for certification of SBWR reactor designs and making appropriate new design concepts witn special emphasis on nmdifications of existmg regulatory assessment vendor test programs for confirming the perform-methods to accom!nodate the review of the ance of novel safety systems. As part of this pro, unique, p ssive salety systems of these reactors.

gram, the NRC will develop, rewiew, and imple-ment advanced reactor safety and policy issues in 2.1.3 Research Accomplishments in FY 1994 its review of such advanced reactor concepts and 2.1.3.1 Support for AP-600 Design Review carry out any mdependent research and analysis determined to be necessary to verify that Confirmatory testing and analysis of the advanced reactor designs have the potential to Westinghouse AP-600 reactor and plant systems provide enhanced margins of safety and that plant are being performed to provide additional con-safety systems will adequately perform their fidence in the NRC's evaluation of the safety of intended functions, the AP-600 design. The most cost-effective means of performing the desired tests was to modify an existing full-height, full-pressure test facility rather 2.1.2 Program Strategy than build a new one. Screening revealed that the Research programs have been m..tiated to support best choice was the Rig of Safety Assessment i

the certification and beensmg of advanced reactor (ROSA) large-scale test facility in the Japan Atomic Energy Research Institute (JAERI). 'Ib designs bemg developed by the nuclear m, dustry confirm these initial results and to determine the and the Department of Energy. Several different designs are bemg considered for certification extent of modification necessary to simulate the under 10 CFR Part 52, Early Site Permits:

AP-600, the Idaho National Engineering Labora-tory was contracted to perform a comparative Standard Design Certifications: and Combm.ed Licenses for Nuclear Power Ilants. These designs study between ROSA and the AP-600 using the RELAP5 code.

fall into two groups. The first group consists of four evolutionary and passive advanced LWR A comparison between the existing ROSA facility types: the advanced boiling water reactor and the AP-600 design showed that modifications (ABWR), the CE System 80+, the AP-600 to the ROSA facility would be needed. Facility advanced pressurized water reactor, and the modifications were completed in February 1994 by simplified boiling water reactor (SBWR). The Sumitomo Heavy Industries, which constructed second group consists of four more nonconven-the ROSA facility and has been maintaining and tional advanced reactor types: the Process operating it for the past several years as a i

Inherent Ultimate Safety (PlUS) reactor, the contractor to JAERI.

Canada Deuterium Uranium (CANDU 3) heavy-water-cooled reactor, the Advanced Liquid-Metal-As of January 30,1995,12 tests have been con-Cooled Reactor (ALMR), and the Modular ducted in ROSA simulating various accident Iligh-Temperature Gas-Cooled Reactor scenarios that would challenge the unique safety (MIITGR).

systems employed in the AP-600 reactor. Small-break loss-of-coolant accidents in different The two evowianary LWR designs (ABWR and locations and different sizes in the primary i

System 80+) have been judged to be similar coolant loop were investigated. Test results

)

enough to the current generation of LWRs so that obtained so far indicate that the reactor will be 2-1 NUREG-1266

2. Standard Reactor Designs

\\

1 effecth cly cooled, as designed, under various 2.1.3.3 Support for CANDU 3 Design Review accident conditions. However, two issues have S

M. i were completed in early FY 1994 been raised by the tests performed to date. One in connection with a preapplication review of the involves the large thermal g~radient found in the CANDU 3 design, and four sigmficant research cold leg where cold water from the passive pr ducts were produced from these studies.The residual heat removal system enters the primary first was a summary of Canadian regulation of system. The other is the possibility of having a CANDU reactors,that identified some contrasts water hammer in the cold leg or in the upper w th NRC regulations. The second identified and plenum as a result of direct contact between classified event sequences (i.e., accident sce,

subcooled water and steam. These issues are n rios), plant systems, and operator actions m being carefully evaluated to determine any safety w ys that would facilitate the application of NRC implications that might alfeet the acceptability of regulations. The third was an assessment of data the AP-600 design' bases that exist as the basis for CANDU safety analyses. And the fourth was a preliminary an lytical study, using Canadian computer codes, Two additional tests will be conducted by June "Ievents mvolymg the design s pos,itive coolant-1995, with six more in 1996. The RELAP5/ MOD 3

.d coefficient of reactivity. Additional work has v

computer code is being assessed against the test been planned to support the formal review for data from ROSA, and necessary modifications are design certification, but that work will not be capabils.ade in the code to improve its prediction being m nutiated until the bulk of the work on the AP-600 ty.

and SBWR is finished.

2.1.3.4 Human Reliability 2.1.3.2 Support for SBWR Design Review Efforts are continuing to develop methods for ssessing the impact on risk of changes in human This program provides confirmatory testing and perf rmance due to the mtroduction of advanced computer code assessment for the General digital displays and controls.

Electric SBWR There are three elements in the program. First, a well-scaled, integral SBWR test Research to establish a technical basis for facility has been designed and will be built at minimum shift staffing (operations) for advanced 1 urdue Umversity. The test facility is called control room designs was initiated in FY 1994 at I,UMA (Purdue Umversity Multi-Dimensional the Halden reactor project. The research is based Integral Test Assembly). Second, tests will be on workload and task allocation studies con-performed in the i UMA facility to produce data ducted on power plant and advanced control for a broad spectrum of loss-of-coolant accidents room simulators and through task network and transients postulated for the SBWR. Third,

""d#li"8' the PUMA data will be used to (a) assess the capabilities of the thermal-hydraulic RELAP5 code for SBWR analysis (b) assess the integral 2.2 Engineering Issues for Advanced performance of the SBWR-unique safety systems Reactor Designs that maintain core and containment cooling, and (c) identify and understand the important 2.2.1 Statement of Problem phenomena observed in the tests.

The safety evaluation of standard / advanced reactors involves (1) new technologies, (2) new PUMA is a low-pressure (150 psig) and reduced-environmental conditions, (3) new systems, height facility. Its design was completed in 1994, (4) new design approaches, and (5) new require-and procurement and fabrication of various ments. New technologies include the expanded use components and instrumentation is under way.

of advanced digitalinstrumentation and control The PUMA facility will be completed by April systems and novel modular construction methods.

1995 and will be ready for testing by July 1995. A New environmental conditions are reflected in the i

total of approimately 40 tests will be performed low-pressure operation of check valves for a i

by August 1996.

planned 60-year operational period. New systems NUREG-1266 2-2

2. Standard Reactor Designs are found in the passive reactor designs where reviewing valve ap; lications and (2) in providing air-conditioning power is not needed for accident information on the behavior of different valve response and the safety classifications differ designs based on research results and other valve markedly from present standards, thereby experiences. Based on the research completed, the affecting probabilistic risk assessments (PRAs)

NRR staff has established a position in the safety and the fragility information needed to perform evaluation reports for advanced boiling water these PRAs. New design approaches incorporate reactors and CE System 80+ regarding the design the use of experience data as a method of seismic requirements for piping systems that interface qualification and the elimination of the operating with the primary system to minimize the potential basis carthquake from design. New requirements for interfacing-system loss-of-coolant accidents.

deal, for example, with NRC goals for contain-The research studies are continuing to define the ment performance under severe accidents and the extent to which the piping systems outside the ability to withstand interfacing-system loss-of-containment should be designed to these coolant accidents. Research is needed to support requirements, considering cost-benefit aspects.

technical positions in safety evaluaCon reports prepared by the NRC staff for the Final Design 2.2.3.2 Ex erience-Based Approach to Seismic Approval, the Design Certification, and the Qualification of Equipment Combined License for standard / advanced reactors dealing with these issues.

The ALWR Equipment Qualification Panel has resolved all issues pertaining to the experience-based approach for seismic qualification of 2.2.2 Program Strategy Group 1 items. Group 1 equ pment has a mature The objective or strategy of engineering research design with little design variability and,in general, has demonstrated characteristics of inherent is to verify the safety and quantify the margins in standard / advanced reactors for those features or seismic ruggedness. Items included in Group 1, criteria substantially different from currently are (1) horizontal centrifugal pumps with electric operating plants and to ensure that NRC goals motor dnvers, (2) vertical centrifugal pumps with and policies are implemented successfully. To deal electric motor drivers, (3) manual valves / check with these issues, analyses, evaluations, data valves, (4) valve assembh,es with electnc motor collections, and limited testing are planned. In perators (MOVs), (5) thermal element assem-some cases, research is used to independently blies, and (6) diesel ge,nerator umts. The ALWR confirm or modify vendor / utility proposals for Equipment Qualification Panel met with vendors design. fabrication, construction, and inspection.

of batteries and transformers to discuss design In other cases, particularly in the passive designs, vanability and failure modes for ALWR designs.

research serves to demonstrate acceptable design In addition, the panel toured a select group of the basis and severe accident response, for example, candidate facilities that had expenenced earth-for structural performance of the AP-600 con-quake damage and would be used in the EPRI/

tainment during passive containment cooling. In ARC data base. A geolog,st has been added as a i

still other cases, research is used to define consultant to the panel to venfy EPRI/ ARC acceptable standards for new features and ground motion estimates.

requirements.

2.2.3.3 ASME Section III 2.2.3 Research Accomplishments in FY 1994 The preliminary evaluation of the basis for revi-sions to ASME Section III design rules for piping 2.2.3.1 Rev.iew of Vendor Data published in the 1995 Wm' ter Addenda has been Under this program the NRC staff has completed completed. It was used to develop a staff position analysis of available vendor data. A listing of in the safety evaluation report of the AP-600.

design. maintenance, and application deficiencies, including historical weaknesses, which could be The NRC staff is interacting with ASME working useful in evaluating advanced light-water reactor group members to understand the bases for the (ALWR) applications has been completed.

published ASME design code changes to facilitate Research efforts have supported NRR (1) in NRC endorsement through 10 CFR Part 50.55a.

2-3 NUREG-1266

2.' Standard Reactor Designs 2.2.3.4 Age Dating of Geologic Materials The most limiting design basis accident evaluated is required by 10 CFR Part 100 and is derived Research activities on the age dating of geolog.ic from the 1962 report TID-14844, " Calculation of materials included the conducting of an inter-Distance Factors for Power and Test Reactor national workshop of 60 renowned geochronol-Sites," which postulates the release of the entire ogists and palcoseismologists. Field tests were core inventory of noble gases and 50% of the core conducted in California, Montana, and South inventory of iodine fission products from the core Carchna where rew and promising techniques into the containment. This " TID source term" is were applied.

used for evaluating the suitability of the reactor design as well as the site. Present regulatory 2.2.3.5 Modular Construction guidance, reflected in Regulatory Guides 1.3 and Preliminary licensing review criteria for structural 1.4, stipulates that this source term is instantan-modules have been developed. Using the de.

cously available for release from the containment veloped criteria, input was provided to NRR to the environment and specifies that the fission licensing staff for the draft safety evaluation Product iodine would primarily be in elemental report for the AP-600 structural modules, form. The TID-14844 source term, originally Information necessary to complete sections of the intended for site evaluation purposes, has also review criteria on design and testing of structural been applied to many aspects of plant design.

modules has been identified in the United States These include requirements for fission product and Japan. The contractor also participated in a cleanup systems, such as sprays and filters, allowable containment leak rate, control room design and calculation review on structural modules for the AP-600 with Westinghouse and habitability, equipment qualification, and others.

Bechtd This TID source term is associated with a severe reactor accident since only major core degrada-The distinguishing construction features, tion and meltmg could result in the release of operational modes, and the applicable building, such large quantities of fission products. How-electrical, and mechanical codes and standards ever, the present source term formulation, which for the advanced reactors have been identified, with Tasks 1 and 2 completed. The applicability of dates from 1962, while providm, g a high level of the codes and standards is ongoing for the CE protection for plant systems, does not reflect recent research findm, gs. As a result, a rigid System 80+, the AP-600, and the advanced and application of the TID source term to the eval-simplified BWRs.

uations of new plant designs may not provide the 2.2.3.6 Equipment Anchorage best engineering solutions on some aspects of future plant designs.

Phase I of a three-phase test program of anchorages used to anchor mechanical and 23.2 Program Strategy component supports to concrete was completed.

Preliminary review of results indicate that for On Ma,y 25,1988, the staff presented to the some anchor types, mainly expansion anchors, Commission an " Integration Plan for Closure of dynamic loads reduce performance.

Severe Accident Issues," SECY-88-147. This plan discussed major elements relating to closure of severe accident issues, including severe accident 2.3 Regulatory Application of New research efforts and related activities in siting, Swrce Terms emergensplanning, and potential changes to existing fcgulations as a result of severe accident 2.3.1 Statement of Problem research findings. On October 4,1990, the staff presented to the Commission a " Staff Study on Potential accidents are evaluated during reactor Source Term Update and Decoupling Siting from licensing as part of the Commission's defense-in-Design," SECY-90-341. This plan discussed the depth policy. Certain accidents, referred to as staff proposal to decouple the issue of reactor

" design basis accidents," are postulated to occur siting from the issue of source term composition and their consequences must be shown to be and dose calculations and to prepare an update i

acceptable.

and revision of the source term given in NUREG-1266 2-4 l

2. Standard Reactor Designs TID-14844. On April 10,1992, the staff presented 2.

NUREG/CR-5747, " Estimate of Radio-to the Commission its " Revised Accident Source nuclide Release Characteristics into Terms for Light-Water Nuclear Power Plants,"

Containment Under Severe Accident SECY-92-127. This paper provided a draft of a Conditions," dated November 1993.

revised source term for use in calculating offsite consequences of design base accidents. On 2.3.3.1 Update of Siting Regulations June 12,1992, the staff presented to the In FY 1994, staff efforts continued on updating Commission its proposed " Revision to 10 CFR 10 CFR Part 100, " Reactor Site Criteria." A Part 100, Revisions to 10 CFR Part 50, New proposed rule to revise Part 100 was first issued Appendix B to 10 CFR Part 100 and New for comment in October 1992. Source term and Appendix S to 10 CFR Part 50," SECY-92-215.

dose calculations were proposed to be eliminated This paper presented proposed rules to revise the for reactor siting by simply specifying a minimum reactor siting criteria. On July 28,1992, the exclusion area distance around the reactor site Commission announced the availability for and by stating population density criteria in the comment (57 FR 33374) of a draft report on rule. An update of the seismic site evaluation

" Accident Source Terms for Light-Water Nuclear criteria proposed to incorporate probabilistic as Power Plants," NUREG-1465. On September 28, well as deterministic methods. Extensive com-1992, the Commission issued an advance notice of ments, both domestic and foreign, were received proposed rulemaking on " Acceptability of Plant favoring the continued use of source term and Performance for Severe Accidents; Scope of dose calculations for reactor siting. In July 1994, Consideration in Safety Regulations"(57 FR the staff recommended that the proposed rule be 44513). Revised accident source terms, as well as withdrawn and that a revised proposed rule be other severe accident considerations, are to be issued that would incorporate basic reactor site incorporated into this rulemaking effort.

criteria and would continue the use of source term and dose calculations for the siting of custom 2.33 Research Accomplishments in FY 1994 plants. Afodifications to standardized plant designs to compensate for poor site character-The Commission's reacter site criteria (10 CFR istics would be discouraged, however. The Com-Part 100) require, irrespective of likelihood, that mission approved the staff's recommendation in an accidental fission product release from the September 1994 to issue a revised proposed rule.

core into containment be assumed to occur and that its offsite radiological consequences be 2.3.3.2 Emergency Planning Regulations evaluated. The criteria for defining the character-In FY 1994, staff efforts continued on emergency istics of that release into containment is derived planning licensing requirements for independent from the 1962 report, TID-14844.

spent fuel storage facilities and monitored retrievable storage facilities. He public comment Since 1962, a ;etter understanding of the timing period for the proposed rule expired in November 5

and nature of the fission product release has been 1993. The staff analyzed comments received and is obtained. As a result, a number of areas of developing the final regulations. In FY 1994, a regulatory activities have been identified that may notice of receipt of petition for rulemaking was benefit from changes introduced as a result of published in the Fedeml Regis;'er (59 FR 17499) source term and severe accident research. A requesting public comment on a petition revised accident source term document-submitted by VEPCO (PRhi-50-60) relating to NUREG-1465, " Accident Source Terms for NRC audits of emergency plans. The public Light-Water Nuclear Power Plants"-which is comment period ended June 1994, and the staff is intended to replace TID-14844, was issued. In currently analyzing comments received. A petition support of this effort, the following documents for rulemaking was also received from VEPCO in were issued:

December 1992 (PRhi-50-58) relating to emer-gency planning exercises. The petition was 1.

NUREG/CR-5901, "A Simplified hiodel of published in the Fedeml Register for public Aerosol Scrubbing by a Water Pool Overlying comment in Af arch 1993 (41 FR 12341). The staff Core Debris Interacting With Concrete,"

is currently evaluating public comments and dated November 1993.

developing a proposed resolution to the petition.

2-5 NUREG-1266

. _ - =. -

2. Standard Reactor Designs In March 1994, a final rulemaking was published updates and clarifies ambiguities that surfaced in in the Federal Register (59 FR 14087) to provide a the implementation of the Commission's revised emergency planning regulation that emergency planning exercise requirements.

l l

1 i

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i NUREG-1266 2-6

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t PART 2--NUCLEAR SAFETY RESEARCII--REACTOR REGULATION SUPPORT 4

3. PLANT PERFORMANCE 3.1 Statement of Problem reactor models interact in a tightly coupled manner at each time step.

A wide range of reactor plant design variations exists in the United States, and the safety of these Our reliance on the computer codes to provide plants must be ensured over a wide range of predictions of reactor response with acceptable normal and abnormal operating conditions. The uncertainties depends on three levels of NRC is required to independently assess experiments and comparisons of experimental licensees' safety analyses and performance in results with code predictions. First are basic designing, constructing, and operating a reactor experiments used to derive empirical formulas for with respect to the safety of the public for the determining basic phenomenological behavior complete spectrum of credible operating within each cell. Second are separate-effect conditions and events.

experiments used to test the code's predictions for a single, complex component such as a steam NRC's task is difficult because straightforward generator. Third are integral-system tests that are testing of all transients in all plant design varia-used to evaluate the code predictions of the tions would not be technically and economically performance of complete reactor cooling systems.

feasible. On the other hand, straightforward and The results of these tests provide feedback to exact theoretical analyses of a reactor's thermal-correct and improve the code and to improve our hydraulic behavior is not possible because mass, understanding of operating transients. Such energy, and momentum exchanges take place over understanding provides the basis for improving P ant operations to reduce the likelihood of l

complicated interfaces between reactor com-ponents, water, and steam and because of the accidents.

moving mechanical interfaces in pumps and the extensive baffle arrangements of steam generators 3.3 Research Accomplishments in FY m the pnmary loops.

g As a result, the NRC must use available experi-3.3.1 Reactor Safety Experiments mental data to validate analytical models for evaluating (1) design basis accidents, (2) the safety 3.3.1.1 High Burnup Fuel Behavior implications of actual events in operating reactors, and (3) hypothetical transient scenarios deter.

By the early 1990s, it,had become clear that mined to be major contributors to risk as a result burnups sn commercial power reactors were of probabilistic risk assessment studies.

exceeding the burnup range that had been used to validate NRC's fuel behavior computer codes and related fuel damage criteria. Fuel suppliers were 3.2 Program Strategy providing data to support the licensing of higher-burnup fuel designs, but the NRC's capability to A dual analytical and experimental approach is independently validate such data had not been used to achieve a firm technical understanding of updated. Accordingly, it was decided to assess the the thermal-hydraulic behavior of the reactor.The need for (1) fuel performance model changes (e.g.,

UO thermal conductivity, fission gas release),

NRC starts by simulating the actual reactor's 2

continuous flow of heat and fluids with a com-(2) fuel performance code updates (i.e., the puterized model consisting of many discrete cells FRAPCON code and resultant effects on fuel exchanging mass, energy, and momentum at each stored energy), and (3) changes in the threshold small but finite, time step. Physicallaws are used criteria for fuel failure under reactivity transients, when possible to calculate all these exchanges.

Empirically derived formulas that are obtained Contracts at three laboratories were placed to from experiments are used as necessary to respond to this need. One is focusing on i

account for such complex effects as friction phenomenological models, one on the modifi-between vapor and liquid. The calculations are cation of computer codes, and the third on plant made for each time step and for each cell. The transient calculations to estimate the impact on 3-1 NUREG-1266

3. Plant Performance reactor safety. During the year, the NRC became 3.3.2 Safety Code Development and aware of new test results on high-burnup fuel Maintenance being obtained in France, Japan, and Russia It is generally not possible to assess the safety suggesting the criteria used to predict fuel failure performance of reactor and plant systems with may need to be revised. Sm, ee no such testmg is tests in full-scale facilities, and an understanding bemg performed in the United States, efforts were of the thermal-hydraulic behavior of these plants made to enter into cooperative arrangements with must be established with the use of computer foreign laboratories to obtam these data. Invita-codes. Most of the NRC's independent analyses tions were extended to these laboratones to for the AP-600 and SBWR will be done with the present prehminary mformation at the,NRC's RELAP code, which is being upgraded for annual Water Reactor Safety Information application to these desi ns. Before anY of the 3

8 Meetmg, and such presentations were made on 2

NRC codes are used for th.is purpose, or released October 26,1994. More definitive results will be f r use by others, they undergo developmental available in 1995. These test results will be used to ssessment and peer review. Revised documen-assess, and perhaps modify, fuel damage criteria tation is also provided for these improved codes.

used by the NRC in licensing.

The upgraded version of RELAP for use on the new passive plant designs will be released in early 3.3.1.2 Thermal-Ilydraulic Phenomena 1995.

Experiments are being performed at the As part of the code maintenance activities for University of Maryland in a scaled (1:4 in height, RELAP and the TRAC code (both PWR and with a 1:500 volume) experimental facility that BWR versions), the NRC conducts an inter-simulates a Babcock and Wilcox reactor. This national Code Applications and Maintenance facility was originally constructed under NRC Program (CAMP). There are now 17 member contract to study small-break loss-of-coolant countries in CAMP, each of which participates in accidents. Following successful completion of that semi-annual meetings and makes cash contribu-program, the facility's mission was shifted to the tions to supplement the NRC code development current study of mixing phenomena associated and assessment programs. Members also provide with boron injection events. Recent studies have code assessment studies, recommend code suggested that dilution of boron (used to control improvements, and make other technical reactivity) may occur that could potentially result contributions to assist in the development and in reactivity transients.

assessment of the codes.

I NUREG-1266 3-2

t

4. HUMAN RELIABILITY 4.1 Statement of Problem 4.3 Research Accomplishments in FY 1994 About half of all safety-related events reported at nuclear power plants and among nuclear materials licensees continue to involve human 4.3.1 Personnel Performance performance. Methods and data are needed to identify, systematically set priorities for, and Work concluded on the development of a method suggest solutions to human performance issues in to assess the effectiveness of training programs at nuclear operations during normal, transient, and nuclear power plants. Measures and supporting emergency situations and during plant documentation for a training effectiveness eval-maintenance.

untion method will be published in a two-volume technical report. Data analyses from a project on the factors that are considered when making 4.2 Program Strategy decisions on operations staffing and on how staffing relates to safe startup, shutdown, and The human reliability research program has three operation of nuclear power plants are now com-objectives:(1) to broaden NRC's understanding of plete. Results of these analyses have been incor-human performance and to identify causes of porated into NUREG/CR-6122, " Staffing human error; (2) to accurately measure human Decision Processes and Issues." A second product performance for enhancing safer operations and of this study was published as NUREG/CR-6123, precluding critical errors; and (3) to develop the "An International Comparison of Commercial technical basis for regulatory requirements, Nuclear Power Plant Staffing Regulations and recommendations, and guidance related to human Practice: 1980-1990." Work continues on a study performance.

to establish a technical basis for minimum shift staffing for both control room crews and for The human reliability research program is divided operational support staff outside the control room into three interrelated program elements: (1) per.

at nuclear power plants based on workload and task a!!ocation. A handbook on the effects of sonnel performance, (2) human-system interfaces, and (3) reliability assessment. The purpose of the environmental factors on human performance for personnel performance element is to develop use by nuclear power plant m, spectors was enhanced methods for collecting and managing published as Volume 1 of NUREG/CR-5680, personnel performance data and to improve "The Impact of Environmental Conditions on Human Performance." A critical review of the understanding of the effects of personnel per.

formance on the safety of nuclear operations and literature was published as Volume 2 of NUREG/

maintenance. In addition, personnel performance CR-5680. 'Iko reports concerning training for research will broaden the understanding of such responding to accidents were published as NUREG/CR-6126 and NUREG/CR-6127. These factors as staffing, qualifications, and training that influence human performance in the nuclear reports describe decisionmaking and stress coping system and will develop information necessary to skills that may be needed to respond to an reduce the negative impact of these influences on accident situation, as well as potential training nuclear safety. Research in the human-system approaches for developing those skills.

interface element will provide the technical basis for guidelines and criteria to evaluate the interface Research was initiated on communication errors between the system and the human user from the in nuclear power plant events to characterize the perspective of safe operations and maintenance.

root cause(s) of these errors, identify potential Lastly, the reliability assessment element includes corrective actions for each category of communi-multidisciplinary research that will integrate cation error, and develop proposed review criteria human and hardware considerations for evalu-and guidelines. A study on whether links exist ating reliability and risk in NRC licensing, between operator effectiveness and the simulator inspection, and regulatory decisions.

training received by operators at multi-unit 4-1 NUREG-1266

4. Human Reliability stations as compared to simulator training at After internal NRC and independent peer review, single-unit stations was initiated.

proposed guidelines in support of the standard review plan for the review of advanced control l

room designs were published as NUREG/CR-t 43.2 Human-System Interfaces 5908, " Advanced Human-System Interface Design Review Guidelines." These guidelines were built Human-system interface research includes NRC on previously validated guidelines available from l

participation in the Organization for Economic other industries, including the aerospace and Cooperation and Development Halden reactor defense industries, and were prepared in both project, which is a multifaceted program that paper form and computerized (interactive) form.

meludes verification and validation of digital r

systems, man-machine interaction, and sur-Work was completed as part of resolving Human veillance and support systems for advanced Factors Generic Issues 5.1, " Local Control control rooms. Information was developed on Stations," and 5.2, " Review Criteria for Human (1) methods and tools for the development and Factors Aspects of Advanced Controls and verification and validation of safety-related Instrumentation," which resulted in the publica-i software, and (2) experience with development tion of NUREG/CR-6146, " Local Control Sta-and quality assurance of software systems at the tions: Human Engineering Issues and Insights,"

Halden project.

and NUREG/CR-6105, " Human Factors Engineering Guidance for the Review of Research continued to evaluate the positive and Advanced Alarm Systems."

negative attributes of standards and computer-aided-software engineering tools for use in the The review guidance published in the three docu-f certification of high-integrity software for nuclear ments described above is presently being inte-power plant safety systems. Research will be grated with additional material to form Revision 1 completed in 1995 on a project co-sponsored by to NUREG-0700, " Human System Interface i

the Electric Power Research Institute on verifi-Design Review Guideline," a draft of which will cation and validation guidelines and quality be issued for public comment in FY 1995.

metrics for digital high-integrity systems.

Work was begun on a new project to develop A project was initiated to independently evaluate, guidance for the review of advanced digital alarm test, and improve upon verification and validation systems.

l guidelines for use in the audit of computer-based Following recommendations from the Advisory safety systems.

Committee on Reactor Safeguards and from the l

Commission, the staff initiated the first phase of a The NRC and the National Institute of Standards project with the National Academies of Sciences I

and Technology issued the proceedings of a jointly and Engineering to conduct a study and workshop sponsored Digital Systems Reliability and Nuclear on a coherent and effective approach to the Safety Workshop (NUREG/CP-0136). The goals regulation of computer-based (digital) systems in I

achieved by the workshop were (1) providmg nuclear safety and control systems. The results of feedback to the NRC from outside experts regard-the full study and workshop will give advice to the l

ing potential safety issues, proposed regulatory NRC on the framework for a coherent and effec-I positions, and research associated with applica-tive regulatory program.

tion of digital systems m, nuclear power plants, and (2) continuing the indepth exposure of the i

NRC staff to digital systems design issues related 433 Reliability Assessment j

to nuclear safety by discussions with experts in NUREG/CR-4639 revisions were issued to com-the state of the art and practice of digital systems.

plete research on collecting, cataloguing, and storing, in a computerized library, estimates of Following the workshop, research was initiated to probabilities of operator error and hardware j

ensure the completeness of the technical bases for failure. Research continued to develop alternative 1

regulatory requirements intended to ensure the quantification methods for incorporating the integrity of safety-related software.

influence of organizational factors into j

NUREG-1266 4-2 1

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4. Human Reliability probabilistic risk assessment (PRA). NbREG/

to accommodate " advanced" instrumentation.

CR-6208, describing the results of data collected Initial studies have been documented in NUREG/

at nuclear power plant simulators to reduce CR-5094, " Functional Issues and Environmental uncertainties associated with operator per-Qualification of Digital Protection Systems of formance in cognitively demanding simulated Advanced Light-Water Nuclear Reactors," where emergencies, was published.

the likely effects of environmental stressors on safety system components and interfaces are Research continued to analyze information from examined. A methodology for identifying the need the simulator portion of the NRC-administered for accelerated aging in qualifying new 1&C operator requalification ex,aminations. Estimates systems for placement in benign environments from th,s source may provide valid error rates for i

was also proposed. Current research involves an use m a nuclear power plant PRA.

experimental investigation into the functional behavior and failure modes that result for a For several years the NRC has been developing reliability and risk analysis tools to evaluate the mier Processor-based safety system under the risk impact of changes of selected requirements in application of environmental stressors, such as the technical specifications. De methods were presence of smoke, electromagnet,c interference, i

completed, and both detailed technical reports radio-frequency interference, temperature, and and a handbook to guide NRC reviewers through humidity. The prototypic safety channel imple-mented for th,s study employs technologies the use of these methods will be issued in FY i

1995 as NUREG/CR-6141.

representative of those proposed for use m, advanced light-water reactors. Environmental The Oak Ridge National Laboratory (ORNL)is tests should reveal any potential system vul-conducting a study to identify functional and nerabilities and help determine the expected effect operating environmental issues arising from the of a stressor on advanced I&C system com-application of new technologies in instrumentation ponents. This information supports a clearer and control (l&C) systems for both current and definition of what (and to what level) stressor next-generation nuclear power plants. The goal of equipment should be qualified to withstand and this program is to establish the technical basis for thus builds the technical basis for supplementing augmenting the equipment qualification process current qualification guidelines.

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4-3 NUREG-1266

5. REACTOR ACCIDENT ANALYSIS 5.1 Reactor Risk Analysis other precursor events, an extensive two-phased project was initiated in 1989 to examine the 5.1.1 Statement of Problem Potential risks of accidents initiated during low-power and shutdown modes of operation. Phase 1, Probabilistic risk analysis (PRA) has been shown completed at the end of 1991, was a coarse screen-to be a systematic and comprehensive method for ing analysis of all operational modes (other than identifying and evaluating the effectiveness of full power) for one BWR and one PWR to provide safety improvements proposed to reduce the timely support for the Office of Nuclear Reactor likelihood and consequences of nuclear power Regulation's (NRR) regulatory analysis and to plant accidents. PRA is used by the NRC staff in guide the Phase 2 effort. A significant finding was a number of ways, including for evaluating the that the traditional concept of technical specifi-level of safety at selected operating plants; for cation modes of operation does not adequately assessing the margins of safety in current require-delineate plant operating boundary conditions ments in light of the Commission's safety goal needed for risk analyses. The Phase 2 effort con-policy; for monitoring plant performance; and for centrated on a specific operating state for each of identifying potential improvements in equipment the two plants, selecting the potentially highest or operator reliability.

risk operating state, based on the Phase 1 results.

In addition, a simplified analysis of potential in-P ant and offsite accident progression and I

l 5.1.2 Program Strategy health consequences of such accidents has been The reactor accident risk analysis research effort Performed and provided to NRR in support of is applied in four ways: (1) providing expert their regulatory activities as documented in review of severe accident PRAs to assess, for NUREG-1449. The complete results of Phase 2 example, the risk implications of accident are being published as NUREG/CR-6143 for the management strategies in order to minimize the BWR and NUREG/CR-6144 for the PWR.

release of radioactive material to the environment fluman Reliability Analysis. As part of an NRC-during severe reactor accidents; (2) developm, g, sponsored program evolving from an assessment venfymg, demonstratmg, and ma, taimng of human reliability issues in low-power and m

methods for analyzing the consequenees of m.

shutdown operations in nuclear power plants, an plant and offsite s processes for use m,cVere accident physical improved approach to human reliability analysis nsk assessment and (HRA)is currently being developed. It is intended developmg and demonstratmg methods for to be fully integrated with PRA methodology and c uantifymg the uncertainty m nsk estimates and to enable a better assessment of the human tie relative contributions of specific issue un.

contribution to plant risk, both during low-power, certainty to the overall uncertainty; (3) reassessmg shutdown and at-power operations.

penodically the frequencies, consequences, and risk of severe accidents in nuclear power plants In FYs 1992 and 1993, a Human Action Classifi-and performing peer review of methods used and cation Scheme for categorizing human actions and results obtained; and (4) developing risk-based associated influences in actual low-power and management tools capable of determining the shutdown events was developed and implemented.

incremental risk reduction associated with These accomplishments were documented in proposed plant design and operational modifi-NUREG/CR-6093,"An Analysis of Operational cations and assisting in the setting of priorities for Experience During Low Power and Shutdown and efforts in licensing and research activities.

a Plan for Addressing Human Reliability Assess-ment Issues."

5.1.3 Research Accomplishments in FY 1994 During FYs 1993 and 1994, work continued on 5.1.3.1 Issue-Oriented Projects (1) the development of a multidisciplinary frame-work for mtegratmg HRA with PRA; (2) the Analpis oflow-Power and Shutdown Accident characterization of errors of commission (EOCs)

Risks. As a result of the Chernobyl accident and and human dependencies, including general 5-1 NUREG-1266

5. Reactor Accident Analysis 1

guidance for their identification and represen-per reactor-year, which is found to be within the tation in PRAs: and (3) the recognition of data range of core damage frequency estimates pro-base improvement needs, including a better vided for similar Westinghouse PWR facilities.

characterization of human actions and their The licensee has subsequently requested modifi-associated performance context (e.g., plant cations to its plant technical specifications based, conditions, performance shaping factors, and in part, on its risk analysis. The RES staff pro-dependencies), as well as a better description of vided a draft safety evaluation report to NRR, an event timeline. These accomplishments are which became a part of the basis for their currently being documented.

regulatory decision.

This framework provided the capability to identify 5.1.3.2 Methods Development Projects factors that influence humans to perform unsafe SAPIIIRE Compurce Tools. The set of computer actions and thereby created a systematic basis for codes called SAPHIRE (System Analysis Pro-evaluating the significance and characteristics of grams for Hands-on Integrated Reliability EOCs and dependency from operational events.

Evaluation) has been updated to version 5.0. This Thus, the framework has enabled important set of codes is to be used in performing proba-aspects of EOCs and dependency to be con-bilistic risk analyses and permit an analyst to sidered in the development of an improved HRA perform many of the functions necessary to methodology and has clarified the regmrements create, quantify, and evaluate the accident risks of for their more realistic inclusion in PRA models.

nuclear power plants. The codes were used By the framework's provision of a single language extensively to perform the low-power and shut-and common structure for relating the different down risk analyses previously described and are dimensions of human-system mteractions, the currently being used for analysis and resolution of evaluations of EOCs and dependencies has been generic safety issues and for evaluating the safety demonstrated to be both tractable and tenable.

aspects of concept plant designs. During 1994, Cons,dering the importance of these issues in PRA data from four more licensed nuclear power i

nuc! ear power plant safety, this change is an plants were added to the SAPHIRE data base important advance. These EOC and dependency and most of the data from previous plant loads capabilities will be refine,d and expanded upon m were updated to version 5.0. Th brings the data subsequent tasks pertammg to the development base total to 17 plants, two of v" h are advanced phase.

concept plants added to suppon she agency's design certification reviews. Courses continued to l

The primary product of the current workscope be provided to both the NRC staff and con-(FYs 1994 and 1995) will be a workable HRA tractors using these codes. The documentation for quantification process that includes the folk) wing:

version 5.0 has been published as NUREG/CR-how to identify and incorporate human failure 6116, and the new codes and user manuals have events in the logic models used in PRAs, what been sent to the Energy Science and Technology information is required for probabilities to be Software Center at ORNL and made available for assigned to these failure events, how this infor-U.S. distribution. The previous version, SAPHIRE mation is used to estimate the probabilities, and 4.17, has been made available to foreign countries how the probabilities are incorporated into the that do not have cooperative agreements with the PRA quaatification process. It is anticipated that NRC.

the final phase (FYs 1995 and 1996) of the project will demonstrate the usefulness and acceptability Consequence Code Benchmark. The NRC has of the developed methodology's implementation successfully completed the work with the Com-guidelines using selected parts of the low power mission of the European Communities (CEC) and and shutdown Level 1 PRAs.

the Organization for Economic Cooperation and Development (OECD) to carry out an inter-South Texas Risk Analysis. In 1992, the staff comparison exercise on probabilistic accident

]

completed a review of the South Texas Project risk consequence codes. The six codes being evaluated analysis and documented the results and findings were MACCS (United States), COSYMA (CEC),

(NUREG/CR-5606). The licensee estimated the CONDOR (UK), OSCAAR (Japan), LENA overall mean core damage frequency to be 2 E-4 (Finland), and ARANO (Sweden). The NUREG-1266 5-2

5. Reactor Accident Analysis intercomparison exercise used six radioactive To identify improvements in PRA techniques e

accident source terms and calculated dose and associated data necessary for each consequences for such measures as whole body category of staff use.

dose and fatal cancers. The results of these comparisons were published in FY 1994 by the The group published NUREG-1489 in March CEC and OECD and demonstrated that the 1994, which included initial guidance to the staff reactor accident consequence predictions cal.

on the use of PRA in screening and analyzing culated by these new codes all agreed within a reactor operational events and on basic terms and reasonable range.

methods used in PRA. The report also contains a number of recommendations for additional guid-Offsite Consequence Uncertainty Analysis. The ance development, improvements to the NRC's NRC has completed a pilot probabilistic con-PRA trammg program, and improvements m PRA sequence uncertainty analysis in cooperation with tools and data bases used by the staff. A draft the CEC. Sixteen internationally renowned Commission pohey statem,ent on the use of PRA atmospheric dispersion and deposition experts s well as an implementation plan has been

- participated in a NUREG-1150-based formal developed to guide staff response to the recommendations.

expert judgment clicitation and evaluation process in which the information needed for the pilot uncertainty study was elicited from the experts.

5.2 Containment Performance The individual expert assessments were aggre-gated to form probabilistic distributions for dry 5.2.1 Statement of Problem deposition velocity, wet deposition parameters, and Gaussian plume dispersion parameters, Core-melt accidents that proceed to vessel failure have the potential to produce high pressures and respectively. The CEC will use the methods.

formulated jomtly dun,ng the pilot study, with the temperatures that might challenge containment NRC providing limited key techm, cal support t integrity. It is known from previous risk studies, i

the CEC, m obtaimng other relevant information and from the experiences at Chernobyl and Three Mile Island, that containment survival or even to complete a probabilistic consequence.

delayed failure has an all-important effect on uncertainty study. The final data base will b shared by the two orgamzations for perform?

minimizing the release of radioactivity to the mg environment in the event of a core-melt accident.

i probabilistic consequence uncertainty analyses.

If realistic assessments of the consequences of core-melt accidents, which so strongly depend on 5.1.3.3 Risk-Related Training and Guidance whether or when containments might fail in the Development course of the accident, are to be made, then an Guidancefor Staff Use of Risk Analysis. In a July understanding of the phenomena that occur in 1991 letter, the NRC's Advisory Committee on containment in the latter stages of the, accident that could lead to contamment failure is essential.

Reactor Safeguards identified a number of concerns with the staff's uses of risk analysis. In response, the NRC's Executive Director for 5.2.2 Program Strategy Operations formed a working group of staff NRC's research efforts in this program element L

management to " consider what unprovements in focus directly on the phenomena believed to be methods and data analysis are possible and most likely to produce high pressures and needed, the role of uncertainty analysis in temperatures that might challenge the contain-different staff uses of PSA,.. " This workm, g ment integrity: high-pressure ejection from the group was organized m early 1992 with the reactor vessel of finely divided particles of molten followm, g object,ves:

core debris; generation of noncondensable and i

flammable gases from the decomposition of o

'Ib develop guidance on consistent and concrete by hot cofe debris; the direct thermal appropriate uses of PRA within the NRC; and chemical attack by molten core debris of structures and engineered safety features; and the

'Ib identify skills and experience necessary for burning or detonation of hydrogen and other e

each category of staff use; and gases produced in the course of the accident.

5-3 NUREG-1266

5. Reactor Accident Analysis NRC's research program dealing with reactor behavior. With these kinds of data, the NRC is containment safety consists of five areas of better able to confirm the adequacy of its require-research. These five research activities address:

ments for the design and reliability of the systems (1) the interaction of molten core debris with that may be used for mitigating the consequences structural concrete, including the ablation of of severe accidents.

concrete structures, heat transfer to structures in the containment and to overlying water, the 5.2.3.1 High-Pressure Melt Ejection-Direct generation of noncondensable and flammable Containment Heating gases, and fission products containing aerosols; (2) direct containment heating by molten debris in certam reactor accidents, degradation of the particles ejected from the vessel at high pressure reactor core can take place while the reactor and hydrogen production resulting from steam coolant system remams pressurized. A molten core left uncooled will dram and relocate to the oxidation of the metallic component of that bottom of the reactor vessel. If the reactor vessel debris;(3) the combustion of hydrogen in the containment, including the potential for detona.

fails, the core melt will be ejected into the con-tion; (4) the development, validation, maintenance, tainment cavity under pressure. If the matenal and application of various computer codes that subseguently should be ejected from the reactor are capable of describing the multiple phenomena cavity mto the surroundmg contamment volumes that occur in severe accident sequences of in the, form of fine particles, thermal energy can interest; and (5) assessment of severe accident be quickly transferred to the containment phenomena, including containment performance, atmosphere pressunzmg it. The metallic com-for advanced reactors (see Section 2.1).

Ponents of the ejected core debris can further oxidize m air or m steam and can gencrate a large quantity of hydrogen and chemical energy that The overall goals of the research are to develop would further pressurize the containment.The techmcal bases for assessmg contamment projected process is called direct containment performance caer the range of risk-sigmficant heating (DCH).

core-melt events, to develop an improved under-standing of the range of phenomena expected As part of the DCH issue resolution plan for during severe reactor accidents, and to develop PWRs, a study was completed for the Zion (Ill.)

improved methods for assessmg fission product plant and documented in NUREG/CR-6075 and behavior. With these kinds of data, the NRC is 6075 Supplement 1,"The Probability of Contain-better able to confirm the adequacy of its ment Failure by Direct Containment Heating in requirements for the design and reliability of the Zion." Both reports have been peer reviewed and systems that may be used for mitigatmg the will be published in December 1994.

effects of severe acc, dents.

i The culmination of extensive experimental and analytical research undertaken principally for the 5.2.3 Research Accomplishments in FY 1994 Zion reactor has produced the finding that DCH loads are significantly lower than once estimated In order to ensure that existing regulations and consequently pose no tangible threat to the adequately protect the public from the conse-containment during a severe accident. This is due quences of severe accidents, the NRC conducts primarily to the inherent design characteristics of research m several areas, among them source many U.S. reactors. Future efforts will explore our term release and transport, core-melt progression, ability to extrapolate these findings to the fuel-coolant interactions, direct containment spectrum of reactor designs.

heating, hydrogen combustion, and melt-concrete interactions. The overall goals of the research are 5.2.3.2 Hydrogen Combustion to develop (1) techmcal bases for assessing con-tainment performance over the range of risk-Significant information exists on hydrogen com-significant core-melt events, (2) an improved bustion to assess the possible threat to contain-understanding of the range of phenomena ex-ment and safety-related equipment. Some pected during severe reactor accidents, and (3) ancillary issues remain related to a better under-improved methods for assessing fission product standing of the likelihood of various modes of NUREG-1266 5-4

5. Reactor Accident Analysis combustion at high temperature and in the sufficient steam had been removed by conden-presence oflarge quantities of steam.

sation from the water sprays. No detonations or accelerated flame propagation were observed in The largest current NRC program in this topical these tests. The combustion mode was character-area comes out of a joint agreement between the ized by multiple deflagrations with relatively small NRC and the Ministry of International Trade and pressure rises. The thermal glow plugs were effec-Industry (MITI) of Japan, managed by the Nu.

tive in burning the hydrogen safely by igniting the clear Power Engineering Corporation (NUPEC).

gases as the mixtures became marginally Under the agreement, a high-temperature flammable.

hydrogen combustion program related to high-speed combustion modes, i.e., detonat, ion and 5.2.3.3 Melt-Concrete Interactions and Debris deflagration to detonation transition, is under way at the Brookhaven National Laboratory. A CoolabilitY small-scale developmental apparatus was con-structed and has provided a preliminary set of In those severe accident scenarios in which the experimental data and solutions to a number of reactor vessel fails, high-temperature core debris design and operational problems for a larger-scale may fall into the reactor cavity where it can high-temperature combustion facility (HTCF).

thermally and chemically interact with structural The construction of the HTCF has been com-c nerete. The major areas of concern associated pleted and high-temperature experiments begun.

with melt-concrete mteractions during a severe As a result of the cooperative agreement with accident are the penetration of the basemat and Japan, the NRC has access to the ongoing hydro-failure of the hner, the generation of radioactive gen research in Japan managed by NUPEC. This crosols and gases, meluding combustible gases, research provides a greatly expanded and nd the overpressurization of the contamment.

improved data base for the validation of analytical tools.

Early experiments on melt-concrete interactions were conducted without an overlying water pool.

A hydrogen research program is also under way More recent experiments on melt-concrete to investigate diffusion flame behavior in low-interactions, otherwise known as debris coolability speed hydrogen combustion. Experiments were experiments, were conducted in the presence of an performed in a small-scale facility to examine the overlying water pool. It has been postulated that influence of ignition source strength on the lean adding water to cover the core debris will effec-flammability limits of hydrogen-air mixtures at tively quench the molten corium and terminate temperatures of 300K and pressure of one bar.

melt-concrete interactions. The currently active The facility has been redesigned to eliminate experimental research on debris coolability, called diffusion flame interference with the walls.

the Melt Attack and Coolability Experiments Construction will be completed during FY 1995.

(MACE) program, was developed as an extension The results will be used to help resolve several of the Advanced Containment Experiments outstanding issues in severe accident behavior, (ACE) program under the sponsorship of NRC, such as high-temperature combustion phenomena EPRI, and other mostly governmental agencies in and detonation initiation by high-temperature several countries. The MACE program is steam-hydrogen-particle jets.

intended to determine the ability of water to cool prototypic ex-vessel core debris of urania-zirconia Experiments have been conducted to determine composition. Four tests, including a scoping test, hydrogen combustion behavior under conditions were conducted in the MACE program in FYs of rapidly condensing steam from water sprays.

1992 and 1993. The latest MACE test, M3, The experimental conditions were nearly proto-performed at a scale more than two times larger typical of those that would be expected in a severe than the previous tests, was conducted in accident in the ABB Combustion Engineering December 1994. This test was designed to provide System 80+ containment. The mixtures were information on the effect of scale on crust forma-initially nonflammable because of dilution by tion, stability, and debris coolability. Analysis of steam. The mixtures were ignited by thermal glow the test data is under way and will guide the plugs as the mixtures became flammable after planning of future activities.

5-5 NUREG-1266

5.- Reactor Accident Analysis 5.2.3.4 Severe Accident Codes the national laboratories and in other organiza-tions in the United States. MELCOR has been Because of the difficulty in performing prototypic applied to the analyses of various plant accident experiments for a variety of severe accident transients, and a large number of code assess-scenarios, substantial reliance must be placed on ments have been completed in FY 1994 by several the development, verification, and validation of United States and international user system-level computer codes for analyzing severe organizations.

accident phenomena. Several codes (hiELCOR, SCDAP/RELAP5, CONTAIN) have been SCDAP/REIAPS is a computer code that has the developed for various stages in severe accidents, capability to perform detailed analyses of the both in-vessel and ex-vessel, for both BWRs and in-vessel progression of LWR severe accidents as PWRs. Additional codes such as COMMIX, well as detailed experiment analyses. The code has VICTORIA, llMS, and IFCI are being developed been in world-wide use for several years. A and maintained to perform specific functions that SCDAP/RELAP5 peer review committee com-require more detailed modeling than the pleted an extensive review of the code in FY 1993 system-level codes.

and identified several areas of modeling and documentation that needed improvement. The improvements completed in FY 1994 include MELCOR is an integrated computer code that models the progression of severe accidents in (1) bnngmg code manuals up to date,(2) mak, g m

the code more reliable and user fnendly, light-water reactor (LWR) power plants. The code (3) streamh,nmg the code output, (4) establishm, g a can be used to evaluate the progression of severe closer link between the code and associated accidents from initiation thmugh containment failure and to estimate severe accident source documentation, and (5) makmg assessment terms as well as their sensitivities and uncertain-n ports available for each version of the released ties in a variety of applications. The NRC has code. Specific SCDAP/RELAI 5 activities emnpleted in FY 1994 include (1) release of the been supporting the MELCOR development and f the code (MOD 3.1) and a five-new msmn assessment program for a number of years.The focus of the development efforts in FY 1994 has Volume set of user and code manuals,(2) comple-tion of BWR control blade / channel box model been to improve capabilities to model the phenomena of in-vessel natural circulation, core testmg,(3) analyses of severe accident sequences structure melting and relocation, and ex-vessel for the Zion (Ill.) and Surry (Va.) plants to Supp rt resolut,qn of the DCII issue,(4) comple-4 i

core-concrete interactions, and to model the tion of nodalization studies for a station blackout performance of passive safety systems in advanced light-water reactors (ALWRs). These

".ccident scenario (high, pressure case) for the and a few other improvements were made in Surry pl nt,(5) completion of a TMI-2 accident response to a number of recommendations made sequence study usmg MOD 3.1, and (6) addition of debns ox,idation and cutectic mteraction models by an independent peer review group convened by the NRC. A significant effort was made in FY for fuel rod claddmg and PWR control rod,

1994 to incorporate CORCON-MOD 3, a stand-m terials. Ongoing work m, eludes (1) continuing to provide code ma, tenance,(2) performing time m

alone core concrete interaction code, into MELCOR. With this work now completed, the nd spatial nodalizatmn studies,(3) performmg

]

NRC has no further plan to support maintenance more assessment studies agamst expenmental of the stand-alone CORCON-MOD 3 code.

data, and (4) contmumg to improve high-pnon,ty j

modeling deficiency items as recommended by the The assessment of MELCOR by NRC contractors continued during this report period as did the CONTAIN is a detailed code for the integrated MELCOR Cooperative Assessment Program.The analysis of containment phenomena. The code goal of the latter work, initiated in FY 1992,is to provides the capability to predict the physical, i

create an international forum for information chemical, and radiological conditions inside a exchange on the applicability, limitations, and reactor containment in the event of a severe operational experiences with MELCOR. The goal accident. One issue currently under investigation of the former work is to broaden the MELCOR is DCli and pressurization of the reactor assessment data base through work conducted in containment atmosphere by molten core materials NUREG-1266 5-6

5. Reactor Accident Analysis ejected following the lower vessel head failure The experimental results will enhance the under pressure. Assessment of the DCH models chemistry models already in VICTORIA in in CONTAIN against experimental data con-simulating severe reactor accidents. Battelle and tinued in FY 1994. Another development is the NRC staff have modeled the retention of I

related to containment analyses for ALWR cesium on stainless steel in this chemical system designs. The industry is developing containment for future implementation in the VICTORIA designs for ALWRs that incorporate passive code.

cooling and decay heat removal features for protection against long-teim containment over-IIMS is a best-estimate, three-dimensional, pressure in severe accidents. The CONTAIN code transient code for analyzing the transport, mixing, 1

was modified in selected areas to model these and burning of hydrogen. The code can model passive A13VR safety features. Finally, a compre-geometrically complex structures with multiple hensive peer review of the code was completed in compartments and can simulate the effects of which code modeling and validation were assessed condensation, heat transfer to walls and internal for the code's intended applications.

structures, chemical kinetics, and fluid turbulence.

During FY 1994, an liMS user's manual was COMMIX is a three-dimensional transient single.

developed to provide the basic information for phase computer code for thermal-hydraulic setting up and running problems with the code.

analysis of single and multicomponent engineering Also, HMS was converted from a main frame systems. The code solves a system of time.

computer code to a workstation code.

j dependent and multidimensional conservation for mass, momentum, energy, and transport IFCI, an Integrated Fuel-Coolant Interactions equations. A number of phenomena encountered computer code, provides a numerical tool for in postulated severe accidents in A1AVRs are calculating and predicting the consequences of inherently multidimensional in nature. The fuel-coolant interactions, including the breakup of COMMIX code is being developed to address melt streams, the expansion work, and the

)

issues such as natural circulation, flow stratifi.

dynamic pressure loads on in-vessel and ex-vessel cation, and the effect of noncondensable gas

.auures. Models in the IFCI code are presently distribution on local condensation and being validated against experimental data. During evaporation for the AP-600 plant.

FY 1994, operational assessment of the IFCI code was completed and a user's manual was published VICTORIA is a computer code designed to (NUREG/CR-6211).

analyze fission product behavior within the reactor coolant system (RCS) during a severe 5.3 Severe Accident Phenomenology accident. The code provides detailed predictions of the fission product release from the fuel and 5.3.1 Statement of Problem the transport in the RCS of radionuclides and nonradioactive materials during core degradation.

Major uncertainties in estimating the probability During FY 1994, assessment and validation of of early containment failure, and the associated models used in the VICTORIA computer code radioactive release, in the event of a severe acci-against existing data bases and against new data dent appear to be significantly related to from several experimental test facilities (e.g.,

uncertainties in the in-vessel progression of the FALCON VI ST)were carried out. An improved accident while the fuel material remains in the fission product chemistry model was implemented reactor pressure vessel. Better understanding is in VICTORIA. In FY 1994, the code was used for being gained of the entire sequence of severe a full plant station blackout analysis for the Surry accident phenomena, including core-melt nuclear power plant.

progression, fission product release, fuel-coolant interactions, hydrogen generation, and response of Battelle Columbus Laboratory has performed a the RCS to fuel melting and relocation. Contain-large number of experiments on boric acid and its ment failure probabilities and related source interaction with a variety of chemical species that terms can now be estimated with less conserva-are expected in the RCS under severe accident tism than in previous analysis to ensure adequate conditions (e.g., cesium hydroxide, cesium iodide).

margin.

l 5-7 NUREG-1266

5. Reactor Accident Analysis 53.2 Program Strategy completed, and the final NUREG-1465 report will be issued in January 1995.

In order to better understand just what happens during a core-melt accident, and thereby reduce The NRC has entered into an agreement with the the uncertainties in both accident behavior and Commissariat a L'Energie Atomique (CEA) of the potential release of radioactivity, the NRC is France to participate in the PHEBUS-FP (fission pursuing a program of research addressing (1) the product) program. The program is sponsored by heatup and meltdown of the core;(2) hydrogen the CEA and the Commission of the European generation;(3) fission product release, transport, Communities and is aimed at studying, under j

and deposition within the RCS;(4) energetic sufficiently prototypic conditions in an in-pile l

fuel-coolant interactions (FCIs) that may occur as facility, those phenomena governing the transport, molten debris falls into the water-filled lower head retention, and chemistry of fission products under or as water is added to molten debris; (5) the severe accident conditions in LWRs. Phenomena mass composition and temperature of the core to be studied are those occurring in the core, in debris at the time of vessel (or RCS) failure; and the primary reactor coolant circuit, and in the (6) the mode of vessel failure. The overall program containment. This agreement is of significant is divided into three main activities:(1) the benefit to the NRC because, at a relatively modest behavior and chemistry of fission products cost, the NRC can participate in the PHEBUS-FP released during core melt, (2) in-vessel core-melt project over its lifetime. The NRC will be able to progression, and (3) fuel-coolant interactions. The obtain integral experimental data to further in-vessel core-melt progression and hydrogen validate its analytical models for fission product generation work has includd in-reactor experi-transport in the reactor coolant system and con-ments, out-of-reactor experiments, examination of tainment and for iodine chemistry in the contain-specimens from TMI-2, and analytical model ment. The experimental data from PHEBUS-FP development. The research on the amount, the is confirmatory in nature and will be used to chemical form, and the behavior of the fission assess the revised source term assumptions used products released from the fuelin the course of a in NUREG-1465.

severe accident requires experiment at high fuel temperatures. Here the core geometry is changing The first PHEBUS-FP test, FPT-0, was suc-l and fission product chemistry and its effect on the cessfully conducted in December 1993. The retention of fission products within the RCS are interpretation of FPT-0 is continuing; lessons significant. The FCI work is focused on the learned from FPT-0 will be taken into account in development and validation of appropriate phe-planning for the next test, FPT-1, which is nomenological and analytical models addressing scheduled for June 1995.

the fundamemal aspects of FCI, namely, melt quenching and FCI energetics.

5.3.3.2 Core-Melt Progression "In-vessel core-melt progression" describes the 5.3.3 Research Accomplishments in FY 1994 state of an LWR core from core uncovery up to 53.3.1 Source Terms reactor vessel melt-through in unrecovered acci-dents or through stabilization of the temperatures

" Source terms" refer to the magnitudes of the and the core geometry in accidents recovered by I

radioactive materials released from a nuclear core reflooding. Melt progression provides the reactor core to the containment atmosphere, initial conditions for assessing the loads that may taking into account the timing of the postulated threaten the integrity of the reactor vessel and the

+

releases and other information needed to calculate containment. Significant results of melt pro-i offsite consequences of a hypothetical severe gression are the, melt mass, composition, I

accident. NRC research in this area is reflected in temperature (superheat), and the rate of release of the updated version of TID-14844, which has the melt from the core and later from the reactor been in use for three decades in connection with vessel if vessel failure occurs. Melt progression plant siting assessments. An extensive review of research also provides information about the the update of TID-14844 published in draft in-vessel hydrogen generation, the conditions that NUREG-1465, " Accident Source Terms for Light govern the in-vessel release of fission products Water Nuclear Power Plants," has been and aerosols and their transport and retention in NUREG-1266 5-8 w

5. Reactor Accident Analysis the primary system, and the core conditions for temperature distributions. Separate melts of assessing accident management strategies.

metallic Zircaloy and control-blade materials are poured into a test assembly at prototypic rates Current NRC research on melt progression is (dribbles), and the melt relocation and blockage focused on two major issues. The first issue is behavior are determined.

determining whether there are any accident conditions for BWRs (and possibly PWRs)in Last year two developmental XR1 tests of the which a metallic core blockage similar to that at experimental system were successfully performed.

Three Mile Island Unit 2 (TMI-2; Pa.) would not In FY 1994, preparations were made for a series be formed. In this case, the metallic melt, and f f ur XR2 experiments under closely prototypic condit,ons to provide definitive data on the i

later the ceramic (fuel) melt, would drain when formed from the core into the water in the lower question of metdhe melt drainage or core block-plenum of the reactor vessel. His would lead to E".under BWR dry core accident conditions. A m j r part of the FY 1994 effort was the develop-release of a mostly metallic (Zircaloy), much lower ment of a new melt delivery system to furmsh the j

temperature melt after the reactor vessel boils dry required separate pours (dribbles) of control and fails. The second issue concerns the condi-blade and Zircak)y melt at prototypic times, rates, tions for melt-through of the growing pool of ceramic melt above the metallic blockage. He temperatures, composition, and location at the entrance to the test assembly. If techmcally melt-through threshold and location determine the feasible, and subject to program review, the XR2 mass of the melt released from the core and later, experiments are to be performed in FY 1995.

potentially, from the reactor vessel.

5.3.3.3 Reactor VesselIntegrity On the issue of bk>ckage of the core by metallic During the late phase of a severe accident, a melt, TMI-2 and the results of in-reactor tests significant amount of core material may relocate and laboratory experiments have mdicated that' downward into the lower head of the reactor for,' wet core conditions (with water m the vessel. When this molten core materialis re-bottom of the core), the relocat, g molten metalh.e m

located into the lower head of the reactor pressure Zircaloy in the core freezes to block the lower vessel, a molten pool forms and can impose a core. All but one of the previous experiments for significant heat load on the reactor vessel lower both PWRs and BWRs were performed for these head. Knowledge of in-vessel and ex-vessel heat wet core conditions, and this one experiment did transfer phenomena to the lower head is needed The emergency operatm, e or drainage question.

to assess the ability of the reactor pressure vessel not address the blockag g procedures for U.S..

to maintain its integrity during a severe accident.

BWRs, however, call for reactor depressurization, When a molten pool forms on the lower head, a which lowers the water level below the reactor solid crust of material forms around the periphery core by flashing so that core heatup occurs with of the pool, but internal heat generation resulting very low steam flow through a dry core., Analy-from radioactive decay of fission products ensures sis of these conditions mdicates that the molten that most of the debris remains molten and,in core metal (and later molten ceramic fuel) might fact, undergoes significant internal natural convec-possibly dram from the core rather than formmg tion in the pool. Detailed understanding of this a blocked core as at TMI-2 with subsequent natural convection process provides information cerarme melt pool growth and melt-through.

on the local heat flux distribution around the inside surface of the crust. This distribution, in A series of new ex-reactor (laboratory) experi-conjunction with the thermal boundary conditions ments to address the question of metalhe melt imposed on the outer crust surface, determines drainage or core plate blockage under BWR dry the fraction of the total heat dissipation that is core accident conditions has been started at the transferred through the upper crust to the inside Sandia National Laboratories (SNL). The experi-of the reactor vessel by radiative heat exchange mental test assemblies are a mockup at full radial and the fraction that must be conducted through scale of a cross section of the lower quarter of a the wall of the reactor vessellower head.

BWR core (and core plate region) where such blockages might occur, and the test assemblies In August 1994, the NRC, in cooperation with 13 have prototypic reactor fuel rods, structure, and countries and under the auspices of the 5-9 NUREG-1266

5. Reactor Accident Analysis Organization for Economic Cooperation and significance both for the expansion work and for Development's (OECD) Nuclear Energy Agency impulsive shock failure of reactor structures.

(NEA), undertook an investigation of melt-vessel interactions to provide data on the internal Since the quantification of the probability of a natural convection flow and local heat flux steam explosion-induced missile from the distribution inside the lower head of the reactor expansion work as a possible mode of contain-pressure vessel for various melt compositions.

ment failure (alpha mode) in the reactor safety This program involves large-scale integral study, WASH-1400, significant progress has been tperiments using molten UO and ZrO in made in understanding the limitations on the i

2 2

representative reactor lower head geometries; formation of such potential missiles by an in-analytical studies; and a number of small-scale vessel steam explosion. Alpha-mode failure was separate effects experiments. This program, not a dominant contributor to early containment named OECD RASPLAV,is being performed at failure in NUREG-1150. The emphasis prior to the Russian Research Center.

NUREG-1150 in fuel-coolant interaction (FCI) research was on the alpha containment failure mode process of in-vessel molten fuel pouring into In order to remove the fraction of heat conducted lower plenum water and the probability of causing through the vessel lower head, the concept of missile generation and containment failure by an flooding the reactor cavity to externally cool the energetic interaction (steam explosion). Current reactor pressure vessel lower head and prevent its emphasis in steam explosion research has shifted failure is being investigated. One major uncertain-to impulsive shock loading of ex-vessel structures.

ty involved in the external cooling of the lower For application of the experimental results on head is the critical heat flux distribution on the FCis, an Integrated Fuel-Coolant Interactions i

bottom curved surface of the reactor vessel. An (lFCl) code has been developed by SNL experimental program is under way at the Pennsylvania State University to address ex-vessel The NRC and the Safety Technology Institute of flooding of the reactor cavity to prevent vessel the Joint Research Center (JRC) of the Com-failure. The program investigates boiling heat-mission of the European Communities at Ispra, transfer on downward-facing surfaces in hemi-Italy, have entered into a technical exchange spherical and toroidal geometries. The results of arrangement to perform a series of experiments in this study include data on the critical heat flux the FARO facility at Ispra on melt breakup and (CHF) and the development of an analytical cooling in water. In this facility, a large mass of model for the CilF on downward-facing surfaces.

reactor fuel (and other prototypic reactor core The experimental apparatus was designed and melt materials)is melted and poured into built during FY 1994 and a series of transient different depths of water at a high pressure that experiments performed. Further experiments, suppresses steam explosion triggering. In the JRC analyses, and CIIF model development will KROTOS facility, steam explosion energetics continue in FY 1995.

(including shock impulse) with prototypic melts are also under investigation. Four melt cooling tests have been performed in FARO, one of 5.3.3.4 Fuel-Coolant Interactions which, in FY 1994, used 150 kg of UO -ZrO melt 2

2 with 3% zirconium. Steam generation and the There are several aspects of the interaction of melt cooling characteristics have been measured ceramic (fuel) melts and of metallic melts with in all these tests. In the KROTOS steam explosion water coolant in the reactor vessel and also experiments at one atmosphere, a series of tests ex-vessel in a flooded reactor cavity that are with tin and with Al O and UO -ZrO ceramic 2 3 2

2 significant in reactor safety assessment. The first melts have been performed. Seven of these tests of these aspects is the non-explosive breakup and have used UO -ZrO melts, five of which were 2

2 cooling of the melts in water with both steam performed in FY 1994. The results are currently generation, and, for metallic melts, oxidation and being analyzed.

hydrogen generation. 'Ihe cooling of the melt is significant both in-vessel, for reactor vessel RES has an ongoing program of FCI research at integrity, and ex-vessel. Explosive melt-coolant the University of Wisconsin. This research interactions (steam explosions) have reactor safety includes (1) simulant material experiments on the i

NUREG-1266 5-10

1 l

5. Reactor Accident Analysis l

mechanisms of both non-explosive and explosive 5.4 Reactor Containment Structural FCIs:(2) tecimical participation in and analysis of Integrity the results of the FARO and KROTOS FCI experiments: and (3) assessment of the IFCI code 5.4.1 Statement of Problem agamst the FARO and KROTOS results. During FY 1994, an initial series of experiments with The major source of risk to the public from the molten tin was completed. and the results were operation of nuclear power plants stems from analyzed and interpreted. A principal result was accidents that lead to a containment failure. The the importance of the fraction of the melt mass regulatory concern is that the failure modes and that actually interacts with the water in a steam associated load levels for containment structures explosion.

cannot be predicted with any real confidence by the methods used for design. This is especially so if the contemplated failure mode is localized leakage. Both assessments of the risk posed by An experimental program has been started at the loads outside the design basis and estimates of Argonne National Laboratory to determine the effectiveness of proposed mitigative steps whether chemical augmentation of the energetics require an ability to predict the way in which a can occur in Zircaloy melt-water steam explo-containment will fail.

sions. Such chemical augmentation can occur in aluminum melt-water steam explosions and has increased the energetics by a factor of up to five.

5.4.2 Program Strategy This possible chemical augmentation of the Research on containment failure modes is based energetics is of particular importance in assessing on the observation that excessive leakage can impulsive shock loads to structures. In FY 1994, occur, basically, from the following sources:

detailed experiment planning and construction of the apparatus were performed. The experiments Failure of the shell, either the containment e

are to be performed in FY 1995.

shell itself (in the case of steel containments) or the liner (in the case of concrete containments);

5.3.3.5 In-flous,e Severe Accident Analysis Leakage at large penetrations as a result of e

Capabihty the inelastic deformations and/or degradation of seals and gaskets:

Growth in the capability of workstation-level Leakage at electrical penetrations due to e

computers provides an opportunity for runm,ng degradation of materials under the high severe accident codes on other than main frame temperatures associated with accident computers. In FY 1994, RES purchased work-scenarios: and stations to enhance the in-house analysis capa-bility at NRC. Reactor plant descriptions, or Leakage through valves due to pressure and e

decks, for analyses using the MELCOR, SCDAP/

temperature effects.

RELAP5, CORCON, CONTAIN, and VICTORIA codes have been installed on the Research related to shell failure or deformations workstations. Typical uses of this in-house of penetrations rests on analyses of and experi-capability have been to review input decks ments on model tests of actual containment developed by NRC contractors, using these decks designs. These tests involve pressurization up to to extend previous analyses. In-house analyses failure levels under ambient temperatures. Since have been used to check new models in the codes seal and gasket materials are adversely affected by and to do bounding calculations to determine the the temperatures associated with severe accidents, appropriateness of the new models.

separate tests focusing on the development of 5-11 NUREG-1266

5. Reactor Accident Analysis leakage are performed under pressure and thickness. This selection of scales allows the temperature conditions, usually at full scale.

model to be small enough for transportation from Examining the possibility of developing leakage Japan to SNL while being thick enough to ensure through electrical penetration assemblies and quality construction.

valves also requires experiments under tempera-ture and pressure conditions at full scale.

The model fabrication, under way at Hitachi Works, Hitachi, Ltd., in Japan, was completed in November 1994, and tus model will be trans-5.4.3 Research Accomplishments in FY 1994 ported to SNL in January 1995.

5.4.3.1 Containment Studies The prestressed concrete contamment vessel The major undertaking in this program for the (PCCV) model will be a scaled representation of next few years will be a cooperative one with the an actual PCCV in Japan, which was designed in Ministry of International Trade and Industry accordance with the Japanese Concrete Con-(MITI) of Japan. Two areas of cooperation have tainment Vessel Design Code. The actual PCCV been identified-one dealing with steel contain.

consists of a hemispherical dome, a cylindrical ments used in both the United States and Japan wall, and a basemat. Two buttresses are used to for BWR designs, the other relating to prestressed anchor the horizontal or "h(x)p" tendons. In the concrete containments. The current generation of vertical direction, a " hairpin" tendon layout is Japanese PWR containments are of prestressed employed. The vertical tendons extend from the concrete designs. In the United States, there are basemat up through the cylinder wall, over the 41 prestressed concrete containments compared dome, and back to the basemat on the opposite to 20 reinforced concrete containments.

side of the containment. They are anchored in a tendon gallery that is inside the basemat. A liner P ate which is made of carbon steel,is placed on l

A reinforced concrete model was chosen for the NRC-sponsored testing at SNL that was per_

the inner surface of the concrete wall, dome, and formed in 1987. Subsequent analyses of the results basemat and forms the containment pressure of that model test have shed light on how poten.

boundary m these areas.

tial failure modes develop in concrete contain-ments. Some of the results are felt to be applica-The basic design of the PCCV model will be ble to prestressed concrete containments as well.

completed by the end of 1994. Construction llowever, there are two main reasons for perform-drawings will be prepared soon for construction activities at SNL that are scheduled for 1995-1997.

ing an additional prestressed containment model Instrumentation of the model will be conducted in test:

1997-1998, partly in parallel with the onsite model Prestressed designs are the most common c nstruction. Testing of the PCCV model will then o

concrete PWR containment type in the take place late m 1998.

United States as stated above.

5.4.3.2 Containment Corros. ion Studies The margin between the ultimate capacity Recent experience suggests the possibility that o

and the design pressure for prestressed corrosion effects may significantly degrade the concrete containments is now thought to be margin that containments have to accommodate somewhat lower than that for reinforced accidents beyond their design basis. Evidence of concrete or steel containments; hence, it is corrosion has been found in both Mark I BWR important to have accurate predictions of the containments and in ice condenser PWR con-ultimate behavior of prestressed concrete tainments. The robustness of containments, as containments.

verified in the tests performed at SNL, showing their capacity to sustain loads well beyond design The steel containment vessel test specimen is a levelis a major support for the Commission's scale model representing some features of an Severe Accident Policy Statement. Thus, we need improved UWR Mark 11 containment vessel in to understand the significant factors relating to Japan. A scale of 1:10 is used for the overall occurrence of corrosion, efficacy of inspection, geometiy of the model with 1:4-scaling of the wall and capacity reduction so as to be able to NUREG-1266 5-12

i

5. Reactor Accident Analysis formulate regulatory requirements that will ensure address inspection ofinaccessible areas in con-the continued availability of sufficient margins.

tainments. Some of the instances of containment degradation suggest the possibility that degrada-Comparison of remaining thickness against mini-tion may have occurred in inaccessible areas. As mum ASME Boiler and Pressure Vessel Code noted in a NUMARC (now NEI) report on PWR requirements is the obvious first line of assess-containments, the state of practice for inspection ment. If the remaining thickness exceeds the limit, of inaccessible areas will have to be improved a decision on adequacy of margin is easy. How-before a resolution of this issue is achieved.

ever, degradation beyond that limit at localized h> cations does not, by itself, suggest loss of 5.5 Severe Accident Policy contamment capacity. It may be that, unless the degradation is especially severe, failure at some Implementation other location will still control. However, the 5.5.1 Statement of Problem clastic analysis methods used for design cannot be extrapolated to provide estimates of actual failure.

A severe accident in a nuclear power plant is an Methods, using the results of research on actual event in which the core is damaged and there is a failure modes of containments, are being sought potential for release of large amounts of fission that can relate containment capacity to amount products. Significant research has been performed and h> cation of degradation. If this effort is on the likelihood, progression, and consequences successful, a basis can be found for judging the of a severe accident as discussed earlier. Much of seriousness of a given degree of degradation at a this work has concentrated on the performance of particular location. The Oak Ridge National the containment during a severe accident, includ-Laboratory initiated a program during FY 1994 to ing potential containment failure mechanisms and assess state-of-the-art nondestructive testing the ability of the containment to mitigate the techniques for examining steel containments and consequences of a severe accident.

the liners of concrete containments. As part of this program, statistically based sampling plans In the Commission policy statement on severe will be developed to provide confidence limits on accidents m nuclear power plants issued on detection of corrosion occurrence. SNL initiated a August 8,1985, the Commission concluded that program during FY 1994 to investigate and existing plants pose no undue risk to the public develop analytical methods to account for the health and safety and that there is no immediate effects of corrosion on the capability of steel need for generic rulemaking related to severe containments to withstand static internal accidents. However, based on NRC and industry overpressurization loads associated with severe experience with plant-specific probabilistic risk accident conditions.

assessments, the Commission recognized the need for a systematic examination of each existing plant 5.433 Rulemaking to identify any plant-specific vulnerabilities to severe accidents. The policy statement indicated In order to improve the state of practice in the intent of the Commission to take all inspection of containments to reduce the chances reasonable steps to reduce the probability of a of having significant undetected degradation due severe accident and, should a severe accident to corrosion, work continued in 1994 on the occur, to mitigate its consequences to the extent rulemaking to incorporate by reference Subsec-possible. As part of the implementation of the tions IWE and IWL of Section XI, Division 1, of Commission's Severe Accident Policy Statement, the ASME Boiler and Pressure Vessel Code into the staff has required individual plant examina-

- 10 CFR 50.55a. Subsection IWE provides rules for tions (IPEs) of all existing plants to identify any the mservice mspection of metal contamments plant-specific vulnerabilities to severe accidents.

and the liners of concrete containments. Subsec-i tion IWL provides rules for the inservice inspec-Much of the work performed to implement the j

tion of the reinforced concrete and the post-Severe Accident Policy Statement has focused on tensioning systems of concrete containments. As research into phenomena that would occur during written, Subsection IWE and Subsection IWL severe accidents and methods to systematically address only the accessible areas of containments.

discover vulnerabilities for severe accidents. This i

A provision was included in the proposed rule to work has shown that the causes and consequences i

5-13 NUREG-1266

5. Reactor Accident Analysis of severe accidents can be greatly influenced by 5.5.3 Research Accomplishments in FY 1994 nuclear power plant operators and that many vulnerabilities to severe accidents can potentially in the 14 years since the Three Mile Island (Pa.)

be eliminated by proper operator actions. The accident, the NRC has sponsored an active TMI-2 accident and other abnormal occurrences program in research on severe accidents at in nuclear power plants have shown that operators nuclear power plants as part of a multifaceted do not stand idle but actively intervene in approach to the assurance of safety in this attempts to control the event. If operators are context. Other elements of this approach include provided with proper guidance and training to improved plant operations, human factor con-take beneficial actions when needed and, most siderations, and probabilistic risk assessments.

importantly, refrain from actions that can have adverse effects, the consequences of a severe accident can potentially be significantly reduced.

The IPE process involves two different efforts.

Smec many accident management strategies do The first is an examination of existing plants for not myolve sigmhcant plant design changes' vulnerabilities to severe accidents resulting from substantial safety benefits can be qu,ckly ach.ieved events occurring within the plant (e.g., equipment i

by ensuring proper operator actions. Ihus, the failures, pipe breaks). The second effort is to consider severe accident vulnerabilities from mitiation of accident management programs at operating plants is a logical result of the JPE external hazards (e g., earthquakes, floods, winds).

This activity is referred to as the IPEEE.

process.

This program element provides for the implemen-Fifteen submittals for internal events were tation of the Commission's Severe Accident Policy received from licensees in FY 1994, making an Statement and the application of the results of overall total of 76 submittals received to date, severe accident research directly to the regulatory Staff evaluations were issued for FitzPatrick process. Modification of the Commission's rules (N.Y.). Surry 1 & 2 (Va.), Millstone (Conn.),

or policies regarding siting, emergency planning, Monticello (Minn.), Palo Verde 1,2, & 3 (Ariz.),

and containment design are examples of areas in Perry 1 (Ohio), Nine Mile 2 (N.Y.), Oyster Creek which the results of severe accident research may (NJ.),11. B. Robinson 2 (N.C.). Browns Ferry 2 affect future changes.

(Tenn.), McGuire 1 & 2 (N.C.), Catawba 1 & 2 (S.C.),and Haddam Neck (Conn.), and draft staff 5.5.2 Program Strategy evaluations were completed for Sequoyah (Tenn.)

and Watts Bar 1 (Tenn.). It is expected that all RES has been given the responsibility for the IPE submittals will be received and reviewed by implementation of the IPE. This implementation the end of calendar year 1996.

has involved developing guidance for performance of the IPE, preparing a generic letter to plant operators requesting the IPE, and developmg The woach for the review of the IPEEE will follow closely that developed for reviewing the review plans and reviewing the results of the II,E.

internal-event IPE submittals. The staff com-submittals. The requirement to correct any identi-fied plant-specific vulnerabih, ties not voluntarily pleted the procurement process to obtain con-tractual assistance for the IPEEE reviews. Sixteen corrected will be determined by the backfit rule Accident management is not required as part of omplete and four partial IPEEE submittals have the IPE process but was highlighted m the 11 E be mived of which four are being evaluated.

generic letter as a future requirement that will make use of the results of the IPE process. Severe Studies began of the IPE results to gain more accident vulnerabilities due to external hazards generic insights. Issues such as the plant-to plant (e.g., earthquakes, ikiods, fires) are being variability in estimated core damage frequency considered under the IPE for External Events results and the reasons for this variability are (IPEEE) program.

being examined.

l NUREG-1266 5-14

1 1

i

6. SAFETY ISSUE RESOLUTION AND REGULATION IMPROVEMENTS 6.1 Earth Sciences science research program to significantly reduce the uncertainty.in seismic hazard estimation in.

6.1.1 Statement of Problem the next decade through emphasizing this type of research.

- Eaithquak's are among the most severe of the e

natural hazards faced by nuclear power plants.

6.1.2 Program Strategy Very large carthquakes would simultaneously challenge the ability of all plant safety systems to The strategy to resolve the s.eismic prob!cm function and, coupled with the likely ioss of offsite involves research to develo'p the methods and data power and dependent safety systems, could pose a that will support the necessary seismic criteria..

unique threat to public safety. As with many development and provide the evaluation tools. The potentially severe conditions, there is much research is focused on (1) improving estimates of uncertainty associated with the design and eval-earthquake hazards by identifying potential uation of nuclear plants for earthquakes. Seismic earthquake sources and determining the propa-hazard in the Eastern and Central United States gation of seismic energy with distance,(2) esti-remains an issue that is not likely to' be easily mating the possible range and likelihood of resolved. These regions contain the highest seismic ground motions at nuclear plant sites, and percentage of nuclear power plants in the United (3) assessing the effect of these ground motions on States.

soil, structures, equipment, and systems of the plants. The integrated results'of this research will IIistorically, the largest carthquake in the United be used to quantify the risk to nuclear plants from States has occurred at New Madrid, Missouri.

earthquakes, to assess the seismic safety margins The geology of the central and castern regions inherent in current or future plant design, and to makes it difficult to estimate earthquake magni-help identify and set priorities for what improve-tudes or seismic parameters for specific locations ments are needed in plant designs or what parts or to ensure a proper design basis for individual of seismic design criteria may be relaxed.

power plants.

A major focus of the NRC research programs in The publication of seismic hazard curves in 1989 geology, seismology, and geophysics continues to by both the NRC (NUREG/CR-5250) and the be identifying and defining potential earthquake Electric Power Research Institute (EPRI) sources or source zones in the Eastern United (NP-6395) marks the end of major efforts to States and using that information in assessing characterize the seismic hazard at U.S. nuclear seismic hazards with respect to nuclear power reactor sites. Although the best information and plants. Many unknowns exist regarding these procedures available were used, they revealed that issues, including a strong basis for seismic large uncertainties still remain in seismic hazard zonation, source mechanisms, characteristics of I

estimates. Also, recent full-scope probabilistic risk ground motions, and site-specific response. The '

assessments, performed as part of the NUREG-NRC is addressing these uncertainties through 1150 effort, continue to show that seismic hazard research that encompasses sustained seismic uncertainties contribute significantly to the overall monitoring, geologic and tectonic studies, uncertainty in nuclear reactor risk estimates, neotectonic investigations,' exploring the earth's These large uncertainties make it difficult to place crust at hypocentral depths, and conducting the contribution of seismic risk into its proper ground motion studies.

perspective, e.g., in the development of individual plant examination guidelines.

The backbone of the NRC program in the Eastern United States has been the seismographic net-Recent successes in the geological, geophysical, works deployed throughout the Eastern and and seismological studies sponsored by RES show Central United States. The NRC is currently -

that it is possible to answer the basic scientific funding seismographic networks in the following questions that underlie these seismic hazard

. regions: Northeastern United States; Virginia; uncertainties. It' is the goal of the NRC earth Charleston, South Carolina; the southern 6-1 NUREG-1266

6. Safety Issue Resolution Appalachian region; the New Madrid (Missouri) 6.13.2 Southeastern Tectonics region; Ohio and Indiana; eastern Kansas; and Oklahoma. An agreement was reached in 1986 A continued search for possible liquefaction features and investigations of landslides and cave between the United States Geological Survey (USGS) and the NRC to jointly support the deposits in the southern Appalachian area did not establishment of the castern portion of a national produce any clear evidence of prehistone carth-seismographic network. The eastern portion of the 9".akes. This activity concluded the search for national network is now fully in place.

evidence of prehistonc earthquakes in the South-east outside of the Charleston, S.C., area. While the search did not result in positive evidence, the fact that clearly liquefiable deposits outside the j

6.1.3 Research Accomplishments in FY 1994 coastal area did not show evidence of liquefaction i

reinforces the impression that the Charleston area The objective of NRC research in earth sciences, is unique with respect to its seismicity.

as related to reactor regulation, is to define potential earthquake ground motions at nuclear In an earlier investigation of paleoliquefaction power plant sites. This information provides a features on the Atlantic Coastal Plain, based on basis for evaluating the effects of earthquakes on evidence for a prehistoric event that occurred the plants and their safety systems.

about 70 kilometers north of Charleston, S.C.,

about 1800 years ago and on the absence of evidence for that event in the Charleston Seismic hazards contribute significantly to overall plant hazards and, because of inherent difficulties cizoseismal area, investigations have been started to determme whether there is more than in defining the seismic hazards, they form an even more significant portion of the overall uncertainty one scismic source for a large carthquake m South Carchna.

in estimating plant hazards. In order to reduce these uncertainties, rescarch into the causes and 6.1.3.3 Palcoscismicity of Southern Illinois and distribution of seismicity is continumg. Research Indiana is also progressmg on improved methods of applying carth science information to estimates of Previous liquefaction studies have found indica-ground motion levels for use in plant design.

tions that a large earthquake, centered near Vmeennes, Ind., occurred between 2,500 and 7,500 6.1.3.1 Seismographic Networks ye rs ago. This earthquake may have been larger than the 1886 Charleston earthquake (magmtude

aPProximately 7.0) but smaller than the 1811-The new National Seismographic Network (NSN) 1812 New Madrid carthquakes (magmtude

was established through a cooperative agreement approximately 8.0). During this report period, between the NRC and the USGS. Includmg coop-investig'itions of the Wabash drainage system erative stations, the NSN operates 32 broadband were extended farther into Illinois and Indiana three-component stations and satellite telemetry, and also into the Anna, Ohio, seismic area.

thus providmg data on sigmficant earthquakes withm, mmutes. During FY 1994, a broad agency Evidence for another, smaller prehistoric earth-quake was found in the Vincennes region, but no announcement was issued solicitmg research evidence for prehistoric events large enough to proposals to analyze NSN data and other avail-cause liquefaction was identified in the Anna, able seismological, geological, and geophysical Ohio, area. The results thus far confirm the data. A number of proposals were received, conservatism of past licensing decisions made resulting m hve research contracts. This research regarding the seismic hazard at nuclear power will continue the type of investigations previously plant sites in this region.

carned out by the umversities operatmg regional networks. The high-quality, broadband, and three-6.13.4 New Madrid Seismic Zone component data of the NSN willlead to new insights into the causes and distribution of Paleoliquefaction studies are being conducted at l

seismicity and the ground motion propagation several locations in the New Madrid seismic zone, l

characteristics of the earth's crust, particularly in particularly near the Missouri-Arkansas state line, the Eastern and Central United States.

to determine the ages and extent of prehistoric NUREG-1266 6-2

6. Safety Issue Resolution earthquakes. Evidence, both geological and methods used to estimate the seismic hazard of archaeological, indicates the occurrence of at least the Diablo Canyon nuclear power plant site.

two prehistoric events.

Preliminary results indicate that the surface Seismic reflection investigations are being par-deformation at Northridge is most likely the result tially supported by the NRC in the New Madrid of strong ground motions and secondary faulting.

seismic zone in the area where waterfalls Evidence pointing to prehistoric events of a appeared in the Mississippi River during the similar kind was also found.

1811-1812 carthquakes. The studies are investi-gating the possibility that these short-lived 6.1.3.7 Strong Ground Motion Studies features were caused by faulting associated with those earthquakes.

The NRC supports several cost-sharing ground rnation programs in cooperation with the USGS.

The investigations are part of the ongoing effort Anmng tkse am sm@tmns@in th Eastern and madng Ngb to estimate the recurrence of the far8e-to-8reat mquency gmumi nm earthquakes (magm.tudes 6 to 8)in the New Central United States, using data from the Madrid seismic zone and to define the causative Landers, Petrolia, and Northridge earthquakes; faults.

attenuation and source parameter studies for the Eastern and Central United States, using data an simng gmund nmdon studes 6.1.3.5 West-Central United States of large intraplate earthquakes, using data from Three suggested Quaternary faults are being teleseismic and regional recordings. The results of investigated, namely the Cheraw and Fowler faults these studies will be used to make more realistic on the Colorado Piedmont and the Harlan County hazard determmations and to reduce uncertainties fault in Nebraska. Preliminary results indicate in the probabilistic estimates.

that the Fowler feature is not a fault but appears to be one because of the current setting of the 6.1.3.8 Geochronolog.ical Studies Quaternary stratigraphy, geomorphic features, The NRC is supporting a research program to i

and jointing in the underlying Pierre Shale. The assemble all state-of the-art information on Cheraw and Harlan County faults appear to be methods for determining the age of geological Quaternary tectonic faults and will be investigated materials. Geochronological analysis of faults, further. The investigations are part of an ongoing paleoliquefaction features, and other paleoseismic effort, which began with the discovery of late features are an important part in determining the i

Holocene displacement on the Meers fault, to seismic and geological hazard of a site. A limited identify other Quaternary faults in the Central field research project to validate new dating United States.

l methods is under way as a part of this project.

The goal of this project is to develop a regulatory 6.1.3.6 Fault Segmentation Studies guide to assist applicants and the regulatory staff During the past 2 years, the surface rupture that waluadng p tential nuckar shes.

occurred during the 1992 Landers earthc[uake was 6.1.3.9 Crustal Strain Measurements studied m detail, and durm, g the last halt of FY 1994 the surface deformation that occurred during A 45-station crustal strain network for the East-the 1994 Northridge earthquake was investigated.

ern and Central United States was established in Geological evidence from faults that ruptured the 1987 and measured for the third time during FY ground surface at Landers indicates that from two 1993. After this strain network was established,it to four prehistoric earthquakes occurred in this became the backbone of a new geodetic network area. These events are estimated to have been for the United States based on Global Positioning about the same size as the 1992 event, suggesting System (GPS) measurements. In the meantime, that the Landers earthquake fault segments high-precision GPS networks have been estab-behaved in a similar manner in the past. These lished in many states, and within the next few findings confirm the validity of using fault years all of the United States will be covered with segmentation to estimate future fault behavior detailed high-precision GPS networks, including and earthquake magnitude, which was one of the continuously operated GPS stations. From 6-3 NUREG-1266

6. Safety issue Resolution preliminary strain analyses, it appears that strains In the 1980s, a considerable effort was made to in the Eastern and Central United States are in better predict the potential response of nuclear 4

the range of 10 per year. Because strain rates in plants to earthquakes greater than those con-this region are so low, many years may be needed sidered in design. Our understanding has been to arrive at meaningful strain determinations.

increased greatly by the testing to failure of Ilowever, with the many high precision GPS equipment and structures, by the gathering and stations now available,it should eventually be synthesis of earthquake experience data from possible to get a very detailed picture of strain non-nuclear facilities, and by the large number of distribution. Information on strain distribution seismie probabilistic risk assessments (PRAs) that and strain rates will then provide a basis for have been made.

refinements in seismic hazard determinations.

This research has generally found that the seismic 6.l.3.10 Probabilistic Seismic Ilazard capacity of important nuclear plant structures and Assessments equipment (when properly anchored)is high. But there remain specific capacity concerns that need During FYs 1993 and 1994, a panel of scientists, to be resolved, such as how to address the po-assembled under the sponsorship of the NRC and tentially harmful effects of relay chatter. The the Department of Energy (DOE) and with input importance of plant-specific walkdown reviews to by the Electric Power Research Institute (EPRI),

find nongeneric vulnerabilities has been noted in conducted a study of probabilistic seismic hazard recent seismic margin studies.

analysis (PSilA) methodologies that is now near-ing its end. The study has the goal of analyzing existing methodologies, namely those developed 6.2.2 Program Strategy by Lawrence Livermore National Laboratory and In recent years, the NRC has supported seismie by EPRI under the NRC and nuclear utih,ty testing and the collection of earthquake experi-sponsorship, respectively, and of denvmg an ence data in order to improve and gain confidence improved methodology that wdl be scientifically in the use of seismic PRAs and seismic margin balanced and usable for regulatory decismns over studies. These data are also being used to support the next decade. A draft of the final report was proposed improvements to seismic design criteria.

completed in November 1994, and a prehm, ary The earthquake resistance of structures, equip-m review meeting was held with the National ment, and piping has been found, in general, to bc l

Academy of Sciences review panel in December.

higher than previously thought. Major efforts in l'he study is bemg peer reviewed by a panel this area were completed in 1990, and the results appointed by the National Academy of Sciences are being successfully used in licensing actions.

to ensure impartiality and objectivity. Consider-Relay chatter is the Ane remaining seismic able weight has been placed on methods of capacity issue that will require additional testing cliciting expert opinions, which are of funda-to resolve.

mental importance in probabilistic hazard estimates. After the final report for this study is Upcoming individual plant examinations and USl issued, the new PSIIA methodology will be A-46 seismic reviews will ese the recent results of verified with a limited amount of site testing.

NRC seismic research.

6.2 Plant Response to Seismic and 6.2.3 Research Accomplishments in FY 1994 Other External Events 6.2.3.1 Revision of Appendix A to 10 CFR Part 100 6.2.1 Statement of Problem In August 1994 the Commission approved the in the 1970s and before, our interest in nuclear staff's recommendation to issue a second plant seismic design was mainly limited to proposed revision of Appendix A. " Seismic and response at design levels (e.g., operating-basis Geologic Siting Criteria for Nuclear Power carthquake (OBE) and safe-shutdown earthquake Plants," to 10 CFR Part 100, " Reactor Site (SSE)) and our knowledge of this was primarily Criteria," for public comment. This revision based on analytical techniques and assumptions.

rellects new information and research results l

NUREG-1266 6-4

l

6. Safety Issue Resolution available since the first proposed revision to the computer codes used in SSI analyses of nuclear regulations was issued and comments were re-power plant structures. Observations will be made ceived from the public on that proposed revision.

on the motions of the reactor building model and l

The second proposed revision to the regulations the surrounding ground during large-scale earth-was published in the Fedeml Register for a 120-day quakes. The expectation is that the test model will comment period on October 17,1994 (59 FR be shaken by numerous earthquakes in this 52255). Draft regulatory guides and standard seismically active area of Taiwan.

review plan sections providing methods accept-able to the NRC staff for implementing the To date there have been several earthquakes proposed regulations were issued for public recorded at the LSST site-one on September 16, l

comment in February 1995.

1993 (4.1 magnitude) and one on January 20,1994 (5.7 magnitude). Instrumentation located on the 6.2.3.2 Seismic Analysis of Piping scale model and in the field along a three-j dimensional strong ground motion array recorded Activities on the program to review significant the recent carthquake data. The LSST program at changes being proposed on portions of the ASME Hualien Taiwan,is a follow-on to the SSI Boiler and Pressure Vessel Code, Section Ill, experiments at Lotung, Taiwan.

Nuclear Power Plant Components, Division 1, continued in FY 1994. ne workscope hicludes the EPRI has organized the Hualien LSST experiment evaluation of completed and ongoing imtiatives; and coordinated participation with the Taiwan for instance, results of the EPRl/NRC piping Power Company (Taipower), the NRC, the Central 4

program, Advanced Reactor Corporation and Research Institute of Electric Power Industry IPIRG activities, NRC staff positions associated (CRIEPI), the Tokyo Electric Power Company with advanced light-water reactor reviews and (TEPCO), the Commissariat a l'Energie climination of the OBE response analysis, and Atomique (CEA), Electricite de France (EdF),

interaction with a peer review panel of nationally Framatome, the Korea Power Engineering Co.

recognized experts.

(KOPEC), and Korea Electric Power Corp.

6.2.3.3 Cooperative International Seismic During this report period, a collaborative effort Programs involving exchange of technical information was The NRC's Partici ation in international seismic established with the Ministry of International P

test programs is beneficial both for the sharing of Trade and Industry (MITI) and Nuclear Power research resources and for gaining different Engineering Corporation (NUPEC) of Japan, in this effort NUPEC is carrying out a seismic perspectives on seismic design issues. The poolm.g proving test program for a main steam line typical of resources allows the development of larger.

of the PWR plants and a feedwater system typical scale tests, an important element in the validation of methods for predictmg the seismic response of the BWR plants. Preliminary tests have begun behavior of nuclear plant systems.

at the shake table of Tadotsu Engineering Laboratory and will continue in 1995. Tests will be conducted at severallevels of seismic excitation.

The Large-Scale Seismic Test (LSST) facility si using energy absorber supports for the pipmg one of the largest in the world for soil / structure systems. The NRC in this collaborative effort will interaction (SSI) research. %c construction of a carry out pre-and post-test analyses to assess the 1/4-scale model of a remforced concrete contam-applicability of currently available analytical ment,10.5 meters m diameter and 16.5 meters models. Data are also being obtained from high (11.1 meters above the ground), was com-NUPEC for seismic proving tests of a computer pleted in March 1993 and a formal dedication system and a reactor shutdown cooling system.

ceremony was held in Huahen, Taiwan, m April 1993.

6.2.3.4 Northridge Earthquake The LSST program was initiated in January 1990, On January 17,1994, a magnitude 6.7 earthquake and it is expected that it will continue for 5 years.

occurred in the San Fernando Valley near the The goal of this program is to collect real town of Northridge, California. This is the same earthquake-induced SSI data in order to evaluate general area affected by the magnitude 6.5 San 6-5 NUREG-1266 e

6. Safety Issue Resolution Fernando earthquake in 1971. Representatives safety issues, including TMI-related issues. The from the NRC Offices of Nuclear Regulatory list was to be based on the potential safety Research and Nuclear Reactor Regulation and significance and cost of implementation of each from an NRC contractor, Lawrence Livermore issue. In December 1983, the original listing and National Laboratory, toured the damaged area.

procedures were approved by the Commission.

The NRC, other government agencies, and the This guidance is reflected in the NRC Policy and nuclear industry continue to study the effects of Planning Guidance, the NRC Strategic Plan, and such earthquakes to improve knowledge of the the NRC Five-Year Plan.

causes, frequency, and severity of earthquakes, seismic wave transmission, local site amplifica-tion, seismically caused soil failure, and perform-63.2 Program Strategy ance of structures, equipment, and piping similar to that found in nuclear power plants. Although A generic safety issue (GSI) is one that involves a there were many failures associated with this safety concern that may affect the design, con-earthquake, many structures, systems, and com.

struction, or operation of all, several, or a class of ponents, even those close to the epicentral region, reactors or facilities and may have a potential for had little significant damage and could be occu.

safety improvements and issuance of new or re-pied or were functional after the earthquake. In vised requirements or guidance. Timely resolution general, well-engineered structures and equipment of these issues is a major NRC concern. A that may have experienced ground motion far in prioritization and management process has been excess of their design remained functional.

established for identifying important issues for Components made of brittle materials, such as immediate action, for eliminating non safety-ceramic insulators and cast iron components, related or non-cost-effective and duplicate issues received significant damage consistent with that from further consideration, and for keeping the observed after other carthquakes.

Commission and the public informed of the resolution of these issnes. Strategies for this program are to provide timely prioritization of Shear Wall Ult, mate Drift Limits proposed new GSIs, climinate the backlog of 6.2.3.5 i

E $ed issues (as resources permit), and issue E - i The ultimate drift limit is defined as the lateral penod c updates on the status and progress displacement at the top of the wall relative to its toward resolution of GSIs.

base normalized by the height of the wall. A research program with the objectives of estab-lishing appropriate values of ultimate drift limits 633 Research Accomplishments in FY 1994 and obtaining the statistics to define this param-eter in a probabilistic sense was completed this 6.3.3.1 Priorities of Generic Safety Issues year. The final report NUREG/CR-6104, " Shear Wall Ultimate Drift Limits," was published in The NRC continued to use risk and cost data in March 1994. The information in this report will be implementing its methodology set out in the 1982 useful in the seismic PRA or seismic margins NRC Annual Report for determining the priority analyses done to identify seismic vulnerabilities to of GSis. In December 1983, a comprehensive list severe accidents (in compliance with Generic of the issues was published in "A Prioritization of Letter 88-20, Supplement 4).

Generic Safety Issues"(NUREG-0933), and this list has been generally updated semi-annually with supplements in June and December. The results f the NRC's continuing effort to identify, 6.3 Generic Safety Issue Resolution pnontize, and resolve GSIs will be m, eluded m, future supplements to NUREG-0933.

63.1 Statement of Problem During FY 1994, the NRC identified no new GSIs, In order to ensure the timely resolution of established priorities for three issues (Thble 6.1),

important safety concerns raised by the staff and and resolved five GSIs (Thble 6.2). 'Ihble 6.3 1

outside sources, the Commission directed the contains the schedules for resolution of the 14 NRC staff to prepare a priority list of all generic unresolved GSIs at the end of FY 1994.

NUREG-1266 6-6

1 i

6. Safety Issue Resolution Table 6.1 Generic Safety issues Prioritized in FY 1994 i

Number Title Priority 158 Performance of Power-Operated Valves Under Design liasis Conditions MEDIUM 165 Spring Actuated Safety and Relief Valve Reliability 111G11 167 Ilydrogen Storage Facility Separation LOW Table 6.2 Generic Safety Issues Resolved in FY 1994 Number Title 57 Effects of Fire Protection System Actuation on Safety-Related Equipment 106 Piping and Use of liighly Combustib!c Gases in Vital Areas I

11-6 4 Decommissioning of Nuclear Reactors I.D.5(3)

On-Line Reactor Surveillance Systems 11.11.2 Obtain 'Ibchnical Data on the Conditions Inside the TMI-2 Containment Structure Table 6.3 Generic Safety issues Scheduled for Resolution Scheduled Issue Resolution Number Title Priority Date 15 Radiation Effects on Reactor Vessel Supports IIIGli TDil 23 Reactor Coolant Pump Seal Failures 111Gli 12/95 165 Spring-Actuated Safety and Relief Valve Reliability 111G11 06/98 26 Automatic Emergency Core Cooling System Switch to MEDIUM 09/95 Recirculation 78 Monitoring of Fatigt.e'Iransient Limits for Reactor MEDIUM 06/95 Coolant System 150 Performance of Safety-Related Power-Operated MEDIUM

&l/96 Valves Under Design Basis Conditions B-17 Criteria for Safety-Related Operator Actions MEDIUM 06/95 11-55 Improve Reliability of Thrget Rock Safety Relief Valves MEDIUM 08/97 B-61 Allowable ECCS Equipment Outage Periods MEDIUM 12/94 33 Control Room liabitability NEARLY RESOLVED 06/95 145 Improve Surveillance and Startup Testing Programs NEARLY RESOLVED 04/95 155.1 More Realistic Source Tbrm Assumptions NEARLY RESOLVED 02/95 166 Adequacy of Fatigue Life of Metal Components NEARLY RESOLVED

'lllD 168 Environmental Qualification of Electrical Equipment NEARLY RESOLVED TilD 6-7 NUREG-1266

6. Safety Issue Resolution 6.33.2 Progress on GSI Resolution that rules are developed in a timely manner. In addition, RES provides support for preparation of Important information has been developed this the regulatory impact analyses (RIAs) that year on the factors contributing to bk>ckage of accompany all rulemaking through the develop-ECCS strainers in IlWR suppression pools.

ment of generic methodology and guidance.

Although this issue had been resolved in 1985 as Technical reviews of all RIAs are performed upon A-43, more recent events at llarseback Unit 2 in request. The NRC Regulatory Agenda Report and Sweden and at the Perry (Ohio) nuclear plant other management information systems associated showed that LOCA-caused fibrous insulation with rulemaking activities are maintained.

debris coupled with sludge and foreign materials in the drywell could block strainers more rapidly Needed reactor-related regulatory products, e.g.,

than previously thought. Restudy of this issue in regulations and regulatory guides, are developed.

the United States is considered prudent because Rulemaking is proposed or initiated, as appro-of the changeout of mirror insulation to fibrous priate, and complex rulemakings that span the insulation by many utilities since the A-43 technical or organizational responsibilities of resolution, which was based on the presence of several groups or that involve novel or complex mirror insulation.

questions of regulatory policy are managed.

Petitions for rulemaking are investigated.

The potential for llWR ECCS strainer blockage due to LOCA-generated debris was studied in detail using a llWR6/MK1 reference plant to 6.4.2 Program Strategy estimate the probability of occurrence and attendant impacts on net pump head suction The purpose of this program is to ensure that (NPilS) margin. The resultr, reported in nuclear reactor facilities are designed, con-NUREG/CR-6224, revealed that severe strainer structed, and operated in a safe manner. There-bh>ckage and loss of NPilS margin could occur fore, a continuing need exists to revise rules and within the first 30 minutes of a loss-of-coolant guides and to develop new ones. The strategies of accident (LOCA)if other materials or particulates this program are to (1) review the effectiveness of such as suppression pool sludge are present in LWR regulatory requirements and guidance and addition to LOCA-generated debris. Studies have make recommendations for revisions: (2) develop also been started at Alden Research Laboratory screening methodology to systematically review to develop a broader experimental data base of requirements and guidance;(3) coordinate and strainer bh>ckage, particulate transport, and review proposed changes to the IAEA safety settling phenomena under simulated post-LOCA standards;(4) develop or assist the development suppression pool turbulence conditions for a of rules and regulatory guides; and (5) continue to variety of fibrous debris and sludge configurations develop and maintain management information for broader applicability to all IlWRs. The NRC is systems for rulemaking.

also participating in an international working group sponsored by the OECD/NEA-CSNI--

6.4.3 Research Accomplishments in FY 1994 Pnncipal Workm, g Group 1-whose charter is to develop an internationally agreed-upon knowledge 6.4.3.1 Elimination of Requirements Marginal to base for assessing the reliability of ECCS Safety recirculation systems, especially as related to strainer blockage.

The NRC has m.stitutionalized an ongoing effort to ch,mmate requirements marginal to safety and reduce regulatory burden by permanently inte-6.4 Reactor Regulatory Standards grating that activity into the regulatory process.

This will satisfy the requirement for a periodic 6.4.1 Statement of Problem review of existing regulations in Section 5 of Executive Order 12866, " Regulatory Planmng and RES has the primary responsibility to manage, Review." The regulatory improvement (RI) coordinate reviews of, and control all NRC program is intended to implement the principle reactor-related rulemaking activities and to adopted by the Commission that all regulatory monitor scheduling of such rulemaking to ensure burdens must be justified and that NRC's NUREG-1266 6-8

6. Safety Issue Resolution regulatory process must be efficient. The reasons The proposed revision would offer an alternative for seeking to remove regulations and license beyond the minimum threshold information conditions marginal to safety are to eliminate or required by the current 10 CFR 2.802(c) to modify requirements where burdens are not encourage any petitioner to submit more detailed commensurate with their safety significance and information and analyses to support the petition.

thus to free up licensee and NRC resources and This information would be of the same type as improve the focus and effectiveness of the body of that currently required to be developed by the regulations. The activities in this program should NRC staff for rulemaking. The revision is result in enhanced regulatory focus in safety-expected to expand the use of the petition process significant areas. As a result, an overall net by reducing or eliminating requirements that increase in safety is expected from the program.

impose a regulatory burden with no commen-Specific policies, framework for rulemakings, and surate safety benefit and to result in faster procedures for the program have been instituted.

disposition of the petitions, as well as more efficient use of NRC staff and industry resources.

As a major action for the RI program, the NRC will propose a revision to its regulations in 6.4.3.2 Other Rulemaking Appendix J to 10 CFR Part 50 concerning con-The Commission issued a final rule on Febru-tamment leakage testmg. Consistent with the ary 9,1994 (59 FR 5934) on requalification pohcies and framework established for the RI requirements for licensed operators for renewal of program, the proposed rule is formulated to adopt licenses (10 CFR Part 55). The amendment deletes performance-oriented and risk-based approaches, is less prescriptive, and allows licensees flexibility the requirement that each licensed operator pass a comprehensive requalification written exam-for cost-effective implementation of the safety objectives in the regulation. The revision would ination and an operating test conducted by the NRC during the term of the operator's 6-year permit greater intervals between required tests, license as a prerequisite for license renewal.

provided that satisfactory performance is.

Forty-two comments were received, the majority actueved on precedmg tests. The nuclear industry of which supported the proposed amendments.

has supported this activity through the collection of data at nuclear power plants and has devehyped The Commission issued a proposed rule for a guideline for implementation of the rule. This public comment on October 24,1994 (59 FR rule revismn is expected to result m greater focus 53372) on procurement of commercial-grade items on safety-sigmficant activities and a sigmficant by nuclear power plant licensees (10 CFR Part burden reduction to the industry.

21). It is expected that the final rule will be published in FY 1995. The proposed amend-The NRC has also initiated action and studies for ments would clarify and add flexibility to the revising its regulations for fire protection of power process of procuring commercial-grade items for reactors under Appendix R to 10 CFR Part 50.

safety-related service by nuclear power plant The NRC is currently conducting a review of licensees. The proposed rule responds to a initiatives for performance-oriented fire protection petition for rulemaking (PRM-214)2) submitted regulation in other industries in the United States by the Nuclear Management and Resources and abroad and in the nuclear industry in other Council (NUMARC), which is now incorporated countries. The NRC is also developing the appli-into the Nuclear Energy Institute (NEI).

cation of PRA for determining the significance of fire protection features and enhanced focus on The Commission issued a proposed rule for fire protection design activities. The nuclear public comment on November 2,1994 (59 FR industry is playing a major role in this rule 54843) on reduction of reporting requirements revision and is expected to submit a petition for imposed on NRC licensees (10 CFR Parts 50,55, rulemaking to adopt performance-oriented and 73). The amendments would reduce reporting approaches in the fire protection area.

requirements currently imposed on water-cooled nuclear power reactor, research and test reactor, The NRC is also in the process of revising 10 and nuclear material licensees. This action would CFR 2.802, " Petition for Rulemaking," to imple-reduce the regulatory burden on NRC licensees ment another aspect of regulatory improvement.

and implements an NRC initiative to review its 6-9 NUREG-1266

l

6. Sity Issue Resolution i

1 current regulations with the intent to revise or promotes preparation of high-quality RIAs, and climinate duplicative or unnecessary reporting implements the policies of the guidelines. During requirements. It is expected that the final rule-this report period, the guidelines were revised in making will be issued late in FY 1995.

response to public comments, and the handbook was modified as a result of internal NRC reviews.

The Commission issued an advance notice of In addition, the NRC continued its re-evaluation proposed rulemaking (ANPR) on November 3, of the current $1000 per person-rem conversion 1993 (58 FR 58664) on standard design certifi-factor, which is integral to the value-impact I

cation for evolutionary light water reactors assessment portion of the RIA. This paper has 1

(10 CFR Part 52). The ANPR requested public been subject to a number of NRC internal comment on the form and content of rules that reviews.

would certify these designs. The Commission anticipates that two applications for design Also to aid analysts in preparing RIAs, the NRC certification may be ready for such rulemakings in Published NUREG/CR-6080," Replacement FY 1995. An applicant for a combined license Energy, Capacity, and Reliability Costs for under 10 CFR Part 52 can use these certified Permanent Nuclear Reactor Shutdowns," and designs without further indepth review by the NUREG/CR-5344, " Replacement Energy Cost NRC.

Analysis Package (RECAP): User's Guide." The cost estimates available from these studies allow The Commiss;on in SECY-94-042 approved the NRC to estimate the costs associated with the withdrawal of six NRC policy statements that have temporary shutdown of a nuclear power reactor in been superseded by subsequent NRC rulemaking order to make safety modifications or its actions. The decision for their withdrawal would permanent loss due to an accident.

not change reporting requirernents on licensees or m any way reduce the protection of the pubbe Durin8 this report Period, the development or health and safety. The policy statements to be review of about 18 safety-related RIAs was e

pleted or imtiated to justify specific regula-withdrawn are:(1) Nuclear Power Plant Access Authorization Program, March 9,1988 (53 FR tory actions for reactor and nonreactor heensees.

7534):(2) Training and Qualification of Nuclear Power Plant Personnel, March 20,1985 (50 FR 6.5 Radiation Protection and IIcalth 11147);(3) Fitness-for-Duty of Nuclear Power EITects Plant Personnel, August 4,1986 (51 FR 27921);

(4) Maintenance of Nuclear Power Plants, 6.5.1 Statement of Problem December 8.1989 (54 FR 50611);(5)Information Flow, July 20,1982 (47 FR 31482); and (6) Plan-The NRC must provide reactor-related radiation ning Basis for Emergency Responses to Nuclear protection standards and guidance that ensure Power Reactor Accidents, October 23,1979 (44 that workers and members of the general pubh,c FR 61123). A notice of withdrawal of these policy are adequately protected from the adverse,

statements will be published in the Fedeml e nseguences of exposure to mmzmg radiation fr m hcensed activities. RES reactor-related Register early in FY 1995.

activities needed to support the program melude 6.4.3.3 Regulatory Analysis developing radiation protection standards; developmg guideh,nes for implementmg these The NRC continued its development of the regula-standards; and planning, developing, and directing tory analysis guidelines (NUREG/BR-0058, Rev.

safety research to provide the information

2) and the regulatory analysis technical evaluation necessary for licensing decisions, inspection and handbook (NUREG/BR-41184). The guidelines enforcement activities, and the standards represent the NRC's policy-setting document with development process. This includes analyzing respect to regulatory impact analyses (RIAs). The available scientific evidence to evaluate the document contains a number of policy decisions relationship between human exposure to ionizing for the preparation of RIAs performed to support radiation and radioactive material and the NRC actions affecting reactor and nonreactor potential occurrence of both late and early licensees. The accompanying handbook provides radiogenic health effects, including the radiation methodological guidance to regulatory analysts, risk to workers and the public, and estimates of NUREG-1266 6-10

6. Safety Issue Resolution the probability of increased incidence of cancer radiological protection. As a result of this new and genetic effects. These analyses are used to guidance, NRC reactor regulations and regulatory provide bases for severe accident consequence guides will have to be revised. The strategies of analysis, probabilistic risk assessment (PRA), the this program are to (1) modify radiation protec-development of safety goals and emergency plans, tion guidance and standards to be consistent with the identification of radiation protection prob-Presidential guidance on radiation protection lems, the allocation of priorities for regulatory requirements and (2) continue to monitor licensee action, and environmental impact assessments.

performance indicators by using the Radiation Recommendations of such organizations as the Exposure Information Reporting System program.

International Commission on Radiological Protection (ICRP) and the National Council on 6.5.3 Research Accomplishments in FY 1994 Radiation Protectm, n and Measurements (NCRP),

Presidential guidance to Federal agencies, con-The NRC maintains a program of research and sensus standards, licensee performance indicators, standards development in radiation protection l

cost and feasibility data, and available technical and health effects intended to ensure continued information also provide bases for developing protection of workers and members of the public regulatory and technical documents related to from radiation and radioactive materials in radiation protection for workers and the public.

connection with reactor licensed activities. The program is currently focused on improvements in health physics measurements, identification and 6.5.2 Program Strategy dissemination of cost-effective dose reduction The Commission's regulatory process requires techniques, assessing health effects consequences that safety enhancements to reactor rules and of postulated reactor accidents, and mom,tormg health effects research.

guidance be systematically screened to ensure that there is substantial increase in public protection 6531 Revision of Part 20 Radiation Protection and that based on analysis the costs are justified.

Standards Realistic values of the dollar-per-person-rem criterion are needed for analysis to justify Staff efforts to facilitate the mandatory implemen-changes, but gaps in knowledge associated with tation of the new rules continued through FY radiation health effects cause uncertainties in 1994. These efforts included development of these analyses. The strategies of this program are training courses, publication of questions and to identify and compensate for uncertainties in answers on Part 20, and publication of regulatory radiation risk coefficients used for health effect guidance. On January 1,1993, the rule became estimates in PRAs and regulatory decisions.

mandatory for all licensees. Thus, activities have l

also focused on issues raised by inspections. In l

When the Commission approved the whole body February 1994, the staff published Revision 1 to l

dosimetry accreditation rule, they directed the NUREG/CR-5569, " Health Physics Positions l

NRC staff to extend the rulemaking to include Data Base," which updated a number of positions I

extremity dosimetry. Therefore, the strategies of to correspond to the revision of Part 20. This data l

this program are to (1) improve regulatory base is also now available on diskette. Several performance for radiation protection by estab-minor corrective rulemakings were completed, and lishing measurement performance criteria and a proposed rule (10 CFR Parts 19 and 20) was accreditation programs in the areas of extremity issued in February 1994 dealing with more dosimetry, bioassay, and air sampling: (2) inves-substantial issues regarding use of " controlled tigate effective new measurement techniques for areas," the definition of occupational and public these areas; (3) establish the data base required exposure, and training requirements (59 FR 5132).

for regulations; and (4) monitor specific indicators to detect improving and declining licensee In February 1994, the staff published performance.

NUREG/CR-6112. " Impact of Reduced Dose i

Limits on NRC Licensed Activities: Major Issues Federal guidance was approved by the President in the Implementation of ICRP/NCRP Dose Limit on occupational radiation protection. Further, the Recommendations," as a di aft report for ICRP has published new recommendations for comment. A critical or wng issue has been how 6-11 NUREG-1266

6. Safety Issue Resolution the agency should respond to the recent recom-6.533 Occupational Exposure Data System mendations of the ICRP on occupational dose limits. De report provided the information The NRC continued to collect and process data in currently available to assess impacts of several the computerized data system called the Radia-alternative approaches, tion Exposure Information Reporting System (REIRS). REIRS provides a permanent record of worker exposures for reactors and several other In an ongoing effort to reduce regulatory burdens categories of licensees. A report on 1992 expo-where such reductions would not reduce health sures, " Occupational Radiation Exposure at and safety, the staff published a proposed rule Commercial Nuclear Power Reactors and Other (10 CFR Part 20)in September 1993 on frequency Facilities 1992," was issued (NUREG-0713, of medical examinations for the use of respiratory Volume 14; December 1993). Compilation of the protection equipment. The proposed rule would statistical reports indicated that approximately remove the requirement for an annual medical 200,000 individuals were monitored and half examination and allow for alternative timeframes received a measurable dose. The average measur-based upon the determination of a physician. He able dose dropped from 031 rem (cSv) in 1990 to rulemaking comments will be considered and final 030 in 1992. The collective dose obtained from actior taken in FY 1995.

summing all the individual doses was 32,000 person-rems (person-cSv). He data base also includes exposure data on individuals who have 6.53.2 f3rookhaven National Laboratory terminated employment with certain licensees.

ALARA Center Data on some 687,000 persons are in the system, most of whom worked at nuclear power plants.

The Ilrookhaven National Laboratory (BNL)

NRC continued to respond to requests for indi-ALARA Center, funded by the NRC, continued vidual exposure data from the system. The data als assist in the exammation of the doses its surveillance and dissemination of DOE and industry dose reduction and ALARA research.

incurred by transient workers as they move from IINL continued work that abstracts national and plant to plant.

international publications discussing dose reduc-tion in areas such as plant chemistry, stress In September 1994, the staff published Generic corrosion cracking, steam generator repair and Letter 94-04," Voluntary Reporting of Additional replacement, robotics, and decontamination. In Occupational Radiation Exposure Data," as a May 1994, Volume 5 of NUREG/CR-4409, " Data mechanism to complete the data available in the Ilase on Dose Reduction Research Projects for REIRS data system on occupational exposure.

Nuclear Power Plants," was published. He report With the revision to Part 20, licensees are required provides a summary of projects that have been to submit only data on the present year's activi-completed or are currently under way to reduce ties. Previously data were collected at the time a doses. This information is particularly important person terminated employment. Thus, in order to to power reactor facilities in the planning stages complete the data base, data were requested for of activities. FINL also continued publication of persons that were employed as of January 1,1994, the newsletter, "ALARA Notes," on about a that were not already covered by termination quarterly schedule. In 1994, BNL focused on reports.

making the data base more easily accessible through an on-line fax system, adding information 6.53.4 National Institute of Standards from overseas contacts, and also continued Technology development of an ALARA handbook. In May 1994, BNL hosted the third ALARA international Interagency Agreement RES-93-01 between the workshop, which was well attended by representa-NRC and the National Institute of Standards and tives from the United States and other countries.

Technology (NIST) involves an ongoing study The proceedings of that conference will be pub-aimed at establishing traceability between NIST lished in FY 1995. The center provided informa-and the Pacific Northwest Laboratories (PNL) for tion and advice on dose reduction to NRC staff neutron irradiations. PNL provides the neutron and licensees.

irradiation to NIST/NVLAP as part of its duties NUREG-1266 6-12

i

6. Safety Issue Resolution as the testing laboratory for dosimeter processor the eye, and the whole body. To date, an Active accreditation run under the NVLAP.

Differential Absorption Spectrometer has been designed, developed, and tested.

6.5.3.5 Electronic Personnel Dosimeters 6.5.3.7 Spent Fuel Heat Removal PNL is presently involved in developing a set of l

performance tests and implementing procedures lhe Oak Ridge National Laboratory, funded by

)

that would permit electronic personnel dosimeters the NRC, is also contmuing to improve the data j

(EPDs) to be used in place of film or thermolumi-base in the guide for BWR and PWR fuel decay nescent dosimeters (TLDs) to establish radiation heat generation by includmg analysis of recent j

doses for radiation workers. The product of this data to provide a basis for evaluatmg the ade-effort is to be a report that could be used by the quacy of the storage system heat removal capa-NRC to evaluate EPDs until such time as an bility to limit fuel rod temperatures.

appropriate ANSI standard for EPDs becomes available. This report would be used as the basis 6.6 Small Business Innovation for a possible future certification program to Research qualify EPDs for use in radiation measurements.

l Pursuant to the Small Business Research and In I)ecember 1993, NUREG/CR-6062, " Perform-Development Enhancement Act of 1992, Public ance of Portable Radiation Survey Instruments,"

Law 102-564, the NRC supports the Small j

was published. This report evaluated the current Business Innovation Research (SBIR) program, status of performance in the portable instrument which stimulates technologicalinnovation by area and is part of an ongoing activity to examine small businesses, strengthens the role of small performance to determine whether new or modi-business in meeting Federal research and fied regulatory standards are necessary.

development needs, increases the commercial application of NRC-supported research results, 6.5.3.6 Gamma Dose Spectrometer and improves the return on investment from Federally funded tesearch for economic and social Work is being carried out under a Small Business benefits to the nation. The NRC has participated Innovative Research Phase Il contract that in the program since its inception in FY 1982, involves the development of a gamma-ray promoting high quality, " cutting-edge" research of dosimeter / spectrometer that will measure the relevance and potential importance to the NRC gamma-ray spectrum over a wide range of mission. One goal of the program is to couple this energies. From this information and the electronic research with follow-on private funding, pursuant signal retrieved from the dosimeter, it will be to possible commercial application. As of FY possible to calculate, through the use of appro-1994, the NRC was supporting 17 SBlR priate algorithms, the dose delivered to the skin, projects-in-progress.

6-13 NUREG-1266

1 i

i PART 3--SAFEGUARDS REGULATION PROGRAM

l

7. NUCLEAR MATERIALS RESEARCll 7.1 Statement of Problem the alk> cation of priorities for regulatory action, and environmental impact assessments. Recom-RES has the primary responsibility to manage, mendations of such organizations as the Intern-coordinate reviews of, and control all NRC ational Commission on Radiological Protection materials-related rulemaking activities and to (ICRP) and the National Council on Radiation monitor scheduling of such rulemaking to ensure Protection and Measurements (NCRP), Presi-that rules are developed in a timely manner. In dential guidance to Federal agencies, consensus addition, RES provides support for preparation of standards, licensee performance indicators, cost the regulatory impact analyses (RIAs) that and feasibility data, and available technical accompany all rulemaking through the develop-information also provide bases for developing ment of generic methodology and guidance.

regulatory and technical documents related to Technical reviews of all RIAs are performed upon radiation protection for workers and the public.

request. The NRC Regulatory Agenda Report and other management information systems associated with rulemaking activities are maintained.

7.2 Program Strategy The purpose of the NRC materials regulatory Needed materials-related regulatory products, e.g.,

program is to ensure that nuclear materials regulations and regulatory guides, are developed.

facilities are designed, constructed, and operated Rulemaking is proposed or initiated, as appro-in a safe manner. Therefore, a continuing need priate, and complex rulemakings that span the exists to revise rules and guides and to develop technical or organizational responsibilities of new ones. The strategies of this program are to several groups or that mvolve novel or complex (1) develop screening methodology to system-questions of regulatory poh,ey are managed.

atically review requirements and guidance; Petitions for rulemaking are investigated.

(2) coordinate and review proposed changes to the IAEA safety standards:(3) develop or assist the The NRC must provide materials-related radia-development of rules and regulatory guides; and tion protection standards and guidance that (4) continue to develop and maintain management ensure that workers and members of the general information systems for rulemaking.

public are adequately protected from the adverse consequences of exposure to ionizing radiation The Commission's regulatory process requires from licensed activities. RES materials-related that safety enhancements to materials rules and activities needed to support the program include guidance be systematically screened to ensure that developing radiation protection standards; there is substantial increase in public protection developing guidelines for implementing these and that based on analysis the costs are justified.

standards; and planning, developing, and directing Realistic values of the dollar-per-person-rem safety research to provide the information criterion are needed for analysis to justify necessary for licensing decisions, inspection and changes, but gaps in knowledge associated with enforcement activities, and the standards develop-radiation health effects cause uncertainties in ment process. This includes analyzing available these analyses. The strategies of this program are scientific evidence to evaluate the relationship to identify and compensate for uncertainties in between human exposure to ionizing radiation radiation risk coefficients used for health effect and radioactive material and the potential estimates in PRAs and regulatory decisions.

occurrence of both late and early radiogenic health effects, including the radiation risk to When the Commission approved the whole body workers and the public, and estimates of the dosimetry accreditation rule, they directed the probability of increased incidence of cancer and NRC staff to extend the rulemaking to include genetic effects. These analyses are used to provide extremity dosimetry. Therefore, the strategies of bases for severe accident consequence analysis, this program are to (.1) improve regulatory per-probabilistic risk assessment (PRA), the develop-formance for radiation protection by establishing ment of safety goals and emergency plans, the measurement performance criteria and accredita-identification of radiation protection problems, tion programs in the areas of extremity dosimetry, 7-1 NUREG-1266 l

t

~. - _. - _

7. Nuclear Materials bioassay, and air sampling:(2) investigate effec-7.3.2 Materials Regulatory Standards tive new measurement techniques for these areas; (3) establish the data base required for regula.

The Commission issued a final rule (10 CFR Parts tions; and (4) monitor specific indicators to detect 30,40,50,70, and 72) allowing self-guarantee as an additional mechanism for financial assurance improving and declining licensee performance, for decommissioning on December 29,1993 (58 FR 68726). This rulemaking is in response to a Federal guidance was approved by the President petition for rulemaking (PRM-30-59) submitted on occupational radiation protection. Further, the by the General Electric Company and Westing-ICRP has published new recommendations for house Electric Corporation. The final rule allows radiological protection. As a result of this new certain financially strong, non-electric utility guidance, NRC materials regulations and regula.

licensees to use a self-guarantee as financial tory guides will have to be revised. The strategies assurance for deco,mmissioning funding. It would of this program are to (1) modify radiation not apply to electnc utility bcensees.

protection guidance and standards to be con-sistent with Presidential guidance on radiation A final rule (10 CFR Part 73) to require a physical protection requirements and (2) continue to fitness program for security personnel at Category monitor licensee performance indicators by using I facilities was published on July 28,1994 (59 FR the Radiation Exposure Information Reporting 38347). The rule adds new requirements for a System program.

physical fitness program and annual performance l

testing or a quarterly site-specific content-based l

performance test.

7.3 Research Accomplishments in FY A proposed rule (10 CFR 72.214) adding a j

1994 standardized HUHOMS cask to the list of approved spent fuel storage casks was published for public comment on June 2,1994 (59 FR 28496). The rule will increase the number of NRC-7.3.1 Maten. ls I,icensee Performance a

certified spent fuel storage casks from which the i

holders of power reactor operating licenses can l

Through its human factors regulatory research choose to store spent fuel under a general license.

}

program, the NRC seeks to improve its under.

It is expected that the final rule will be issued standing and to maintain its requirements early in FY 1995.

I concerning the effect of human performance on the safety procedures involving the medical and A proposed rule (10 CFR Parts 30,32, and 35) on industrial use of nuclear materials, the medical use of byproduct material was pub-l lished for public comment in July 1993 (58 FR 33396). This action was taken in response to a Reports are being prepared on the results of petition for rulemaking (PRM-35-9). He final comprehensive human factor evaluations of the rulemaking sent for EDO/ Commission approval teletherapy and remote after-loading brachy-in September 1994 is intended to provide greater therapy systems. The first volume for each set of flexibility by allowing properly qualified nuclear evaluation results includes identification of human pharmacists and authorized users who are factor problems within each system, alternative physicians more discretion to prepare radioactive approaches to solving those problems, and an drugs containing byproduct material for medical assessment of those approaches with respect to use. The proposed rule would also allow research their relative ability to solve system human factor involving human subjects using byproduct problems. The remaining volumes for each system material and the medical use of radiolabeled i

evaluation will contain support for the findings biologics. It is expected that the final rulemaking described in the first volume: specifically, the will be completed early in FY 1995.

results of job and task analyses, as well as indepth studies of human-system interface, procedures, A proposed rule (10 CFR 72.214) that would i

training, and organizational practices and policies amend the regulations to allow cask VSC-24 to for each of the systems, store spent fuel with control components in the 9

NUREG-1266 7-2

7. Nuclear Materials storage casks is being developed. The holders of being irradiated is in air in a room that is accessi-power reactor operating licenses can use approved ble to personnel when the source is shielded) and spent fuel storage casks under a general license to underwater irradiators in which the source always store spent fuel at the reactor site.

remains shielded under water and the product is irradiated under water. Draft Regulatory Guide A proposed rule (10 CFR Part 73) to update DG-0003 " Guide for the Preparation of Appli-nuclear power reactor physical protection require-cations for Licenses for Non-Self-Contained ments is being developed. It is expected that the Irradiators," was published for comment in proposed rule will be published for comment in January 1994. The guide is related to the irradi-FY 1995.

ator rulemaking and describes the information that an applicant should submit for a new license A proposed rule (10 CFR Part 70) on domestic application or renewal license application.

licensing of special nuclear materials is being developed. The proposed rewrite of Part 70 would 7.3.3.2 Sewer Disposal amend the Commission's regulations to provide performance-based rather than prescriptive-based in Februa'Y 1994, the staff published an advance regulations for special nuclear material licensees.

notice of proposed rulemaking (ANPR) on The rewrite will also develop regulations that are disposal of radioactive material by release mto granted according to risk and clarify existing samtary sewer systems (59 FR 9146). Regulations

• I0 CFR Part 20 currently permit disposal into a j

requirements. As an additional requirement, s mtary sewer of specified quantities of soluble licensees with large quantities of special nuclear material with the additional constramt of meetmg material would have their safety programs based on an integrated safety analysis.

concentratmn values m Table 3 of Appendix B to the regulation. Tins rule will also respond to a i

petition for rulemaking (PRM-20-22) submitted A petit. ion for rulemak.mg from Advanced by the Northeast Ohio Regional Sewer District.

Medical Systems, Inc. (PRM-32-3) was dem. d on The ANPR requested comments on the appropri-e April 12,1994 (59 FR 17286). The petition re.

ateness of current NRC regulations and solicited quested the Commission to amend its regulations comments on a number of possible alternative because the petitioner beheved the reqmrements approaches to the form and content of the m Part 32, which are applicable to ongmal regulations. Interest in this area continued with manufacturers and supphers, were not equally the publication of a GAO report, and the NRC applicable to manufacturers and supphers of staff is currently contracting with the Pacific replacement parts. The petition was demed Northwest Laboratories (PNL) for additional because the existmg NRC regulatmns apply information related to sewer chemistry to deter-equally to manufacturers and supphers of both mine what types of regulatory changes may be ongmal and replacement parts, thereby ensunng appropriate.

the m, tegrity of these parts.

7.3.3.3 Radiography 7.3.3 1 ate in s diation Protection and in February 1994 (59 FR 9429), the staff published a proposed rule for 10 CFR Part 34, " Licenses for 7.3.3.1 Irradiator Rulemaking Radiography and Radiation Safety Requirements for Radiographic Operations." This portion of the On February 9,1993, the NRC published (58 FR Commission's regulations covers the conduct of 7715) a final rule on " Licenses and Radiation radiography using sealed sources and has not Safety Requirements for Irradiators." The rule been the subject of a complete revision for a num-established a new Part 36 to specify radiation ber of years. The proposed rule would respond to safety requirements and licensing requirements a petition for rulemaking (PRM-34-04) submitted for the use of licensed radioactive materials in by the International Union of Operating Engi-irradiators. Irradiators use gamma radiation to neers, Local No. 2, and represented a complete irradiate products to change their characteristics revision to this part of the Commission's in some way. The safety requirements apply to regulations, including proposals for certification panoramic irradiators (those in which the material of radiographers and implementation of a 7-3 NUREG-1266

7, Nuclear Materials two-person rule for work with radioactive sources.

present regulation. The Commission plans to The proposals took into account recent regulatory analyze comments and consider final rulemaking approaches of a number of Agreement States and action in FY 1995, the Conference of Radiation Control Program 7.3.3 6 Improvement of IIcalth Effects Models Directors. Interest of the Agreement States has been significant, and the NRC staff plans to hold Revision 2, Part I to NUREG/CR-4214. " Health a workshop early in FY 1995 with the States to Effects Models for Nuclear Power Plant Accident discuss the issues and possible resolutions.

Consequences Analysis," published in October 1993, contains an introduction, integration, and In related activities, NUREG/CR-4833, "Large summary of health effects models and risk Area Self-Powered Gamma Ray Detector: Phase coefficients intended for use in severe accident II Development of a Source Position Monitor for analyses, probabilistic risk assessments, Use on Industrial Radiographic Units," was emergency response planning, and safety goal and j

published. This work resulted from a Small cost / benefit analyses. Leading to modification of Business Innovative Research contract and the models presented in NUREG/CR-4214 are examined the feasibility for a source position the reports of the United States Scientific monitor as an additional safeguard to preventing Committee on the Effects of Atomic Radiation overexposures resulting from disconnected (UNSCEAR,1988), the National Academy of sources during radiographic operations.

Sciences / National Research Council BEIR V Committee (NAS/NRC,1990), and other revised 7.3.3.4 Uranium Mill Tailings recommendations of ICRP-60 (ICRP 1991).

In November 1993, the staff published a proposed 7.3.3.7 Embryo / Fetal Dose from Maternal Intake rule (10 CFR Part 40, Appendix A) on uranium mill tailings to conform the NRC regulations to A study to improve understanding of the contri-Environmental Protection Agency (EPA) regula-bution of maternal radionuclide burdens to tions under the Clean Air Act and support prenatal radiation exposure was continued in FY recision of certain EPA Clean Air Act require-1994 with significant progress. In October 1993, ments as outlined under a memorandum of NUREG/CR-5631, Revision 1, Addendum 1, understanding and a settlement agreement

" Contribution of Maternal Radionuclide Burdens between EPA. several States, and environmental to Prenatal Radiation Doses: Relationships t

organizations. The final rule was published in Between Annual Limits on Intake and Prenatal June 1994 (59 FR 28220) and EPA published its Doses," was published. The report provides an recision at the end of June.

expansion of the methodology presented earlier by examining the relationship between published 7.3.3.5 Patient Release Criteria Annual Limits on Intake in 10 CFR Part 20 and the dose to an embryo / fetus. Research that will In June 1994 (59 FR 30724), a proposed rule was permit inclusion of additional radionuclides, such published on entena for the release of patients as technetium, molybdenum, and additional adnu,mstered radioactive matenal. At the same transuranic elements, began in FY 1993 and time, Draft Regulatory Guide DG-8015, "Releas$

continued in FY.1994. The methods and data of Patients Admmistered Radioactive Matenals,,

developed under this project have been used by was published for comment. Critena for release of the NRC in preparing Regulatory Guide 8.36, patients is currently contamed in 10 CFR 35.75,

.. Radiation Dose to Embryo / Fetus," which and is specified in terms of a quantity of matenal describes acceptable methods of compliance with (30 mci) m the patient. Tlus rulemaking action 20.1208 of 10 CFR Part 20. The guide might be addressed the requests of three petitions for revised to incorporate the information presented rulemakmg: PRM-20-20 from Dr. Carol S.

in the addendum. The methods developed under Marcus and PRM-35-10/10a from the Amen.can this project are also useful in calculating doses in College of Nuclear Medicme. The petitioners cases of accidental releases of radioactive requested that the Commission adopt a dose limit materials of 5 mSv (0.5 rem) for individuals exposed to patients who have been administered radioactive In December 1993, the NRC placed a Letter material rather than the activity limit in the Report from PNL (PNie8977). " Dose to the NUREG-1266 7-4

7. Nuclear Materials Embryo / Fetus from Selected Radiopharma-and package designs. The system is currently used ceuticals-Preliminary Recommendations," in the at ORNL in support of several tasks funded by Public Document Room in support of ongoing NMSS. In particular SCALE-4 is used by ORNL rulemaking activities related to establishing limits and NRC staff for criticality safety analyses for dose to the embryo / fetus as a result of medical relevant to licensing issues. Valid criticality safety treatments. This rulemaking effort will continue in analyses require validation of both methods FY 1995.

applied and the user who applies them. The goal of this project is to validate the Criticality Safety 7.3.3.8 Criticality and Fuel Cycle Safety Analyses Sequences within the SCALE-4 system The final Regulatory Guide 3.68, " Nuclear by analyzing a large number of benchmark critical Criticality Safety Training," was published in eXPenments whose parameters (ennchment, April 1994. This regulatory guide was developed geometry, fissile fuel / moderator ratio, etc.) cover to provide guidance to licensees on an appro-the range of interest within the NMSS Fuel Cycle priate nuclear criticality safety training program Safety Branch. The work will be documented m a for the use of special nuclear material, especially rep rt that willinclude a desenption of the the prevention of criticality accidents.

entical experiments modeled, calculational results, quantification of trends m calculated k-effectives The Los Alamos National Laboratory, funded by for different types of experiments, and recom-the NRC, continued its examination and revision mended calculational uncertainties to be applied.

of TID-7016. " Nuclear Safety Guide," for simplifi-i cation of use, evaluation against new experimental 7.3.4 Uranium Enrichment data, and use of current computational codes. The document is a standard guide and reference used in Rbruary 1994, the staff published a proposed by industry and the Office of Nuclear Material rule (10 CFR Part 76) on certification of gaseous Safety and Safeguards (NMSS) staff for initial diffusion plants (59 FR 6792) to solicit comment criticality safety evaluations.

on the standards that will be used by the NRC for certification of the operations of the gaseous The Oak P.idge National Laboratory (ORNL),

diffusion enrichment facilities leased by the U.S.

funded by the NRC, continued its methods Enrichment Corporation from the Department of validation of the criticality analytical sequences in Energy. The rule covered both the certification SCALE-4 using ENDF/B-V cross-section data.

process and the standards to be used to judge The validation effort will qualify the applicability acceptable performance for certification. Under of SCALE-4 to criticality safety problems the enabling legislation, the final rule was to be covering the range of interest within the NMSS completed by the end of October 1994. The final Fuel Cycle Safety Branch. The SCALE code rule (10 CFR Parts 19,20,21,26,51,70,71,73,74, system was developed at ORNL for criticality, 76, and 95) was published on September 23,1994 shielding, and thermal analysis of nuclear facility (59 FR 48944).

7-5 NUREG-1266

~_

8. LOW-LEVEL WASTE DISPOSAL NRC research in support of regulatory activities Disposal criteria for LLW have evolved as for low level-waste (LLW) disposal facilities is experience, knowledge, public awareness, and focused on making more realistic assessments of political controversy have grown. In particular, j

the overall performance of disposal systems. The through the low-Level Radioactive Waste Policy results of NRC LLW research are also useful to Amendments Act of 1985, the Congress has

)

the States regulating LLW disposal and are made required the NRC to provide guidance for l

available to the States through NRC-sponsored regulatory decisionmaking regarding engineered j

workshops, participation by NRC contractors in LLW disposal methods. This change has forums sponsored by other agencies, as well as the broadened the scope of NRC LLW research.

I conventional method of publication in journals.

8.2 Program Strategy NRC research in support of licensing activities for 8.1 Statement of Problem LLW disposal facilities is exammmg enhance-ments and alternatives to shallow land burial, Disposal of. LLW m.volves issues concerning waste LLW waste forms, infiltration of water, radio-form and waste package mtegnty, transport of nuclide migration in the soil, hydrology and radionuclides through the disposal facility contaminant transport, performance assessment, environment, and evaluation of long-term doses and LLW source term modeling. The NRC's LLW from releases of radionuclides beyond the dis-research staff also prepares rulemakings that posal facility environment. Research is required to affect LLW disposal.

establish regulatory critena and license apphea-tion assessment information to permit sound The diverse LLW regulatory user community evaluation of proposals for disposal facilities and makes the coordination and definition of LLW to ensure that all regulatory requirements, par-research and the dissemination of associated ticularly those on radionuclide release limits, will products a much more complicated undertaking be met. Performmg the needed research m a than similar activities for the high level waste timely manner is made more urgent and complex program. Because many States are licensers of by two factors. First, the Low-Level Radioactive LLW disposal and are k>oking to the NRC for Waste Policy Amendments Act of 1985 (P.L technical support in their LLW licensing and 99-240) sets a very tight schedule for establishing regulatory programs, NRC's LLW research has to facilities within individual States or compacts of be more prescriptive and developmental than the

{

States.

IILW research program.

Second, the States and compacts of States have 8.3 Research Accomplishments in FY chosen to consider alternative disposal methods to 1994 shallow land burial. Certain of these alternatives must be critically examine,d by tightly focused 8.3.1 Materials and Engineering research to determine their acceptability and to give guidance to the States and compacts.

8.3.1.1 Engineered Enhancements and Alternatives to Shallow Land Burial The direction of the LLW research program has Many States and State compacts are considering responded to legislative action, the changing engineered enhancements for the disposal of LLW.

policy of States now responsible for disposal, and Several concepts have been proposed-the lessons learned from the history of shallow particularly the use of concrete engineered land burial of wastes at a number of sites for barriers to contain LLW. NRC research conducted several decades. Vague and differing criteria as to at the National Institute of Standards and Tech-site suitability, waste package design, etc., have nology (NIST) has investigated the durability of been emphiyed and may characterize future concrete while the Idaho National Engineering efforts.

Laboratory (INEL) completed their evaluation of 8-1 NUREG-1266

8. Low-Level Waste Disposal concrete barriers in limiting radionuclide trans-missioning Management Plan), UMTRA port (NUREG/CR4070). Three reports of the (Uranium Mill Tailings Remedial Action), and NIST work are being prepared as NUREG hazardous waste sites as well. Two designs are documents that address (1) a new method to proving to be particularly effective. One, called determine chloride diffusion coefficients in bioengineering water management, not only concrete (2) the evaluation of stress-induced reduced water infiltration to a negligible amount microcracks on solute transport through concrete, but also dewatered the cells to which it was and (3) the evaluation of the effects of stresses applied lience this cover lends itself to use as a caused by sulfate attack in concrete. NIST also remedial action cover for sites susceptible to has completed a computer program for modeling subsidence. The New York State Energy Research the degradation of concrete for LLW performance and Development Administration finished con-assessment applications. The model incorporates struction in 1993 of a bioengineering water synergistic degradation mechanisms, the effects of management cover over such a trench at the West cracks and joints, and the precipitation of Wiley LLW disposal facility. A second promising concrete dissolution products to predict concrete cover consists of a conductive layer barrier placed hydraulic properties.

below a resistive layer barrier. This cover has functioned perfectly since its installation in 8.3.1.2 LLW Waste Forms January 1990.

Research conducted at INEL on the stability of PNL has developed an infiltration evaluation nuc! car reactor decontamination waste was com-methodology (NUREG/CR-5523) and has pleted. These studies were aimed at determining separately modeled infiltration and moisture radionuclide and chelating agent releases, as well redistribution using a field experiment data set as the compressive strength of the cement solidi-(NUREG/CR-5998). Various infiltration fied waste. Results have been published as estimation approaches have also been examined NUREG/CR reports; test results are also being by PNL (NUREG/CR-6114). Future work will summarized in papers that will be published in focus on applying the infiltration evaluation scientific literature. Field lysimeter studies methodology m an arid site using data from a containing radioactive ion-exchange resins related cooperative research study with the U.S.

solidified in cement and vinyl ester-styrene are Geological Survey.

being conducted at the Oak Ridge and Argonne National Laboratories to determine radionuclide release rates under environmental conditions.

8.3.2 Hydrology and Geochemistry Studies are being completed at INEL to investi-8.3.2.1 Radionuclide Migration in Soil gate biodegradation of LLW by microorganisms to ensure that the stability requirements of 10 Current models of radionuclide retardation in CFR Part 61 are met. Studies at the Pacific soils introduce sigmficant conservatisms into Northwest Laboratories (PNL) to determine current assessments of performance of LLW scaling factors for assessing hard-to-measure disposal. This conservatism is necessitated by the radionuclides in LLW are continuing. Also con.

quantitative uncertainty as to the degree of tinuing at PNL are studies to determine the effect retardation m vanous soil types under various of naturally occurring radionuclide-chelating conditions. To reduce this uncertainty, and hence complexes in soils on radionuclide transport.

permit more realistic assessment of actual expected performance of an LLW disposal facility, 8.3.1.3 Infiltration of Water the NRC is developing more rudistic retardation models based on filed observations and laboratory The University of California at Berkeley,in experiments. Of particular interest is the role in cooperation with the University of Maryland,is radionuclide transport played by naturally continuing to field test a variety of covers for produced organic complexants. Observations LLW disposal at the Maryland Agricultural made by PNL at the Hanford, Wash., site Experiment Station in Beltsville, Maryland. These (NUREG/CR-3712 and -4030) and at Chalk River covers are not only applicable to any LLW dis-Nuclear Laboratory in Canada posal method that includes an earthen cover, but (NUREG/CR-4879, Vols.1 and 2) found radio-are applicable to LLW, SDMP (Site Decom-nuclide transport through soils at rates faster than NUREG-1266 8-2

8. Low-Level Waste Disposal predicted by current transport models. This improve the quality and uniformity of information includes radionuclides (e.g., Fe-55, Co-60, Ni-63, regarding actual quantities and characteristics of Pu, and Am) generally considered unlikely candi-LLW disposed at LLW disposal facilities through dates for mobilization based on their presently the use of standardized NRC forms when the i

understood geochemical behavior. Preliminary waste is shipped. In turn, the more accurate and evidence suggests that naturally produced organic complete information on what is actually received l

complexes and microparticulates played a signifi-at a disposal facility will facilitate more realistic cant role in enhancing migration.

assessments of expected disposal facility perform-ance. It is expected that the final rule will be 8.3.2.2 Hydrology and Contaminant Transport published in the first quarter of calendar year i

PNL has evaluated and developed a data set from an earlier field study involving subsurface injec-tion of radioactive tracers in heterogeneous 8.3.5 Environmental Policy and unsaturated porous media at the llanford site.

Decommissioning The data sets reported in NUREG/CR-5996 cover 8.3.5.1 Decommissioning Cost Reassessment a period of 10 years and will allow confirmatory In October 1993, Volumes 1 and 2 of analyses of existmg flow and transport models that may be used in LLW performance assess-NUREG/CR-5884, " Revised Analysis of Decom-ment. Work has been completed by the Massa-missioning for the Reference Pressurized Water chusetts Institute of Technology and I,rinceton Reactor Power Station," were published for public comment. This analysis is the first of two Umversity on the application of stochastic.

methods for simulatmg flow and transport in documents resulting from an ongoing reassess-heterogeneous soils.

ment of decommissioning costs for commercial les power reactors, using the experience gained in the last 20 years and information 8.3J Compliance, Assessment, and Modeling available on costs of transport and disposal of waste materials. The draft report indicated that 8.3.3.1 Performance Assessment the waste disposal component could be significant Research is continuing work to develop a realistic depending on the waste site assumed. In support and computationally tractable performance f the revised analysis, the staff published assessment methodology. The current capabilities NUREG/CR-6054, "Estimatmg Pressunzed Water and limitations of performance assessment Reactor Decommissionmg Costs, in October models have been evaluated by the Sandia 1993. He report contams the computer program National Laboratories, and the results have been developed by PNL for doing cost assessments.

published in NUREG/CR-5927, Volume 1, which Similar work is under way for BWR facilities, and deals with modeling approaches, and Volume 2, the reports will be published early m FY 1995.

which deals with validation needs.

8.3.5.2 Radiological Criteria for Decommissioning 8.3.3.2 LLW Source Term Modeling ne NRC continued in FY 1994 with its enhanced participatory rulemaking approach for establish-Dur.mg FY 1994, extensions were made to the ing radiological criteria for decommissionir.g. In existing LLW source term code, BLT (breach, January 1994, NUREG/CR-6156, " Summary of leach, and transport), developed by the Brook-Comments Received from Workshops on Radio-haven National Laboratory, to incorporate logical Criteria for Decommissioning," was additional geochemistry and gaseous release. The published to provide the comments received code is currently being tested and documented.

during the seven workshops held across the country on the issues and possible approaches to 8.3.4 Low Level-Waste Regulatory Standards the rulemaking. In February 1994, the NRC staff published a draft of the rulemaking and support A proposed rule to amend 10 CFR Parts 20 and statement for public comment. Numerous com-61 to revise low-level-waste shipment manifest ments were received on the draft, and these were information and reporting was published for used to prepare a formal proposed rulemaking comment in April 1992. This rule is intended to package for Commission consideration. The 8-3 NUREG-12f6

8 low-Level Waste Disposal proposed rule was published for public comment 8.3.5.4 Timeliness on August,22,1994 (59 FR 43200). The comment In July 1994 the final rule (10 CFR Parts 2. 30,40, period expires m late December 1994, and the 70, and 72) on timeliness in decommissioning of staff anticipates holding a special workshop on materials facilities was published (59 FR 36026).

issues of pubh,c participation and the use of site-The rule amended the Commission's regulations specific advisory boards during the comment to establish timeliness criteria for decommission-E""

ing nuclear sites or separate buildings or areas following permanent cessation of licensed In support of the proposed rule, a number of activities. The principal effect of these amend-additional documents were published, including ments is to formalize and codify the NRC's NUREG-14%, " Generic Environmental Impact requirements for timeliness in decommissioning of Statement in Support of Rulemaking on Radio-materials facilities.

logical Criteria for Decommissioning of NRC-Licensed Nuclear Facilities"; NUREG-1500, 8.3.5.5 Safety Issues Related to Permanently

" Working Draft Regulatory Guide on Release Shutdown Reactors Criteria for Decommissioning: NRC Staff's Draft Brookhaven National Laboratory continued its for Comment,; and NUREG-1501,, Background determination of technical and safety criteria that as a Residual Radioactivity Criterion for should remain as part of decommissioning Decommissioning.,

regulations under 10 CFR Part 50 when a licensee r

initiates action to permanently shut down the 8.3.5.3 Decommissioning Funding nuclear reactor in preparation for decommission-ing activities. This project will develop a com-In June 1994, the Commission published a parison of the safety requirements for a shutdown proposed rule (10 CFR Parts 30,40,70, and 72) versus an operating nuclear power reactor after i

on clarification of decommissioning funding the reactor has permanently shut down. It will r

requirements (59 FR 32158). The proposed rule also perform financial assurance analysis for was intended to clarify when decommissioning offsite liability requirements for shutdown funding assurance was required and to provide reactors. it will examine the environmental impact that assurance would be available after operations of the potentialincrease in the spent fuel trans-were terminated and decommissioning initiated.

port and radiological exposure to the public in the i

The staff will analyze comments and prepare a event the licensees prefer to ship and store their i

final rulemaking package during FY 1995.

spent fuel.

?

l NUREG-1266 8-4 P

l PART 4--ASSESSING TIIE SAFETY OF IIIGII-LEVEL WASTE DISPOSAL

=

1 i

9. HIGH-LEVEL WASTE RESEARCH The Nuclear Waste Policy Act of 1982 requires the 9.1 Statement of Problem Department of Energy (DOE) to dispose of high-level radioactive waste (IILW), which can be spent The liLW disposal policy for the United States is reactor fuel or the byproduct of reprocessing defined by the Atomic Energy Act, the Energy spent fuel, in a deep geologic repository. The act Reorganization Act, the Nuclear Waste Policy Act, further requires the DOE to apply for a license and the Nuclear Waste Policy Amendments Act from the NRC to dispose of HLW.

(NWPAA). The last, signed into law in 1987, provides for the development of a geologic repository for the permanent disposal of high-level ne NRC maintains an active 11LW research pro-radioactive waste m the State of Nevada at Yucca l

l gram of theoretical study and laboratory and field Mountain and assigns responsibility for repository experiments directed at understanding the physi-development to the DOE. According to the cal processes that control and determine reposi-Federal Government's Reorganization Plan No. 3 tory performance in the unsaturated volcanic tuff of 1970, HLW environmental standards develop-at the Yucca Mountain (Nev.) site currently under ment is the responsibility of the Environmental consideration by the DOE as directed by the Protection Agency (EPA), and the Energy Reorg-Congress in December 1987. The goal of the anization Act' assigns the regulation of HLW NRC's HLW research is to provide models.

disposal to protect public health and safety and methods, data, and technical information to the environment to the NRC.

support the staff's independent judgments as to the appropriateness and adequacy of DOE's An HLW repository poses problems involving demonstration of compliance of the HLW reposi-regulatory considerations and uncertainties tory with NRC requirements specified in 10 CFR related to waste emplacement, monitoring, and Part 60 and with the Environmental Protection performance assessment that are unique in the Agency's llLW standard, incorporated by history of the NRC. Much of this um,queness reference into Part 60. The program is divided stems from the type of facility, first-of-its-kind i

into three parts: engineered systems research, geologic disposal installation, its very long which examines issues related to controlled performance time (specified as 10,000 years by the release of radionuclides, containment of waste, EPA), and the fact that it will be placed in low and the engineering-geology interface in the permeability / low flow geologic systems that have repository; geologic systems research, which not been investigated previously because of their examines issues related to the hydrology, low economic value. The NRC must have an geochemistry, and geology of the repository site; independent capability to evaluate the DOE and performance assessmeret research, which safety analyses and decide whether long-term i

integrates mathematical models from the other releases predicted by DOE will be within research into NRC's IILW performance assess-established limits. The NRC research program ment methodology. Key technical issues being objective is to provide the technical capability addressed include methods to assess the long-necessary to evaluate DOE's site characterization term performance of the packages containing the activities as required by the NWPAA and to HLW, the potential for volcanic and seismic assess DOE's license application when it is events, and flow and transport mechanisms in submitted.

unsaturated fractured rocks.

9,2 Program Strategy Most NRC HLW research is conducted by the I

Center for Nuclear Waste Regulatory Analyses The research program has been guided by the (CNWRA), a division of the Southwest Research need to provide the technical foundation for NRC Institute in San Antonio, Texas. However, a development of a set of regulations and a significant portion of NRC HLW research on licensing process for the review and licensing of hydrology is being conducted at the University of the HLW repository. 'Ihis framework for NRC Arizona.

review will allow the formal licensing activities 9-1 NUREG-1266

9. High level Waste Research and the supporting research to be focused on the 9.3.1.2 Containment significant technical issues.

10 CFR Part 60 contains a criterion for the mini-mum lifetime of HLW containment within waste At present, the NRC has active research programs packages placed in the repository. CNWRA is in hydrology, geology, materials science, geo-conducting confirmatory research on the behavior chemistry, and several other disciplines related to of waste package materials in the expected reposi-HLW management. The research combines tory environment. During FY 1994, research was theoretical study with laboratory and field done on stress-corrosion cracking, repassivation experiments to identify and quantify the physical potentials for long-term corrosion of stainless processes and phenomena important to waste steel, corrosion of copper-based waste package isolation so that the NRC can assess repository container materials, effects of surface conditions performance and quantify the uncertainties on the corrosion of waste package container associated with characterization and measurement materials, and crevice corrosion of stainless steel.

of these processes. All this work is integrated into Work also was initiated on microbial corrosion of an independent HLW performance assessment waste package container materials.

methodology. Effort is also required to validate many of the models that underlie the method-9.3.1.3 Engineering-Geology Interface ology. The ultimate goal of the NRC's HlW 10 CFR Part 60 requires that the repository's research program is to provide the techmcal bas.is engineered and geologic systems function together to support the b,eensmg staff's mdependent review so as not to compromise repository safety.

of the appropriateness and adequacy of the CNWRA has been conducting two projects on DOE s demonstration of comphance with 10 CFR coupled processes deriving from the engineered Part 60 and the EPA's HLW standard. In additmn, system's interaction with its surrounding geologic NRC's waste management research seeks to system. One project, on the redistribution of provide techmcal support to the bcensmg staff in liquid water by emplaced HLW, is using their mteractions with DOE, the State of Nevada, laboratory-based similitude experiments and and other participants and interested parties and theoretical simulations to assess models of this to develop regulatory standards to support the redistribution. Work in FY 1994 produced a heensmg of the disposal and management of simplified thermosyphon model of the redistri-high-level radioactive wastes.

bution process and examined pressure-driven heat flows in unsaturated media. in the other project, on rock-mechanical aspects of repository 9.3 Research Accomplishments in FY performance, CNWRA researchers finished a 199.g study of the effect of mine scismicity on ground-water hydrology; finished research on rock-joint characteristics; issued an evaluation of 9.3.1 Engineered Systems Research the rock mechanics simulator UDEC; and supported NRC's continued participation in 9.3.1.1 Controlled Release DECOVALEX-an international cooperative effort to test the validity of mathematical models 10 CFR Part 60 contains a criterion for the of thermal-hydrological-mechanicalinteractions.

maximum rate of release of radioactive material NRC also provided financial support to the from the repository's engineered barrier system.

Swedish Nuclear Power Inspectorate for admimstration of DECOVALEX.

Research on controlled release is being done at the natural analogue site at Pena Blanca, Mexico, by CNWRA. This site is located in an unsaturated 9.3.2 Geologic Systems Research tuff environment similar to that at Yucca Moun-9.3.2.1 Hydrology i

tam. A uramum ore body is serving as a surrogate for disposed spent fuel, and limits on the expected Because transport by ground water is considered range of spent fuel behavior in oxidizing chemical to be the most likely path for radionuclide trans-environments like those of Yucca Mountain are port from an HLW facility to the accessible being developed.

environment, the NRC is actively studying NUREG-1266 9-2

9. High-Level Waste Research i

ground-water infiltration, recharge, flow, and developed and tested in the laboratory a transport processes. At an experimental site, the double-layer surface complexation model that Apache Leap'Ibff site operated by the University meets both objectives.

of Arizona,in partially saturated fractured rock' similar to that at the Yucca Mountain site, In a workshop organized and conducted by research continued in FY 1994 on testing CNWRA researchers, ways in which natural and hydrologic site characterization methods and on archaeological analogues can be used to build scale effects in fluid flow and radionuclide confidence in the conceptual and mathematical transport in unsaturated media. Results from models used in HLW performance assessment theoretical work conc:ucted at both the CNWRA were addressed.

and the University of Arizona suggested that 93.23 Geology scalmg of certam aspects of permeability measurements may be universal and not site CNWRA has two projects that are investigating specific as previously believed. In FY 1994, the (1) techniques to estimate the likelihood of CNWRA also completed a project on stochastic occurrence of volcanos in the Yucca Mountain analysis of large-scale flow and transport in area that are alternative to the method currently unsaturated fractured rock masses. The project used by the DOE, and (2) possible consequences developed an efficient method for estimating to HLW disposal of a volcano at Yucca Mountain.

effective permeabilities measured in unsaturated During FY 1994, CNWRA found that other fractured media and developed a methodology for methods may suggest a higher likelihood of a probabilistic estimation of ground-water travel volcano at Yucca Mountain than the method i

time. CNWRA is continuing to study hydrology currently used by the DOE. The application of on a regional scale as well as a local scale.

seismic tomographic methods to provide insights as to the possible consequences of basaltic 93.2.2 Geochemistry volcanism-the type that most likely would occur at Yucca Mountain-also was examined.

Knowledge and application of the geochemical conditions at Yucca Mountain are important t W Mo M

Red understanding many aspects of repository performance, including waste package corrosion.

The NRC will assess DOE's demonstration of radionuclide release and transport, and alteration compliance with both the NRC's requirements for of ground-water flow paths. During FY 1994, HLW disposal given in 10 CFR Part 60 and EIWs CNWRA finished a project on geochemical effects HLW standard. Use of a performance assessment on mass transport in unsaturated media. In its methodology, independent of DOE's performance final phases, the project examined the thermo-assessment methodology,is a key element in dynamics of ion exchange in the zeolite mineral NRC's strategy to review that demonstration of clinoptilolite, common in tuffs like those of Yucca compliance. lo support implementation of that Mountain. This mineral is expected to play a key strategy, research is being conducted at the role in controlling radionuclide transport in the CNWRA on the development of performance Yucca Mountain repository.

assessment tools. The tools are being used in their current state of development in the joint NRC-A significant problem with addressing the CNWRA HLW lterative Performance Assessment gcochemistry of radionuclide transport is that the (IPA) effort, which is providing insights as to the complexity of the chemistry makes calculations processes and phenomena that may be critical to difficult and time consuming. Simplified geo-repository performance. It is anticipated that as chemical models that have been developed to the performance assessment tools become more makc transport calculations tractable oversimplify robust, the IPA effort also will assist in setting

~

the chemistry to the point that even so-called priorities for future HLW research.

bounding calculations may not be truly bounding.

For this reason, the NRC asked the CNWRA to In FY 1994, the CNWRA reviewed the scenario determine whether some model could be methodology used in IPA's latest exercises, developed that is sufficiently realistic to retain developed a mathematical model of infiltration credibility of the results and yet be calculationally that was applied in IPA, examined film flow in tractable. The CNWRA has subsequently fractures, performed laboratory permeability tests 9-3 NUREG-1266

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9. High-Ixvel Waste Research on tuff samples from the Pena Blanca analogue, increase the efficiency of performance assessment provided training on the flow and transport calculations with improved numerical algorithms simulator PORFLOW, and examined ways to and massively parallel computation.

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NUREG-1266 9-4

APPENDIX FY 1994 Regulatory Products from the Office of Nuclear Regulatory Research Date Regulatory Product Description l

Part 1-NUCLEAR SAFETY RESEARCll-REACTOR LICENSING SUPPORT REACTOR AGING AND RENEWAL Pressure Vessel Safety and Piping Integrity September 1994 Draft Regulatory Guide DG-1028, " Periodic Testing of Electric Power and Protection Systems."

October 1994 Regulatory Guide Revision 30 to Regulatory Guide 1.84,

" Design and Fabrication Code Case

. Acceptability-ASME Section III, Division 1."

October 1994 Regulatory Guide Revision 30 to Regulatory Guide 1.85,

" Materials Code Case Acceptability-ASME Section III, Division 1."

October 1994 Regulatory Guide Revision 11 to Regulatory Guide 1.147, " Inservice Inspection Code Case Acceptability-ASME Section XI, Division 1."

October 1994 Draft Regulatory Guide DG-1027, " Format and Content of Application for Approval for Thermal Annealing of Reactor Pressure Vessels."

October 1994 Proposed Rule Fracture toughness requirements for light-water reactor pressure vessels.

STANDARD REACTOR DESIGNS Regulatory Application of New Source Terms January 1994 SECY-94-017

" Options with Regard to Revising 10 CFR Part 100."

April 1994 EDO Memo to Commission Review of MIT Ph.D Thesis on Chernobyl Source Term July 1994 SECY-94-194

" Proposed Revisions to 10 CFR Part 100."

Part 2-NUCLEAR SAFETY RESEARCII-REACTOR REGULATION SUPPORT REACTOR ACCIDENT ANALYSIS Severe Accident Policy Implementation May 1994 SECY-94-134

" Status of the Individual Plant Examinations and the Individual Plant Examination of External Events Insights Program."

A-1 NUREG-1266

Appendix Date Regulatory Product Description SAFETY ISSUE RESOLUTION AND REGUIATION IMPROVEMENTS Earth Sciences October 1994 Proposed Rule Proposed 9100.23, " Geologic and Seismic Siting Factors," issued for public comment.

Generic Safety Issue Resolution October 1993-Generic Safety Issues For generic safety issues prioritized and September 1994 resolved in FY 1994, see Tables 6.1 and 6.2.

Reactor Regulatory Standards November 1993 Advance Notice An advance notice of proposed rulemaking, of Proposed Rulemaking 10 CFR Part 52, concerning standard design certification for evolutionary light-water reactors. The contemplated rulemaking would define the form and content of rules that would certify the designs.

December 1993 Volume 14 of A report on 1992 exposures, " Occupational NUREG-0713 Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities,1992."

It provides a compilation of the statistical reports of individual exposures.

February 1994 Final Rule The NRC regulation,10 CFR Part 55, on requahfication requirements for licensed operators for renewal of licenses. The rule deletes the requirement that each licensed operator pass a comprehensive requalification written examination and an operating test during the 6-year license term as prerequisite for license renewal.

February 1994 Proposed Rule The proposed rule,10 CFR Parts 19 and 20, regarding use of " controlled areas," the definition of occupational and public exposure, and training requirements.

September 1994 Proposed Rule The NRC regulation,10 CFR Part 20, on frequency of medical examinations for use of respiratory protection equipment. The proposed amendment would remove the requirement for an annaal medical exami-nation and allow for alternative timeframes.

NUREG-1266 A-2

l i

1 Appendix D:te Regulatory Product Description October 1994 Proposed Rule The NRC regulation,10 CFR Part 21, on procurement of commercial grade items by reactor licensees. The proposed rule j

responds to a petition for rulemaking (PRM-21-02) submitted by the Nuclear Management and Resources Council (NUMARC), which is now incorporated into the Nuclear Energy Institute (NEI). The proposed amendment would clarify and add flexibility for procuring items for safety-related service.

November 1994 Proposed Rule

'Ihe NRC regulation,10 CFR Parts 50,55, and 73, on reduction of reporting requirements imposed on NRC licensees.

l The proposed amendments would reduce reporting requirements on power reactors, research and test reactors, and nuclear material licensees.

Part 3-SAFEGUARDS REGULATION PROGRAM Nueicar Materials November 1993 Proposed Rule The NRC regulation,10 CFR Part 40, Appendix A, on uranium mill tailings. The propo.;ed rule would conform NRC regulations to Environmental Protection Agency (EPA) regulations under the Clean Air Act and support recision of certain EPA Clean Air Act requirements.

December 1993 Final Rule A final rule,10 CFR Parts 30,40,50,70, and 72, to allow self-guarantee as an additional mechanism for financial assurance for decommissioning. This rulemaking is in response to a petition for rulemaking (PRM-30-59) submitted by the General Electric Company and Westinghouse Electric Corporation. The final rule applies to certain financially strong, non-electric utility licensees and allows the use of self-guarantee as financial assurance for decommissioning funding. It would not apply to electric utility licensees.

January 1994 Draft Regulatory DG-0003," Guide for the Preparation of Guide Applications for Licenses for A-3 NUREG-1266

Appendix Date Regulatory Product Description Non-Self-Contained Irradiators." This guide is related to 10 CFR Part 36 and describes the information that an applicant should submit for a new or renewed license application.

Februaq 1994 Advance Notice An advance notice of proposed rulemaking of Proposed Rulemaking (ANPR),10 CFR Part 20, on disposal of radioactive material by release in sanitary sewer systems. The ANPR is in response to a petition for rulemaking (PRM-20-22) submitted by the Northeast Ohio Regional Sewer District.The ANPR requested comments on the appropriateness of current NRC regulations and solicited comments on possible alternative approaches.

February 1994 Proposed Rule The NRC regulation,10 CFR Part 34, on the conduct of radiography using sealed sources.

The proposed rule responds to a petition for rulemaking (PRM-34-04) submitted by the International Union of Operating Engineers, Local No. 2. The proposed rule represents a complete revision to this part of the Com-mission's regulations, including certification of radiographers and implementation of a two-person rule for work with radioactive sources.

February 1994 Proposed Rule A proposed rulemaking,10 CFR Part 76, on certification of gaseous diffusion plants. The proposed rule solicits comments on the standards for certification of the operation of gaseous diffusion enrichment facilities.

June 1994 Proposed Rule A proposed rule,10 CFR 72.214, adding a standardized HUHOMS cask to the list of approved spent fuel storage casks. The rule would increase the number of NRC-certified spent fuel storage casks available under a general license.

June 1994 Final Rule The final rule to amend NRC regulations,10 CFR Part 40, Appendix A, on uranium mill tailings. See proposed rule above (November 1993).

NUREG-1266 A-4

Appendix 1

i Date Regulatory Product Description

)

June 1994 Proposed Rule A proposed rule to amend NRC 10 CFR 35.75, on criteria for release of patients administered radioactive material.

The rulemaking action addressed the requests of three petitions for rulemaking:

PRM-20-20 from Dr. Carol S. Marcus and PRM-35-10/10a from the American College of Nuclear Medicine. He proposed amendment would specify a dose limit of 5 mSv (0.5 rem) rather than the limit of 30 mci currently specified.

June 1994 Draft Regulatory DG-8015," Release of Patients Administered Guide Radioactive Materials." The draft guide provides guidance on the proposed rule to amend 10 CFR 35.75 (see above).

July 1994 Final Rule The NRC regulations,10 CFR Part 73, on physical fitness programs for security personnel at Category I fuel cycle facilities.

1 The amendment requires physical fitness training programs as well as annual performance testing for specific security i

force personnel at facilities authorized to possess formula quantities of strategic special nuclear material.

September 1994 Final Rule The NRC regulations,10 CFR Parts 19,20, 21, 26, 51, 70, 71, 73, 74, 76, and 95, for certification of the operations of the gaseous diffusion enrichment facilities leased by the U. S. Enrichment Corporation from the Department of Energy. See proposed rule above (February 1994).

14w Level Waste Disposal June 1994 Proposed Rule The proposed rule to amend NRC regulations,10 CFR Parts 30,40,70, and 73, on clarification of decommissioning funding requirements. The amendments would clarify when decommissioning funding assurance was required and provide that assurance would be available after operations were terminated and decommissioning initiated.

A-5 NUREG-1266

-Appendix Date Regulatory Product Description July 1994 Final Rule The rule on timeliness in decommissioning of materials facilities,10 CFR Parts 2,30,40, 70, and 72. The rule establishes timeliness criteria for decommissioning nuclear sites or separate buildings or areas following permanent cessation of licensed activities.

August 1994 Proposed Rule The proposed rule,10 CFR Parts 20,30,40, 50,51,70, and 72, to amend the NRC regulations on radiological criteria for decommissioning. The proposed rule is based on comments received from seven workshops and a draft of the rulemaking published in February 1994.

August 1994 NUREG-1496

" Generic Environmental Impact Statement in Support of Rulemaking on Radiological Criteria for Decommissioning of NRC-Licensed Nuclear Facilities." This document supports the rulemaking on this topic.

August 1994 NUREG-1500

" Working Draft Regulatory Guide on Release Criteria for Decommissioning: NRC Staff's Draft for Comment." This document supports the rulemaking on this topic.

August 1994 NUREG-1501

" Background as a Residual Radioactivity Criterion for Decommissioning." This document supports the rulemaking on this topic.

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NUREG-1266 A-6

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2. TITLE AND SUBTITLE NRC Safety Research in Support of Regulation - FY 1994 3.

DATE REPORT PUBLISHED Mo#4 T M YE.R June 1995

4. FIN OR GR ANT NUMBER L

6. AUTHOR (S)

6. TYPE OF REPORT Regulatory

1. PE RIOD COVE R ED tonclusone Desess FY 1994 B.

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10. SUPPLEMENTARY NOTES I

I1. ABSTRACT (Joo.o,e or =,ss This report, the tenth 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 FY 1994.

The goal of the Office of Nuclear Regulatory Research (RES) is to ensure the avail-ability of sound technical bases for timely rulemaking and related decisions in support of NRC regulatory / licensing / inspection activities.

RES also has responsibili-ties related to the resolution of generic safety issues and to the review of licensee submittals regarding individual plant examinations.

It is the responsibility of RES to conduct the NRC's rulemaking process, including the issuance of regulatory guides and rules that govern NRC licensea activities.

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