ML20070D247
| ML20070D247 | |
| Person / Time | |
|---|---|
| Issue date: | 06/30/1994 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| NUREG-1266, NUREG-1266-V08, NUREG-1266-V8, NUDOCS 9407070294 | |
| Download: ML20070D247 (100) | |
Text
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N U REG-1266 Vol. 8
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NRC Sa ~ety Research in Supaort 0:? Regu:a~: ion FY 1993 U.S. Nuclear Regulatory Commission OITice of Nuclear Regulatory Researcli
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AVAILABILITY NOTICE Availabihty of Reference Materials Cited in NRC Publications Most occuments cited in NRC pubhcations will be available from one of the following sources:
1.
The NRC Pubhc Document Room. 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2.
The Superintendent of Documents. U.S. Government Printing Othee, Mail Stop SSOP, Washington, DC 20402-9328 3.
The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.
l Referenced documents available for inspection and copying for a fee from the NRC Pubhc Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; bcensee event reports; vendor reports and correspondence; Commission papers; and appbcant and licensee docu.
ments and correspondence, The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, international agreement reports, grant pubhcations, and NRC booklets and brochures. Also available are regulatory guides, NRC regulations in the Code of Federal Regulations, and Nu-clear Regulatory Commission Issaances.
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by the public, Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American Na-tional Standards Institute,1430 Broadway, New York, NY 10018.
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NUREG-1266 Vol.8 l
l NRC Safety Research in
~
Support of Regulation - FY 1993 Manuscript Cornpleted: April 1994 Date Published: June 1994 OITice of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
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AllSTRACT This report, the ninth in a series of annual reports, was focus on safety considerations for license renewal be-prepared in response to congressional inquiries concern.
comes timely, ing how nuclear regulatory research is used it summa-The primary purpose of performing regulatory research is rizes the accomplishments of the Office of Nuclear Itegu-to develop and provide the Commission and its staff with latory itesearch during FY 1993. A special emphasis on sound technical bases for regulatory decisions on the safe accomplishments in reclear power plant aging research operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to de-1 reflects recognitica that a number of plants are entering velop data and related information for the purpcsc of the final portion of their original 40-year operating li-revising the Commission's rules, regulatory guides, or censes and that, in addition to current aging effects, a other guidance.
i iii NUIREG-1266
Contents Page Abstract..........................................................................................
iii j
liighlighe.........................................................................................
vii Part 1-Nucletr Safety Research--Reactor Licensing Support 1.
Reactor A@g and Licensing Renewal...........................................................
1-1 1.1 Reactor Vessel &tfety and Piping Integrity....................................................
1-1 1.2 Aging of Reactor Componen ts............................................................
1-6 l
2.
S t a nda rd Reac t o r D eM an s.......................................................................
2-1 1
2.1 Engineering Issues for Advanced Reactor Designs............................................
2-1 l
2.2 Systems Performance of Advanced Reactors 2-2 2.3 Adva nced Reacto r Risk Analysis............................................................
2-4 2.4 Regulatory Application of New Source Terms.........................,...................
2-5 Part 2-Nuclear Safety Research-Reactor Regulation Support 3.
P la n t Pe rfo rm a n c e.............................................................................. 3-1 3.1 S ta t e m e n t o f Prob l e m...................................................................... 3-1 3.2 Prog ra m S t ra t e gy...................................................................... 3-1 3.3 Resea rch Acco m plishm e n ts in FY 1993......................................................
3-1 4.
I l u ma n Re l iab il i ty...........................................................................
4-1 i
4.1 Statement of Problem.....................
4-1 4.2 Prog ra m S t ra t e gy........................................................................
4-1 4.3 Resea rch Accomplish m e n ts in FY 1993......................................................
4-1 5.
React or Accid e n t Analysis......................................................................
5-1 5.1 Reactor Risk Analysis............................
5-1 5.2 Co n tain m e n t Pe rfo rm a nce...............................................................
5-2 5.3 Severe Accident Phenomenology.......
5-7 5.4 Reactor Containment Struct ural Ir.Writy................................................... 5-10 5.5 Severe Accident Policy Implementation...................................................... 5-11 6.
Safety issue Resolution and Regulation Improvements.............................................. 6-1 6.1 Ea rt h Sci e n c e s............................................................................
6-1 6.2 Plant Response to Scismic and Other External Events.........................................
.6-7 6.3 G en e rie Safe ty Issu e Resol u tion.............................................................
6-9 6.4 Reactor Regulatory Standards............................................................... 6-9 6.5 Radiation Protection and IIcalth Effects.................................................... 6-13 6.6 Small Business Innovation Research........................................................ 6-16 Part 3-Nuc! car Materials Licensing and Regulation Support 7.
Nucl ea r Ma t e rials.............................................................................
7-1 1
7.1 Materials Regulatory Research.............................................................
7-1 8.
Low-Level Wa st e D isposal..................................................................... 8-1 8.1 S ta t em e n t o f Probl em......................................................................
8-1 8.2 Prog ra m S t ra t e gy..........................................................................
8-1 8.3 Research Accomplish ments in FY 1993.......................................................
8-1 v
Contents (cont)
Page ihrt 4-Assessing the Safety of liigh. Level Waste Disposal 9-1 9.
l iigh-Lev el Waste Re searc h...................................................................
9-1 9.1 S t a t e m e n t o f Probl e m....................................................................
9-1 9.2 Progra m S t ra t e gy..........................................................................
9-1 9.3 Research Accomplish m en ts in FY 1993......................................................
APPENDIX-FY 1993 Regulatory Products from the Office of Nuclear Regulatory Research................ A-1 1hbles 6.1 Generic Safety issues Prioritized in FY 1993................................................... 6-10 6.2 Generic Safety issues Resolved in FY 1993..................................................... 6-10 6.3 Generic Safety Issues Scheduled for Resolution............................................... 6-11 6.4 Rulemaking Actions Processed During FY 1993.................................................. 6-13 l
l
)
NUREG-1266 vi 1
________m
IIIGilLIGilTS Part 1--NUCLEAR SAFETY
^ draft regulation and draft regulatory guide ad-RESEARCll-REACTOR dressing the engingering and metaHurgical aspecu of thermal anneahng for U.S. plants were devel-LICENSING SUPPORT oped.
i l
Reactor Aging and License Renewal Aging of Reactor Components j
The process is continuing in the development of in.
e Pressure Vessel Safety and Piping Integrily formation on, as well as mitigation and managing methods for, the aging process of safety-significant A draft regulatory guide was issued that would pro-components of commercial nuclear power plants.
o vide evaluation methods for demonstrating that ma-Final reports on the following subjects were issued, i
terials with low resistance to a " ductile tearing" fail.
BWR internals; PWR internals; check valves; auxil-ute mode meet the requirements of Appendix G to iary feedwater system (follow-on study); aging, con.
10 CFR Part 50 on Charpy upper-shelf energy (the dition monitoring, and loss-of-coolant accident indication of reactor vessel toughness). Generie (LOCA) tests of Class 1E electrical cables; record-analyses using the draft regulatory guide methodolo-keeping; HEPA filters and adsorbers; and essential gy demonstrated that adequate pressure vessel in-IIVAC chillers 'Ihe results will be used by the NRC tegrity exists for pressure vessels with Charpy upper-in identifying and developing policies and guidelines
)
shelf energy values well below the values specified in for making operating plant management decisions
]
Appendix G.
that may safely extend the life of the plants.
)
A draft regulatory guide was issued that would pro, The risk-based methodology for assessing aging in o
o nuclear power plants and for d among risk contributors and mam, efining prioritie vide acceptable methods for determining calcula-tenance activities tional and dosimetry methods for determining pres-sure vessel neutron fluence for power reactors. In s subject to uncertainties because of limited avail-support of this draft guide, new neutron cross, able agmg data and because of various modeling as.
section libraries were developed that apply the latest sumptions (e.g, modeling of effects of test and main.
evaluated nuclear data files.
tenance). Research m 1993 focused on developing sensitivity and uncertainty analyses to define pnort-ties addressing data and modeling uncertainties and A mechan. tically based model was developed for o
is was documented in draft NUREG/CR-6405.
predicting steam generator tube failures and leak rates of degraded tubes under normal and main steam line break conditions. This model was used by Standard Reactor Des,gns i
the NRC'lhsk Group on Interim Plugging Criteria, which was charged with evaluating industry propos-Engineering Issues for Advanced Reactor l
als on the subject.
Designs 1
A probabilistic methodology was developed to esti-o As part of NRC participat. ion m Working Group 3 mate the failure probability for low pressure piping (pressure vesselintegrity) of the Joint Coordinating subjected to reactor coolant pressures and tempera.
Committee on Civilian Nuclear Reactor Safety, an effort continued on test reactor irradiations m the tures during an intersystem loss-of-coolant accident.
United States and Russia, Russian scientists visited The method is applicable to both carbon and stain.
the United States, and a Russian scientist was ac-less steel piping and other piping compqnents, in-cepted for a 1 year assignment to Oak Ridge Nation' cluding flanges, valves, pumps, and heat exchanger al Laboratory to investigate radiation embrittiement tubes. The methodology can be used for piping re-and thermal-anneahng effects.
views for operating reactors and for Advanced Light-Water Reactor (ALWR) design certification reviews, o
Interim fatigue design curves were developed and The Oak Ridge National Laboratory is conducting e
the information provided in NUREG/CR-5999; a study to identify functional and environmental they more accurately describe fatigue life in the issues arising from the application of new technolo-high temperature water characteristics of LWR gies to the instrument and control (l&C) systems in coolant systemt Fatigue tests are in progress to vali-both present and next generation of nuclear power date and/or update the proposed design curves.
plants. Preliminary results were published in draft vii NUREG-1266
liighlights NUREG/CR-5904. IIcavy emphasis was placed on mentation of the Commission's emergency planning the applicability of military and industrial standards exercise requirements.
for the qualification of such s stems. Generally such In March 1993, a petition for proposed rulemaking standards were shown to be overly conservative as e
applied to digital I&C systems for nuclear plants.
by VEPCO was published. Proposed rulemaking would change the frequency of emergency exercises fr m annual to biennial.
Systems Performance of Advanced Reactors e
Modifications to convert the ROSA facility for Part 2--NUCLEAR SAFETY M'
testing were defined and the hardware m-RESEARCH--REACTOR REGULATION SUPPORT A contract to build a loop for confirmatory ter ting of e
the Simplified floiting Water Reactor (Sim R) was Plant Perforniance awarded to Purdue University.
In order to share international experience with A review of the data base supporting the CANDU 3 thermal-hydraulic system codes, over 14 countries e
design has been completed.
have joined the Code Applications and Maintenance Program, thereby providing $390,000 in financial i
Improvements were made to the models in the RE.
support.
e LAPS code for the AP600 and SilWR designs.
Human Reliability Advanced Reactor Risk Analysis The Human Performance Investigation Process is a The passive reliability project was continued systematic method for investigating the root causes e
throughout FY 1993, it is presently focusing on the of human performance problems during events at Westinghouse AP600 design and developed a candi-nuclear power plants. Field testing of the method date methodology for quantifying the uncertainty was completed and is being used in current event in-distribution in the core damage frequency arising vestigations. Future training on the method will be from uncertainty in the modeling of the natural pro-provided by NRC's Technical'Itaining Center.
cesses. Project documentation will be completed in
'Ihe NRC requested and received a report on the les-FY 1994, e
sons learned from the study of over 10 years of ex-perience with test and evaluation methods for Regulatory Application of New Source Terms computer-based operator support systems at the Work continued on a replacement forTID-14844. In Halden Reactor Project in Norway, e
support of this effort,fourcontract NUREG reports The NRC initiated and conducted an international were issued.
workshop on digital systems reliability and nuclear Staff efforts continued on revising 10 CFR Part 100, safety. The workshop provided feedback from ex-
" Reactor Site Criteria." In an SRM dated March 28, perts from nine different countries regarding poten-1994, the Commission approved the staff's recom.
tial safety issues, proposed regulatory positions, and research.
mendation provided in SECY-94-017, " Options with Regard to Revising 10 CFR Part 100, Reactor i
l Site Criteria," that the proposed rule issued for com-Reactor Accident Analysis j
ment in October 1992 not be adopted and that Part l
100 be revised to emphasize siting aspects by includ-Reactor Risk Analysis mg baste site criteria.
A simplified analysis of potential in-plant and offsite In May 1993, a proposed rule was published on the accident progression and the health consequences of emergency planning licensing requirements for in-accidents initiated during low-power and shutdown dependent spent fuel storage facilities and moni.
operating conditions has been performed and pro-tored retrievable storage facilities.
vided to the Office of Nuclear Reactor Regulation (NRR) in support of their regulatory activities, as in June 1993, a proposed rule was published on re-documented in NUREG-1449.The results of Phase e
vised emergency planning that would update and 2 of this work will be published as contractor clarify ambiguities that have surfaced in the imple-NUREG reports early in FY 1994.
NUREG-1266 viii
Highlights The South 1bxas nuclear power plant requested o
An extensive series ofintegral effects testing for di-e modifications to its plant technical specifications rect containment heating (DCil) at different scales based, in part, on its risk analysis. 'the RES staff is was completed. 'IWo reports were completed, one to now working with NRR on the acceptability of the describe the DCH issue resolution process for the requested modifications and expects to complete Zion and Surry plants and PWRs in general, and the this work early in FY 1994.
other to document the probability of containment failure by DCIi at Zion.These reports are undergo-PRA data from three more licensed nuclear power ing peer review and will be completed in FY 1994.
o plants were added to the SAPHIRE (System Analy-i The independent peer review of the SCDAP/RE-l sis Programs for Hands-on Integrated Reliability e
Evaluation) data base; most of the data from pre.
LAP severe accident computer code was completed, vious plant loads was updated to Version 5.0.
and improvements to the code based on peer review j
recommendations were initiated. An independent o
The NRC continued working with the Commission peer review of the CONTAIN code began in FY 1993 of the European Communities and the Organization and will be comp!cted in FY 1994.The peer review of for Economic Cooperation and Development to per-NRC's third major severe accident systems code, form an intercomparison exercise on probabilistic MELCOR, was completed and improvements to the accident consequence codes.
code made in FY 1993 to address peer review com-ments.
o A survey and evaluation of aging risk assessment methods and applications is being performed. A The late-phase core melt progression experiment, e
draft contractor NUREG report has been received; MP-2, was completed as were two initial ex-reactor the final report will be published early in FY 1994, tests of core damage progression in BWR dry core accident conditions. Results of these experiments,as
,lhe PRA Working Group, organized to provide well as plans for future research in core melt pro-o guidance to the staff on the use of PRA, completed a gression, will be reviewed early in FY 1994 by an ex-draft report in October 1993.The report includes im-pcrt peer review group to provide input on the need for additional research in this area.
tial guidance on the use of PRA in screening and analyzing reactor operational events and on basic terms and methods used in PRA. The report also Severe Accident Policy Implementation contains a number of recommendations for addition-Twenty-six new IPE (internal-event) submittals were al guidance development, improvements to the received (a total of 63 to date) and seven were eva-NRC's PRA training program, and improvements in luated. It is expected that all IPE submittals will be PRA tools and data bases used by the staff. It is ex-pected that the final report will be published in received and reviewed by the end of CY 1995.
March 1994.
Four IPEEE (external-event) submittals were re-in response to a request from the Office of Analysis o
and Evaluation of Operational Data (AEOD), RES has developed a new course that is mtended to treat Safety issue Resolution and Regulation reactor safety in a broad sense. TWo presentations of Improvements the course have been offered at NRC's 'Ibchnical
" Raining Center in Chattanooga during 1993. With Earth Sciences the developmental work completed, responsibility cn et the course will be turned over to Several noteworthy developments occurred in stu-e dies related to the strong ground motions. By study-ing 97 carthquakes recorded by the Eastern Canada Network, a model for attenuation of ground motions Containment Performance was developed, and information was obtained about propagation and source characteristics in the East-o ne TMI-2 Vessel Investigation Project was com-ern United States, pleted and final technical reports issued. Examina-tions of reactor vessel lower-head samples were Another ground motion study concerned rupture e
completed, and the results were used as input to per-histories of eastern North American earthquakes.
form calculations of potential reactor vessel failure Such large events as Miramichi, Nahanni, Ungava, modes. %e conclusions and results of this project and Saguenay were included in the study and the were presented at a 3-day public meeting in October time histories of movement on the earthquake 1993 in Boston, Massachusetts.
failure surface were obtained for these events by ix NUREG-1266
I Highlights inverting teleseismic and strong motion recordings.
computer system and a reactor shutdown cooling sys-
'lhese studies provide improved understanding of tem.
the earthquake source mechanism in the Eastern United States and also allow for better source mod-Generie Safety Issue Resolution cling for ground motion predictions.
During FY 1993, the NRC identified five new gener-In an update of the earlier work, a stochastic ground ic issues, established priorities for 12 issues (see motion model was extended to include four broad
'lhble 6.1), and resolved 10 issues (see Table 6.2).
site categories for the ground motion prediction tak-Thble 6.3 contains the schedules for resolution of all ing into account local site geology. In the previous unresolved issues, work, the model was limited to the deep site condi-tions.
Reactor Regulatory Standards in April 1993, the NRC conducted a public workshop e
Plant Response to Se.isnue and Other on elimination of requirements marginal to safety.
External Events The purpose of the workshop was to provide infor-mation on the NRC program, solicit comments from The NRC published for public comment the pro.
the pubhe and regulated industry on the program, e
posed revision of Appendix A," Seismic and Geolog-and discuss a number of specific tmtiatives being con-ic Siting Criteria for Nuclear Power Plants," to sidered.
10 CFR Part 100. Responses were received from ap.
proximately 47 domestic and foreign commenters.
The NRC issued a draft report, " Proposed Regulato-e Ihe staff is reviewmg all the comments and will re" ry Analysis Guidelines,"(NUREG/BR-0058, Revi-vise the proposed rule as appropriate in FY 1994.
sion 2) for public comment in September 1993. The Proposed guidelines represent the NRC's policy-Seismic testing of relays to determine the influence setting document with respect to regulatory impact e
of relay chatter on circuit breaker tripping was com-nalyses (RIAs), in addition, a draft report, Regu pleted in FY 1993. Results obtained indicate.that the I tory Analysis Tbchmcal Evaluation Handbook,-
acceptability of relay chatter depends on the circuit (NUREG/IlR-0184)wasissued.Thepurposeof the parameters of the specific circuit. These results will handbook is to provide guidance to regulatory ana-facilitate implementation of the resolution of USl lysts, to promote preparation of high-quality RIAs, A-46," Seismic Qualification of Equipment in Oper-and to implement the pohcies of the guidehnes.
ating Plants," and the IPEEE program.
A final rule,10 CFR Part 50, on training and qualifi-During FY 1993, the NRC initiated a program to re-cation of nuclear power plant personnel was issued e
view significant changes being proposed to portions in April 1993. The rule as revised codifies existing in-of the ASME Boiler and Pressure Vessel Code, Sec-dustry practices related to personnel training and tion III, Nucicar Power Plant Components, Division qualification and meets the directives of the Nuclear 1, which deal with scismic design of piping systems.
Waste Policy Act of 1982.
The objectives of this program are to (1) assist the NRC staff in developing regulatory changes and per-A final rule and six regulatory guides relating to the form supporting research activities as needed, and effective implementation of the new 10 CFR Part 20 (2) evaluate the cumulative impact of proposed rule were issued in FY 1992. Three additional regu-changes on the overall safety margins of the piping latory guides needed for the implementation of the systems. This program will be completed in 1995, al-revised Part 20 rule were issued in June and July lowing the staff to develop its position on the piping 1993.
design requirements.
A proposed rule,10 CFR Part 55, on requalification A collaborative effort involving exchange of techni-requirements for licensed' operators for renewal of e
calinformation was established with the Ministry of licenses was issued in May 1993. The proposed International'Irade and Industry and Nuclear Power amendment would delete the requirement that each Engineering Corporation (NUPEC) of Japan. In this licensed operator pass a comprehensive requalifica-effort, NU PEC is carrying out a scismic proving test tion written examination and an operating test con-program for a main steamline typical of the PWR ducted by the NRC during the term of the operator's plants and a feedwater system typical of the llWR 6-year license as a prerequisite for license renewal.
plants. The NRC will carry out pre-and post test A final rule,10 CFR 50.65, on monitoring the effec-analyses to assess the applicability of currently avail-able analytical models, in addition, data will also be tiveness of maintenance at nuclear power plants was obtained from NUPEC for scismic proving tests of a issued in June 1993. The rule requires that the NUREG-1266 x
l
Ifighlights 3
licensee monitor the performance or condition of A final rule on licenses and radiation safety require-certain structures, systems, and components (SSCs) ments for irradiators was published in February 1993.
against licensee-established goals so as to provide
'lhe rule established a new lbrt 36 to specify radi-reasonable assurance that those SSCs will capable of ation safety requirements and beensing require-performing their intended functions.
ments for the use of licensed radioactive materials in irradiators.
In June 1933, Regulatory Guide 1.160," Monitoring the Effectiveness of Maintenance at Nuclear Power Plants " which endorses a NUM ARC guidance doc-Uran.ittIn Eriricliiiierit ument, was issued.
The Commission is reviewing the proposed rulemak-ing developed to amend 10 CFR Ibrt 76," Regulation Part 3--NUCLEAll MATEllIALS Governing the Operation of Gaseous Diffusion Faci-LICF'NSING AND nues." ms rulemaking, required by the Energy ItEGULA,I, ION SUPPOlt,I, Policy Act of 1992, would establish both the procc-dural and technical requirements for certification of the operation of the gaseous diffusion facilities by Nuclear Materials U.S. Enrichment Corporation.
o A proposed rule,10 CFR lbrts 30,40,50,70, and 72, that would allow self-guarantee as an additional LOW-LCVel WaSic D.ISpOSal mechanism for financial assurance was published in A final rule to amend 10 CFR Ibrt 61 tc clarify that January 19)3. 'Ihis proposed rule is in response to a petition for rulemaking submitted by the General the requirements related to the performance ofiand Electric Company and Westinghouse Electric Cor.
disposal facilities for low level waste are applicable poration.
to aboveground disposal (i.e., built on the ground without an carthen cover) was published in June o
A proposed rule,10 CFR Ibrts 30,32, and 35, on the 1993.
medical use of byproduct material was published in j
July 17)3. This action, taken in response to a petition A proposed rule on radiological criteria for decom-1 for rulemaking, is intended to provide greater flexi-missioning is being pursued under an Enhanced Ibr-bility by allowing properly qualified nuclear pharma-ticipatory Rulemaking process. This process was ini-4 cists and authonzed users who are physicians greater tiated to actively solicit early input from a wide discretion to prepare radioactive drugs containing spectrum of interests. Seven public workshops were byproduct material for medical use.
held across the United States, and these were fol-lowed by eight generic environmental impact state-A proposed rule,10 CFR lbrt 73, on a physical fit-ment (GEIS) scoping meetings in four cities. The o
ucss program for security personnel at Category I fa-EPA participated in these workshops and meetings cilities was published for comment in October 1993, and is a cooperating agency in the development of the GEIS.
A final rule, Appendix 11 to 10 CFR lbrt 73, on day-o firing qualifications for security personnel at Cate-A final rule,10 CFR lbrt 20, on disposal of waste oil gory I fuel cycle facilities was published in August by incineration at nuclear power plants was pub-1993.
lished in December 1992. The rulemaking action re-sponds to a petition for rulemaking originally filed by A final rule,10 CFR lbrt 73, amending the regula-Edison Electric Institute and the Utility Nuclear o
tions covering the physical protection of special nu-Waste Management Group.
clear material at fixed sites, was published in March 1993.
A final rule (10 CFR Parts 30,40,70, and 72) was pub-lished in July 1993 to amend the NRC's decommis-o A final rule,10 CFR Parts 26,70, and 73, on fitness sioning regulations to require holders of a specific for duty for Category I facilities and shipments was license for possession of byproduct material, source
' published in June 1993.
material, special nuclear material, and independent storage of spent nuclear fuel and high-level waste to o
A final rule and a proposed rule,10 CFR 72.214, add-prepare and maintain additional documentation ing two casks to the list of approved spent fuel stor-identifying areas where licensed materials and
, age casks, were published in April 1993.
equipment were stored and used.
.i xi NUREG-1266
- 1
liighlights A proposed rulemaking,10 CFR lbrts 30,40,70, and Research (SillR) program to stimulate technological e
72, on timeliness in decommissioning a materials fa-innovation by small businesses, strengthen the role cility was published for comment in January 1993.
of small insiness in ineeting Federal research and
'Ihe NRC issued, in final form NUREG/CR-5512, development needs, increase the commercial appll-e Volume 1, " Residual Radioactive Contamination cation of NRC-supported research results, and im-from Decommissioning: lbchnical Ilasis for 'Itans-prove the return on investment from Federally lating Contamination Levels to Annual"Ibtal Effec-funded research for economic and social benefits to live Dose Equivalent," in October 1992.
the nation. l'articipation in this program has contin-ued since the program was established in FY 1982.
Office Program In FY 1993, the NRC was supporting 20 SillR The NRC supports the Small Ilusiness Innovation projects-in-progress.
e NUREG-1266 xii l
PART 1--NUCLEAR SAFETY RESEARCII--REACTOR LICENSING SUPPORT i
L llEACTOlt AGING AND LICENSING llENEWAL
'this program is conducted to ensure that reactor plant and that sufficient critical experiments are conducted to systems and related components perform as designed dur-validate those procedures and methods.
ing normal operation and transient and accident condi-tions and ensure that their functional integrity and oper-Ensuring the structural integrity of the pressure boundary ability can be maintained over the life of the plant. 'Ihe has been at the center of several recent well publicized program includes the reactor system pressure boundary.
regulatory issues-for example, the 1984 decision to re-Failure to maintain pressure boundary integrity could quire an accelerated schedule of five 13WR inspections compromise the ability to cool the reactor core and could due to cracking in the coolant pipes; the 1991 review of the lead to a loss-of-coolant accident accompanied by sclease Yankee Rowe plant; and the 1992 review of the "Rojan of hazardous fission products.
plant steam generators. Additionally, incidents of cracks and leaks in piping and steam generator tubes have high-lighted the need for materials data, analysis methods, and insputi n tahniques for these components.
1.1 Ileactor Vessel Safety aml Piping Illtegrity Much of the work is completed and has been put in prac-tice through several regulations, regulatory guides, and parts of the standard review plan, as well as through na-1.1.1 Statement of Problem tional codes and standards.The remaining work is provid-ing the basis for both confirming and revising some of the The reactor system pressure boundary of a light water earlier regulatory positions, with the overall aim of pro-reactor (LWR)is the principal boundary enclosing the viding a stable, fully validated regulatory framework for nuclear fuel core and the water coolant used to maintain ensuring the integrity of the primary pressure boundary suitably low temperatures of the fuct cladding and to for the foreseeable future.
conduct the heat from the fission reaction and convert the water coolant into steam for electricity generation. 'Ihe 1.1.2 Program Strategy primary system includes the reactor pressure vessel, pri-mary coolant piping, primary pumps, and steam genera-The approach used for this element is to develop analyti-tors for pressurized water reactors (PWRs). For boiling cal procedures for predicting continuing integrity or con-water reactors (IlWRs), the primary system includes the ditions for-failure and to ensure that an adequate exper-steam line piping out to the first isolation valve. This imental basis exists to validate those procedures. He boundary must be kept intact and fully serviceable at all most critical facet of pressure vessel integrity is embrittle-times to ensure that water coolant is always available to ment of the pressure vessel steel as a result of exposure to cover the fuel core so that the heat generateo during neutrons escaping from the fuel core during normal ser-power operation or from decay following shutdown can vice. Experiments are conducted to develop a base of always be safely conducted away, thus precluding a core information on all the factors that will cause embrittle-meltdown accident. The principles of ensuring the struc-ment to increase during service life. Much work is done to turalintegrity of the primary system components are em-establish correlations between small-specimen behavior bodied in the elements of fracture mechanics used to and thick section behavior to ensure that the analyses predict conditions for failure. These elements are performed to assess structural integrity are valid. Similar-(1) knowledge of the materialpropertics(strength, tough-ly, the ability to predict integrity in piping has required ness, embrittlement, etc.), especially the changes in those testing of full sized sections of pipe having a variety of properties that can occur as a consequence of nuclear cracks to determine if such cracks could cause failure operations;(2) knowledge of the pressure and other load-during either normal service or accident conditions. For ings that can be applied to the components either from both vessels and piping, knowledge of the rate at which normal operations or from accidents; and (i) knowledge cracks grow is very important to ensure that a component of the presence and size of cracks or other flaws in the will not fail during its forthcoming operational period.
components. The regulations, codes, guides, etc., that Experiments are conducted on a wide variety of pertinent pertain to the structuralintegrity of LWRs were written to materials under a range of typical and expected exposure ensure that possible combinations of material properties, conditions to determine the maximum bounding rates of h> ads, and flaws will yield adequate margins against fail-crack growth. Detection and sizing of flaws and cracks in ure of primary system components.The goal of the reac-all primary system components are conducted by the in-i l
tor vessel and piping integrity element is to ensure that dustry through periodic inservice inspections at shut-appropriate analytical procedures and inspection meth-downs. 'Ib ensure that the inspections reliably detect and ods exist for assessing the safety of comp (ments during accurately si/c the flaws, extensive tests are conducted normal operation and transient and accident conditions with inspection teams drawn from the industry using l-1 NUR EG-1266
-.. =
- 1. itcactor Aging typical equipment and techniques on samples whose flaw under design basis and hypothetical accident condi-conditions are known. From the results, it is possible to tions.11asic work is being performed by researchers determine which techniques are effective and the magni-at the Oak Ridge National Laboratory (ORNL),with tude of the error bands for flaw detection and si/ing.
additional efforts being performed by researchers at improvements in methods are proposed and qualification llrown University, the University of Illinois,'Ibxas procedures developed that can provide better assurance A&M University, and the U.S. Navy's Naval Surface of flaw detection in future inspections and for sizing flaws Warfare Center (NSWC), Annapolis Detachment, more accurately. Materials and components removed These researchers are developing state-of the-art from actual service are used to measure material proper-analysis methods and evaluating those methods ties after years of service, to evaluate the extent of corro-against test data developed as part of this program by sion, and to validate the existence of flaws that have pre.
ORNI, the National Institute for Standards and viously been identified and had their size estimated.
Technology (NIST), and the NSWC.The initial work has been very promising, and the program has been continued to permit evaluation of test geometries 1.1.3 Research Accoinplishments in FY 1993 that are more typical of reactor pressure vesseis.
Additionally, the researchers are coordinating their 1.1.3.1 Pressure Vessel Safety work with international efforts in this technical area.
This area of NRC research focuses on ensuring the struc.
Collaborative efforts with a European Community tural integrity of the reactor system pressure boundary, program are expected to provide results from a test i.e., keeping it free from damage and leaktight. 'lhe un-that will closely simulate a reactor pressure vessel derlying concern in ensuring the integrity of the pressure subjected to accident loads. This wdl provide a more realistle validation of the revised analysis methods, boundary is that failure to do so could compromise the operator's ability to cool the reactor core and could lead to During FY 1993, the results of several efforts were a loss-of-coolant accident (LOCA) accompanied by the put to use in performing generic analyses of reactor release of hazardous fission products.
pressure vessels fabricated from materials with a low resistance to a " ductile tearing" failure mode. In the The research program in this area is a broad based pro-carly 1970's, the NRC recognized that some pressu re gram, initiated in 1%7.The original program was focused vessels were fabricated using steel plates and some solely on the properties and fracture behavior of the reac' weld materials that did not provide the high resis-tor pressure vessel-the large, thick walled steel cylinder tance to this failure mode exhibited by most of the that houses and supports the reactor core. As the full plates, forgings, and welds used in reactor pressure challenge of ensuring the integrity of this critical compo' vessels.The NRC issued Appendix G to 10 CFR Part nent was realized, the scope of the research program was 50 in 1973 to provide explicit requirements on the expanded to include irradiation damage, service induced Charpy upper-shelf energy-a measure of the duc-cracking mechanisms, and methods for periodically in' tile tearing resistance of these materials-for both spectmg the pressure vessel, new construction and for operating plants. However, it was recognized that some of the early vessels did The pressure vessel safety program is closely tied to the not meet these requirements. Herefore, Unre-regulatory efforts to ensure pressure vessel integrity.The solved Safety issue A-11," Reactor Vessel Materials technical efforts in the research program-fracture analy-
'Ibughness," was established to develop methods to sis and radiation embrittlernent-are central to sound evaluate the integrity of pressure vessels that did not regulatory positions addressing the safe operation of the satisfy the Appendix G requirements.That issue was pressure vessel. For example, efforts to revise the basis for resolved in 1982 with the publication of " Resolution determining the allowable operating pressure and tem-of the 'lhsk A-ll Reactor Vessel Materials Tbugh-perature to preclude brittle failure of the pressure vessel ness Safety Issue" (NUREG-0744). A key aspect of are drawing on research results from the pressure vessel the overall resolution was a request to the American safety program.
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to recommend criteria 1.
fracture Analysis The part of the pressure vessel that would satisfy the Appendix 0 requirement to safety program that includes fracture analysis meth-demonstrate margms of safety equivalent to those in ods has had a particularly large role in the overall the ASME Code.The Codecommittee respondedin program during FY 1993. The fracture analysis work February 1991 with a set of evaluation criteria that involves an ongoing program to develop, and reduce were reviewed and accepted by the NRC.The Code l
to practice, advanced analysis methods that will im.
also developed an analysis method (Code Case prove the ability to predict the allowable pressures N-512) that was similar to the NUREG-0744 recom-and temperatures for the pressure vessel and the mended method but that had the benefit of several ability to evaluate the integrity of the pressure vessel years of review and experience by the Code NUREG-1266 1-2
1, Reactor Aging -
l committee members and by the NRC staff. Unfortu-removed from the cancelled Midland Unit I reactor nately, Code Case N-512 (Section XI, Division 1, pressure vessel. The materials being irradiated are February 1993) did not address cornplete details of representative of the so-called " limiting" material in all the potential loading conditions for reactor pres-several operating nuclear power plants. Additional-sure vessels, nor did it include guidance on deter-ly, these materials are being irradiated in the surveil-mining appropriate material properties for use in the lance programs of an operating power plant as part evaluation me: hod.
of an NRC-industiy coordinated research effort, When the results from each of these programs are
'Ihe research staff undertook an effort to address the available in the late 1990's, they will provide impor-shottcomings of the Code Case during FY 1993, tant information about the embritt!cment trends for
'these staff efforts include development of a draft these materials and equally important information regulatory guide that would expand the Code Case's about the differences between test reactor and pow.
guidance to include evaluation methods pertinent to er reactor irradiation conditions and about the all service loading conditions and would provide spe, mechanisms controlling embrittlement of these ma-cific guidance on estimating material properties. In tenals.
developing this draft guidance, the staff drew on results from past NRC-funded research efforts at During FY 1993, ORNL processed the updated Eva-ORNL and the l'acific Northwest utboratories luated Nuclear Data File, Version 11-VI, to develop (PNL) to provide a comprehensive fracture analysis neutron cross section libraries that can be used in methodology. Additionally, results from a Phase !!
evaluating the neutron fluence for power reactors.
Small ilusiness Innovative Research (SillR) pro.
'these cross section librancs are needed to predict gram were used as an acceptable method for estimat, the neutron fluence, which is an essential m, put in ing the material properties. 'Ihe draft regulatory estimating the level of radiation embrittlement for guide was published at the end of September 1993.
reactor pressure vessels.':he work will be completed in early 1994, and the updated cross section libraries In addition to developing this draft regulatmy guide, will be available for use shortly thereafter. In addi-the research staff worked with researchers at ORNL tion to the cross section library work, researchers at to perform generie analyses using the draft regulato-OM have worked with researchers m the Czech ry guide methodology in support of the regulatory Repubh,c, and other East European researchers, in staff efforts to evaluate responses to Generie Letter performmg calculations to predict the results of 92-01. " Reactor Vessel Structural Integrity." This carefuHy controlled ' benchmark expen,ments con.
work demonstrated that, on a generic basis, ade.
ducted by the Czech researchers. this cont;muing quate pressure vessel integrity was ensured even work is providm, g important data to the NRC s pro-with Charpy upper shelf energy values below the gmm to yaHdat,e neutron nuence calculabon meth.
value specified in Appendix G to 10 CFR lbrt 50, ods and ts providing technology transfer and valida-tton of the methods being used by the different laboratoriesJihis work contributed to the staff's ef-2.
Radiation Embrittlement. One concern m. ensuring fort to prepare a draft regulatory guide on "Calcula-the integrity of the reactor pressure vessel is em*
brittlement of the pressure vessel steel caused 16 tional and Dosimetry hiethods for Determining Pressure Vessel Neutron Fluence," which was pub-neutrons escapmg from the reactor core during not-I shed for public comment in September 1993.
mal operationJihese neutrons impinge on the pres.
sure vessel wall and, through a complex process, Work continued in FY 1993 to compile and evaluate reduce the ability of the steel to resist fracture. Em.
embrittlement trends using the power reactor pres-brittlement increases with continued operation. 'Ib sure vessel material surveillance data. 'these data ensure the continued safe operation of pressure ves-are reported to the NRC in accordance with Appen.
sets, the research program includes a significant ef*
dix 11 to 10 CFR lbr 50 and reflect embrittlement fort to quantify the effects of neutron radiation em-trends for reactor pressure vessels irradiated under brittiement, to understand the mechanisms that typical power reactor conditions. 'the OR NL work to control this process, and to find methods to mitigate compile these data into a comprehensive data base the embrittlement and recover the original fracture has provided the basis for work by blodeling and toughness-Computing Services to develop statistically based models for predicting radiation embrittlement. Ad-During FY 1993, the radiation embrittlement re.
ditionally, the ORNL data base has been used by the search efforts moved forward on several fronts. 'Ibst regulatory staff in both plant-specific and generic reactor irradiations were initiated by ORNL, using evaluations.The ORNL work is a continuiug effort the University of hiichigan test reactor, to evaluate while the hiodeling and Computing Services work is the effects of neutron radiation on weld mateilats expected to be completed in late 1994 or early 1995, 1-3 NUREG-1266
- 1. Reactor Aging This work will enable the NRC to evaluate the need 1.1.3.2 Steam Generator'Ibbe Integrity l
for further revision to Regulatory Guide 1.99, which provides the methods for estimating radiation em-Steam generator tube integrity and identification of deg-brittlement and is a fundamental part of the NRC's radation continues to be an important area in the research l
approach to ensuring pressure vessel safety.
program. The thin-walled steam generator tubes are an integral part of the reactor system pressure boundary.
j The research to ur.derstand the mechanisms of radi.
'Ibbe failures could Icad to a LOCA and containment ation embrittlement continued in FY 1993, with sig.
bypass resulting in the escape of radioactive fission prod-nificant advances being made by the University of ucts directly to the environment. During FY 1993, the California at Santa llarbara (UCSil) and ORNL, in research staff worked closely with the regulatory staff in i
conjunction with researchers in the United King.
the review and evaluation of industry-proposed interim
)
dom, in modeling the complex interactions among (or alternative) criteria for plugging those tubes where the impinging neutrons and the atoms in the pres.
stress-corrosion-cracking degradation has been detected i
sure vessel steel. This work is closely integrated with at the tube-to-tube-support-plate intersections. The re-the experimental work being done at UCSil, at scarch staff developed a mechanistically based model for ORNL, and in Eumpe. Understanding the control-Predicting tube failure and cumulative leak rate of de-l ling mechanisms is essential t'o confidently extrapo.
graded tubes under normal and main steam line break late the empirical r.iodels of radiation embrittlement conditions.The model was useful forcomparisons to leak to imique pira-specific operating circumstances.
rates predicted by the industry's statistically based model Th e prorm,in the mechanisms research is provid.
and for gaining an understanding of the important param-eters that influence failure and leak rates.This modelwas int, assurance that the empirical models are conser.
vative and is helping to define the limits of extrapola.
used by an NRC'Ihsk Group on Interim Plugging Criteria, tion for those modelr.
which was charged with evalaating the industry proposal.
The research staff and research contractors from the Ar-g nne National Laboratory (ANL), PNL, and ORNL con-The interactions with researches in Russia under tributed to the task group's effort to develop an NRC the auspices of the Joint Coordinating Committee consensus position on mterim pluggm, g entena for steam on Civilian Nuclear Reactor Safety (JCCCNRS) generator tubes, which was published in NUREG-1477.
Working Group 3 were continued in FY 1993. "Ihe
.I'he research activities focusing on steam generator tube scope of the continuing efforts has expanded from integrity are continuing and are focused on validation of radiation embrittlement to include the general sub-inspection results, integrity prediction methods, and leak ject of pressure vessel integrity analysis methods.
rate estimation methods, During FY 1993, coordinated test reactor irradi-ations were continued in the United States and in Russia, and ORNL hosted a visiting scientist on a 1.1.3.3 Piping Integrity 1-year assignment to investigate radiation embrittle-ment and thermal-annealing effects. The Russian The piping integrity research prograro has bem an impor-scientist's work at ORNL has provided valuable in-tant part of the overall pressure bwdary integrity re-formation and data concerning the ability of thermal search program for many years. It took on heightened annealing to mitigate the effects of radiation em-important in the early 1980's v the emphasis on inter-brittlement. Work in this area is continuing.
granular stress corrosion crackns,.JSCC)in BWR pip-ing systems, and the work on environmentally assisted The interactions with the Russians have provided cracking and piping fracture behavior became high-prior-first-hand examination of their procedures for con-ity activities. Over the last decade, research on IGSCC has ducting thermal-annealing treatments of their reac.
largely been completed and has contributed to the NRC's tor pressure vessels. Additionally, the NRC started position on this issue. The pipe fracture research is draw-research into the metallurgical effects of thermal ing to a close, with the work during FY 1993 being focused annealing many years ago and in the mid-1980's on conducting large-scale pipe fracture tests to provide funded work to examine the engineering feasibility final validation of the methods used by the NRC and by of this process for U.S. plant drigns and conditions.
the ASME Code to ensure that piping will not fail under During FY 1993, the research staff made use of this accident conditions. Further, work to evaluate the signifi-experience end the results of prior research to draft a cance of the reactor operating temperature on the frac-regulation and regulatory guide addressing both the ture behavior of cast stainless steel piping and compo-engineering and metallurgical aspects of thermal nents was completed during FY 1993, and the results of annealing for U.S. plant designs. The perform-the research have been provided to, and are being used by, ance-based regulation and guidance are undergoing the regulatory staff in plant-specific evaluations. Also, the final internal reviews and should be published for research program on environmentally assisted cracking comment in early 1994.
has been developing data on the effects of the water NUREG-1266 1-4
- 1. Reactor Aging coolant on the fatigue life of pressure boundary compo-approach and include programs addressing each of the nents.
major considerations in providing structural integrity-analysis methods, material properties, and inspection Environmentally Assisted Cracking. Irradiation-assisted techniques. The research program for inspection procc-l stress corrosion cracking (I ASCC) of core internal compo-dures and technologies provides an independent basis for nents fabricated from solution-annealed austenitic stain-cvaluating the efficacy and reliability of industry inspec-l less steels and high-nickel alk>ys has been observed in tion techniques. The program includes studies of im-
)
both IlWRs and PWRs. Failures caused by IASCC have proved methods for selecting components for inspection become more numerous as reactors age and core materi-and strategies for setting the sample size and inspection als accumulate higher fluence. Although many of the intervals to provide a reliable overall inspection process.
affected components can be replaced, some safety-signifi-The program also includes the inspection technologies 4
cant components would be difficult or impractical to re-and methods necessary to ensure reliable detection and place. Additional tests were perfomed in FY 1993 to accurate sizing of flaws. Finally, the program includes a compare the susceptibility of conventional commercial focused effort to transfer this technology to practitioners purity type 304 stainless steel, which is the material cur-in the NRC regional and headquarters offices.
rently used in most reactors, with special high-purity heats of type 304 stainless steel--materials that have been sug-Cast Stainless SteelInspectability. Cast stainless steel has gested as a replacement for the current materials. Both been widely used in piping systems and pipe components types of materials were irradiated in commercial power in both the nuclear and non-nuclear industry. Owing to its reactors and, contrary to expectations, the susceptibility unique microst ructural characteristics, this material is dif-of high-purity materials to IASCC was higher than those ficult to inspect reliably. During FY 1993, PNL research-of the commercial purity materials. Microstructural eval-ers developed and evaluated a technique that is based on uations of the irradiated materials are in progress to de-equipment and procedures developed in earlier NRC re.
termine the underlying mechanisms for this mode of ma-scarch, e.g., the use of specially designed low-frequency terial degradation. Numerous heats of stainless steel with transducers coupled with the Synthetic Aperture Focus-a range of compositions are being irradiated in the Halden ing Technique for Ultrasonic Testing (SAFT-UT). The reactor in Norway as part of an investigation of material tech nique was evaluated in a blind test coordinated by the compositions and water chemistries that promote suscep.
Electric Power Research Institute (EPRI) in conjunction tibility to this type of cracking.
with Northeast Utilitics.The PNL technique showed im-proved results over those currently being used and ranked Fatigue is a potentially significant damage mechanism in in the top group of techniques tested. The majority of LWR pressure boundary system components.The repeti-cracks in the samples were detected, but the " false call" tive nature of the service conditions in nuclear power rate was higher than desirable. Further work is planned to plant systems, particularly the piping systems, lead to pro-refine the technique to improve crack detection and to gressively greater damage. This phenomenon has been lower the false call rate.
recognized for many years, and design criteria were in-cluded to account for it. The early design Codes provided Risk-Based Inspections. The NRC has funded research to a less rigorous fatigue analysis than does the current de-evaluate methods for establishing an overall inspection sign Code. Ilowever, recent data suggest that the effects program-components to be inspected, sample size, and of the water coolant on the expected fatigue life of typical inspection frequency-based on risk concepts. The materials are not adequately addressed by the design Co.
successful application of the risk-based methods in a dem-des-past or present. Since no consensus design proce-onstration project using the Surry Unit 1 (in Va.) nuclear dure is availabic, data from ongoing tests and from the power plant led the ASME Code to establish a working literature were evaluated during FY 1993 to identify the group under the subcommittee responsible for Section XI key environmental variables that influence fatigue life (the inservice inspection section of the Code) to consider and to define the material and loading conditions for changes in the Code's approach.
which additional information is needed. During FY 1993, interim design curves were developed and published that During FY 1993, the NRC supported the working group's more adequately describe fatigue life in the high-temper.
efforts through active participation by the research staff i
ature water characteristics that could exist in LWR cool.
and researchers from PNL.'Ihe working group has moved ant systems. Fatigue tests are in progress to validate and/
quickly to develop initial proposals to revise Code criteria or update the proposed design curves.
for the inspection of pressure vessels and piping. The expected benefits of implementing risk-based inservice L L3.4 inspection Procedures and Technologics inspection programs include enhanced plant safety and the climination of costly inspections, particularly those The NRC's approach to ensuring the integrity of the reac-inspections t hat involve higher levels of occupational radi-tor system pressure boundary builds on the overall de-ation exposures when inspecting " low risk" components.
fensc-in-depth concept. Research activities parallel this Completion of this work will provide Code revisions that 1-5 NUREG-1266
- 1. Reactor Aging promote reliable detection of defects before degradation 1.2 Aging of Reactor Components can compromise the structural integrity of " risk impor-tant" components.
1.2.1 Statement of Problem
^8 "S.a ts au mammums, systems, angmpo Surface Eughness &aluation. As insPcction capabilities nents m vanous degrecs and has the potential to increase have improved, and emphasis has shifted toward reliable risk to public health and safety if its effects are not con-detection of flaws, the surface roughness and local geom-trolled. In order to ensure continuous safe operation, etry conditions of the region being inspected have become measures must be taken to monitor key structures, sys-more important. Ilowever, there are no ASME Code tems, and components and interfaces to detect aging deg-requirements for surface conditions for inservice inspec-radation and to mitigate its effects through maintenance, tions. The surface conditions for plant components, which repair, or replacement. For an older plant approaching were specified by the Code for construction radiography, the end of its design life and for which extended operation can vary appreciably. 'Ib address this issue, the NRC has beyond its original license period of 40 years is contem-entered into a cooperative program with EPRI.
plated, aging becomes a critical conce rn and will clearly be crucial to any assessment of the safety implications of j
Working under EPRI funding, the Iowa State University Center for Nondestructive Evaluation is developing com-The NRC and the nuclear industry have initiated a signifi-puter codes for four different inspection modes to de-cant effort aimed at renewing plant licenses beyond their scribe ultrasound interactions and responses. Working original term of 40 years. According to an early Depart-under NRC funding, PNL will expenmentally validate inent of Energy study, the projected net benefit to the and use the computer c*s to assess vanous surface United States economy can oc on the order of $230 billion conditions, These studies wdl form a techmcal basis for through the year 2030, assuming a 20-year period of ex-proposing Code requirements for surface roughness and tended operation for current plants.The benefit reflects changes in geometry. I\\vo of the four codes were com-both the lower fuel cost of the nuclear plants and reduced pletcd and expenmentally validatcd in FY 1993, and eval-outlays for replacement of generating capacity. The li-uations of acceptable surface conditions were started.
cense renewal rule,10 CFR Part 54, " Requirements for Renewal of Operating Licenses for Nuclear Power Pipnts," was issued in final form in December 1991.The j
Intemational Reliability Studies. The NRC is an active par-amti i f rm of draft Regulatory Guide DG-1009, Stan-ticipant and Icader in the Program for the Inspection of dard Format and Content for Applications to Renew Nu-Steel Coniponents, Phase III(PISC III).This internation-clear Power Plant Operatmg Licenses, was issued for al program, organized in 1986 by the Community of Euro-comment m 1991. Since publicat,on of the final hcense i
pean Countries (CEC) and the Committee for the Safety mnewal rule, a number of significant policy issues have of Nuclear Installations of the organization for Economic been identified. As a result, the Commtssion is m the Cooperation and Development's Nuclear Energy pr cess of amending the license renewal rule.
Agency, is assessing the effectiveness of nondestructive testing technologies and procedures for inservice inspec-tion (ISI) of nuclear power plant components. The CEC L2.2 Program Strategy and participating organizations have invested an esti-mated $40 million la the program, including contributions NRC staff effort in aging is being pursued in several areas, of material, inspection services, and manpower.The prod.
including technical and scientific research to identify the ucts from this program will assist regulators and Code effects of aging on the key safety-related components of bodies in establishing technical bases for improving ISI the plant and to examine methods for mitigating such requirements, effects. Specifically, the strategy is to achieve, relative to each component, the following results:
The focus of the PISC 111 program is on the nondestruc-1.
Identify and characterize aging and service wear ef-tive testing of realistic LWR primary circuit components fects that, if unmitigated, could cause degradation oi containing realistic flaws.The experimental portion of the structures, systems, and components and thereby PISC 111 program was completed during calendar year impair plant safety.
1993, and the results will be released during a syrnposium to be held in the spring of 1994. As the results become 2.
Develop methods of inspection, surveillance, and 1
available, the NRC will assess the adequacy of current ISI monitoring and of evaluating residuallife of struc-requirements and, as appropriate, will recommend tures, systems, and components that will permit changes to the ASME Code and to NRC regulations and compensatory action to counter significant aging ef-regulatory guides.
fccts prior to loss of safety function.
NUREG-1266 1-6
- 1. Reactor Aging i
l 1
3.
Evaluate the effectiveness of maintenance, repair, information on equipment aging that n ny be considered and replacement practices, current and proposed, in for the implementation of the license renewal rule.
mitigating the effects and diminishing the rate and the extent of degradation caused by aging.
Standard Technical Specifications Aging Assessment. The NPAR program is evaluating the Standard Tbchnical Specifications (STS) for selected nuclear power plant sys-1.2.3 Research Accomplishments in FY 1993 tems and components to determine the effectiveness of current surveillance requirements (SRs) to detect age-re-1.2.3.1 Aging Research lated degradation effects.The purpose of these SRs is to Aging affects all nuclear reactor structures, systems, and ensure the operability and availability of safety-related components. If aging degradation is not detected and systems and components by venfymg and demonstrating corrected,it can increase risks to public health and safety, that they are capable cf performing their required func-Failures of safety-related comp (ments have occurred in tions. Agmg effects have not always been recognized or the past because of such age-related degradation pro-addressed explicitly in the SRs; many sigmficant forms of cesses as corrosion, embrittlement wcar,and fatigue.The aging degradation may not be detected prior to failure objective of aging research is to develop the technical and, in some cases, the test methods and frequency of N ses for continuous safe operation of nuclear power testing may even contribute to premature degradation.
t mnts as they progress through their design life; to define the operative aging mechanisms; and to confirm existing
,I'he STS aging assessments were completed during FY and/or developing recommendations for new detection 1993 for check valves, the auxiliary feedwater system, and and mit;gation methods in order to prevent or mitigate the the reactor protection system. These assessments re-deleterious effects of the aging process.
viewed the current surveillance and testing requirements to determme their effectiveness in detecting degraded conditions in systems, structures, and components (SSCs)
The Nuclear Plant Aging Research (NPAR) program pro-pri r to failure; to determme if the current survedlance vides technicalinformation usefulin understanding the effects of aging on the 1,afety functions of electrical and and test methods and frequency of testing contribute to premature agmg degradation; to identify, based on the mechanical components of commercial nuclear power results of the mdepth aging assessments of specific sys-plants. During FY.1993, preliminary or comprehensive tems and components conducted under the NPAR pro-aging assessments were completed or final reports were issued for the following safety-related components, sys-gram, tk parameten or m&aton usdul to momtor &
degraded state of SSCs; and to develop recommendations tems, and associated special topics:
for inspection, surveillance, trendmg, and condition mon-itoring methods to be incorporated as part of SRs to evalu-o HWR Internals (NUREG/CR-5754) ate age-related degradation. Work was also initiated to o
PWR Internals (NUREG/CR-6048) combine the 10 STS aging assessments completed into a NUREG/CR report, o
Check Valves (NUREG/CR-5944) o Auxiliary Feedwater System (Follow-on Study)
Aging Management NUREG Update. For several years the (NUREGICR-5404, Vol. 2)
NPAR program has been developing technical under-standing of the processes that, through time-dependent o
Aging, Condition Monitoring, and Loss-of-Coolant age-related degradation of SSCs, could reduce operation-Accident (LOCA) 'Ibsts of Class 1E Electrical al safety margins in operating nuclear power plants below Cables (Ethylene Propylene Rubber Cables-acceptable limits. The results from the NPAR program, NUREG/CR-5772, Vol. 2, and Miscellaneous and other complementary aging management programs, Cables-NUREG/CR-5772, Vol. 3) were compiled and critically reviewed in a draft report, "A o
Recordkeeping (NUREG/CR-5848)
Review of Information Useful for Managing Aging" (NUREG/CR-5562). This draft report has been a valu-o ilEPA Filters and Adsorbers (NUREG/CR-6029) able resource for the NRC staff in such tasks as preparing o
Essential llVAC Chillers (NUREG/CR-6043) the draft standard review plan for the review of license renewal applications for nuclear power plants and in the o
Fans and Blowers review of aging issues related to license renewal.
impact of Aging on Accident Precursors o
NUREG/CR-5562 was extensively revised and updated e
Standard Technical Specifications Aging Evaluation during FY 1993 to provide additional insights and techni-cal guidance for aging management, including (l) identify-7echnicalInformationfor License Renewal. License renewal ing SSCs in which age-related degradation should be man-
)
is a high-priority activity for the NRC and for the nuclear aged, (2) understanding aging mechanisms and identifying power industry.The NPAR program is providing technical degradation sites in these SSCs, and (3) managing 1-7 NUREG-1266
L Reactor Aging 1
i degradation through effective monitoring and mainte-contain safety-related equipment that is essential to [ilant nance programs or by modifications to operating condi-safety. Without cooling, control roora temperature can tions.
rapidly rise leading to operator stress and causing the electronic equipment to give erroneous readings and spu-Information Management. The aging assessments con-rious alarms and to start to fail.The newer digital controls ducted by the NPAR program have generated an exten-are even more sensitive to high temperatures than the sive volume of valuable information on aging processes older analog controls. An NPAR aging assessment of and effective methods for detecting and mitigating aging these essential chillers is in progress. The initial assess-degradation in safety significant systems and components, ment results are documented in " Aging Assessment of Improved methods are needed to make this information Essential Chillers Used in Nuclear Power Plants" readily available to the NRC regulatory staff, nuclear (NUREG/CR-6043). This work shows that chillers are power plant licensees, other government agencies, and affected by vibration, excessive temperatures and pressur-the public. This need was addressed during FY 1993 by es, thermal cycling, chemical attack, and poor quality developing a demonstration electronic document.Techni-coohng water. Aging is accelerated by moisture and non-cal and regulatory information can be rapidly and effi-condensable gases such as air, dirt, and other contamina-ciently computer searched for specific words and topics, tion within the refrigerant containment system; by exces-The electronic document also includes a description and sive start /stop cycling; and by operating below the rated aging overview for the systems and components of inter-capacity. 'lhe primary cause of chiller failures is a lack of est, a compilation of the aging assessment information, adequate condition monitoring and failure to perform and colored line drawings that illustrate the various types scheduled maintenance. The comprehensive assessment and components where aging mechanisms are operative.
now in progress will identify actions that could help re-duce chiller failures and premature aging through effec-1.2.3.2 Components, Systems, and Facilitics tive monitoring and preventive maintenance programs; develop effective procedures for maintaining the reliabil-Engineered Safety Featurcs. Engineered safety feature sys-ity of essential chillers used in all modes, including the tems are systems to c<mtrol and mitigate specific occur-eme rgency standby-only mode; and provide guidelines for rences that might challenge the mtegrity of the reactor a safer and more effective transition to the mandated new and/or adversely affect plant personnel or the general chiller refrigerants.
public. Issues related to the aging and scivice wear of these systems at commercial reactor facilities could im-Jervice Water System. The NPAR service water system pact both public and plant safety. NPAR aging assess-aging assessment was completed with the publication of ments have been conducted for nuclear air treatment and
" Nuclear Service Water System Aging Degradation cooling system fans and for the high-efficiency particulate Assessment" (NUREG/CR-5379). However, an ad-air (llEPA) filters and activated carbon beds that remove vanced power plant monitoring and diagnostic system, radioactive particulates and volatile radionuclides that which was developed as an outgrowth of the NPAR scrvice otherwise could be substantial contributors to public dose.
water system aging assessment, has been successfully The fan aging assessment revealed that aging degradation tested at a U.S. Marine Corps base.
appears to be an important factor when considering fan failure and that breakdown can impact both plant and Emergency Diesel Generators. Although the emergency public safety. Details compiled from surveys containing dicsc' generator aging assessment was completed earlier, information concerning aging and wear effects suggest these is continuing work related to the incorporation of that bearings are the component most frequently linked the findings and recommendations in relevant codes and to fan failure.'ihe investigation also indicated that moni-standards such as those of the Institute of Electrical and toring techniques that will detect irregularities arising Electronic Engineers (IEEE). Guide document P-1205, from improper lubrication, cooling, alignment, and bal-
" Guide for Assessing, Monitoring and Mitigating Aging ance will aid in counteracting many of the aging effects Effects on Class 1E Equipment Used in Nucl ar Power that could impair fan performance. The aging assessment Generating Stations," was published as the result of of the HEPA filters and activated carbon adsorbers identi-Working Group 3.4 effort. It included an appendix on fied heat, moisture, radiation, airborne particles, and con-dicscl generators derived from NPAR information. From taminants as key stressors, with resulting aging mecha-Working Group 4.2, IEEE Std. 387," Standard Criteria for nisms and degradation ranging from particle loading to Diesel-Generator Units Applied As Standby Power Sup-degraded scalant and gasket properties. This work is doc-plies for Nuclear Power Generating Stations," has been umented in " Phase I Assessment of Nuclear Air Treat-approved for final balloting.This new revision of the stan-ment System HEPA Filters and Adsorbers" (NUREG/
dard includes the aging results and information from the CR-6029).
NPAR rescarch.
Chilkrs. Essential chillers are required in nuclear plants Aging Assessment and Mitigation of Major LWR Compo-to cool rooms, such as the main reactor control room, that nems. Intrinsic to the general exploration of reactor aging NUREG-1266 1-8
- 1. Reactor Aging is the assessment and mitigation of aging damage to major was developed in 1993 to incorporate age dependence in components and structures. The objective of this aging PRAs that does not require absolute age-dependent com-assessment task, an element of the NPAR program, is to ponent failure rates. Instead, the aging of a component is identify, develop, and evaluate various aging management expressed in terms of relative aging rates that are found to techniques for the major LWR components and struc-be fairly constant across different components and differ-tures. The approach is to gauge the degradation of the ent plants. A draft report (NUREG/CR-6067) was com-major LWR components and structures by the synergistic pleted on the proposed aging data assessment methodolo-influences of radiation embrittlement, thermal fatigue, gy.
stress corrosion cracking, thermal embrittlement, erosion corrosion, and so forth.
Also in 1993, an important application of the risk-based methods resulted in the development of PRA-based ap-Research completed in this area in 1992 focused on devel-proaches for identifying safety-related motor-operated oping insights for aging management for selected LWR valves (MOVs) having the most impact on plant risk cov-I comrxments and structures to ensure continued safe op-cred under Generic Letter 89-10, " Safety-Related MOV eration. The studies also included the evaluation of ad-Testing and S urveillance." Dynamic tests and surveillance vanced inspection and monitoring methods for character-tests, in accordance with GL 89-10, could then be per-izing the aging damage.1hc results will be useful to the formed on those MOVs having the largest risk impact.
l NRC in identifying and resolving safety issues associated The risk importance of single MOVs and the interaction I
j with LWR aging degradation and developing policies and of multiple MOVs can be analyzed using this approach. A
}
guidelines for making operation plant aging management draft NUREG/CR documenting the results of this work is decisions that may safely extend its operation. The major being prepared.
components assessed in 1993 are the LWR metal contain-ments and the L.WR reinforced and prestressed concrete In addition to the above-described effort, work was initi-containments. PWR reactor pressure vessels and the ated in 1993 to identify and prioritize, based on their risk PWR coolant piping research will continue. Results of significance, environmental stressors associated with these assessments are being documented in a multivo-advanced digital instrumentation and control (I&C) sys-l lume report, NUREG/CR-5314.
tems in nuclear power plants. Analog I&C systems in nuclear power plants are becoming obsolete and are being In addition to the above efforts, a draft report (NUREG/
replaced by digital systems. Digital I&C systems are vul-CR-5824) that discusses the identification of advanced nerable to common environmental stressors, e.g.,
monitoring methods for estimating stresses causing fa-moisture / humidity, temperature.1hc effects of each en-tigue damage was completed.
vironmental stressor are being identified and importance measures are being developed to rank the stressors.
PRA BasedMethodologyforAgingAssessmentsandCompo-These risk-based approaches are being tested using nent Prioritization. The risk-based methodology for assess-plant. specific PRAs.
ing aging in nuclear power plants and for defining priori-ties among risk contributions and maintenance activities Aging of Passive Components. In earlier efforts, a method-(published in previous years, NUREG/CR-5587 and ology was developed to include the effects of aging on NUREG/CR-5510)is subject to uncertaintics due tolim-passive components (pipes, structure, and supports) and ited available aging data and due to various modeling the resulting impact on plant risk. The methodology is assumptions (e.g., modeling of effects of test and mainte-based on probabilistic structural analysis for calculating nance). Tb address the validity of results obtained from the failure probability of these components. The failure the risk-based approach, research in 1993 focused on de-calculation can be incorporated in a PRA for the plant veloping sensitivity and uncertainty analyses to define that will calculate the effects of this failure on plant risk.
priorities addressing data and modeling uncertainties.
During this fiscal year, approaches were investigated that This work was documented in draft N UREG/CR-6045 in can be applied to a large number of passive components 1993.
that exist in a nuclear plant. A screening approach can be used to identify those components that age and contribute The application of age-dependent risk methodology re-most to risk.Two approaches were investigated. The first quires age-dependent comp (ment failure rates; however, is a simple probabilistic structural analysis approach and these age-dependent component failure rates are not the second is an approach called failure attributes. The generally available and need to be estimated from limited simple probabilistic structural analysis is an approxima-recorded plant failure data and plant maintenance logs. A tion of the large, complex structural probabilistic comput-major limitation of the age-dependent methodology has er codes. The second approach uses the attributes that been the lack of recorded comp ( nent aging data and ap-have been shown to most affect aging and failure. These proaches to develop aging failure rates based on the avail-approaches, including a screening approach, will be docu-abic information. lb address this limitation, an approach mented in a report in FY 1994.
1-9 NUREG-1266
- 1. Reactor Aging Valve age at failure The draft report (NUREGiCR-5730) that describes the e
methodology for including the effects of aging on passive Plant age at failure o
components was reviewed internally by other NRC re-System in which the valve was used scarch staff, and responses to their comments have been developed. A revised NUREG/CR-5730 is almost com.
General operating status of the system pleted and will be recirculated for general internal NRC e
Valve manufacturer review. Calculations were also completed to investigate the effects of passive components on the risk of contain-Failure mode ment failure..An approach was developed to identify the Extent of degradation o
passive components that most contribute to risk.
Detection method Aging Effects on Motor-Operated Vah e Performance. In 1993 e
Affected arcas an investigation was initiated to assess the effects of aging i
I on the operability of MOVs and to identify those safe-The characterized data were analyzed for relative failure ty-related MOVs in typical PWR and UWR plants that are rates for each of the characterized parameters. Cross most susceptible to internal or external environmental tabulations of the parametric data were also made. Some aging effects. The investigation has drawn from the Nu-notable observations were:
clear Plant Aging Research reports by ORNL on valve l
There was not a strong relationship between valve aging and from the reports by the Idaho National Engi-e neering Laboratory on high-pressure injection systems.
age and failure rate.
The current investigation includes the reviewof the inser-The largest valves ( > 10 inches) experienced abou:
vice testing programs for four BWR and five PWR plants to identify safety-related MOVs. From this review it has twice the failure rate of smaller valve sizes. The been determined that the majority of the safety related largest valve's failures were also most prone to be valves are either butterfly, globe, or gate valves con-the most significant failures, structed of carbon steel. Internal and external environ-The emergency service water, main feedwater, die-mental conditions are being combined with valve type and material of construction to identify those most susceptible sci starting air, and main steam systems experienced to aging effects. Final identification will also include the the highea failure rates.
safety signific:mcc as well as maintenance and testing Just over half (54%) of the failures were detected by routinely performed. A draft contractor NUREG report is being prepared to report the results of the investigation.
programmatic means, including inservice testing, Also, experiments are under way to compare valve sliding surveillance testing, and other periodic inspection surface friction factors for corroded and uncorroded valve programs.
material specimens and measurement methods to detect Normally operating systems experienced only slight-the changes being evaluated.
ly greater failure rates than did systems that are used nly during shutdown or only when tested.
Check Vahe failure Data Characterization. A detailed re-view of historical check valve failure data for operating The results of the study are being used by ASME Code nuclear plants m the United States was conducted by Working Group OM-22 in support of code development ORNL,Ihe results were published as NUREG/
activities. Follow-on studies for failures occurring during CR-5944. The source of data for the review was the Nu-
"## b ""
clear Plant Reliability Data System (NPRDS), operated by the Institute of Nuclear Power Operations (INPO). Nar-Detection of Pump Degradation. This Phase 11 Nuclear rat,ves describmg the failures of check valves during the Plant Aging Research study examines the leading causes i
period of 1984-1990 were imtially reviewed to eliminate of pump degradation and describes the existing methods those failures that only mvolved external leakage because used in domestic and overseas nuclear facilities to diag-the purpose of the study was to better understand those nose pump problems. Research results are being pub-failures that affect the mternals of the valves. A total of lished in the report, " Detection of Pump Degradation" 4,680 failure narratives were reviewed; of this total,1,227 (NUREG/CR-6089), which evaluates the criteria man-were identified as mvolving int ernal degradation and were dated by required pump testing at the United States nu-further characterized.
clear power plants and comparcs them to those features that are characteristic of state-of-the-art diagnostic pro-The characterization of the failed valves included the grams and practices currently implemented by other ma-following parameters:
jor industrics. Degradation that is caused by low-flow pump operation is also discussed along with new analysis e
Valve size techniques that may be used to ascertain unstable flow. In NUREG-1266 1-10
- 1. Reactor Aging i
i addition, since many pump operational problems can be Steam Turbine Drives. Steam turbine drives for safety-attributed to the pump driver, motor current analysis related pumps are used at most of the commercial nuclear methods are presented that can assist in the determina-power plants in the United States /lbrbine-driven pumps tion of specific kinds of motor degradation, are used in PWR AFW systems. 'Ibrbine-driven pumps are used at IlWR plants in the reactor core isolation cool-Vibration spectral analysis is widely accepted as a power-ing and high-pressure coolant injection systems. The i
turbine-driven pumps provide a means of heat removal fut diagnostic tool for determining numerous types of from the reactor coolant system in the event of station pump degradation, such as misalignment, unbalance, blackout.
h>oseness, and vanous bearing anomahes. Many nuclear plant maintenance departments use vibration spectral analysis to diagnose pump problems.,lhermography and Evaluation of failure records indicates that the turbine lubrication analysis are other important diagnostic tech-8 g
P g
g nologies that have made significant improvements within reported turbine failures. "Ihere are multiple sources of the past decade, both in their case of application and their g vernor problems, with dirty or water-contaminated oil diagnostic capabilities 'lhe next major thrust of develop-and setpoint drift accounting for a large percentage of the ment in diagnostic methodologies is focused on the devel-failures (33%). Oil prob! ems (18%) can be mitigated by opment of expert systems. Low flow operation, which was periodic chemical analysis of oil samples. Setpoint drift often performed by using minimum flow loops to conduct (15%)can be mitigated by periodic calibration of the gov-required ASME Code testing, has been observed to cause ernor. Currently there are no specific regulatory require-pump degradation through destructive low-flow phenom-ments for either chemical analysis or governor calibra-ena. Motor power analysis techniques have also been tion, and current maintenance practices have rot been J
developed that may assist in the determination of the consistent, although there are indications of improve-onset of unstable flow conditions as well as enable the ment in recent years m that regard. The results of the pump analyst to determine the most efficient operational study are being published as NUREG/CR-5857.
ranges of a particular pump system.
Cables. The research program conducted at the Sandia i
National Laboratories on the aging effects on electrical Amiliary feedwater System. In the Phase I study, a thor-cables was completed in 1993 and the three-volume final ough review of system controls and functions was per-report (NUREG/CR-5772, Vols.1,2, and 3) was issued.
formed, and several limitations of current maintenance
'Ihis program cvaluated the capability of cable types com-and surveillance practice were identified, such as the fail-monly found in operating nuc! car power plants to meet u re to verify many safety-related control functions by peri-equipment qualification standards for periods of 20,40, odic testing and the degradation of auxiliary feedwater and 60 years of operation. Several of the cables tested in (AFW) pumps by testing at low flow. The follow-on study this program failed at a fairly carly age. NRC Information categorizes the limitations in current monitoring /operat-Notice 93-33 summarized the cabic test results and ing practice identified in Phase I and evaluates failure alerted the utilitics to possible deficiencies. Research on modes and component degradation caused by these prac-the aging degradation of cable connectors and penetra-tices. The Phase I findings have applicability to all plants tions is currently under way, in that they point out typical testing omissions or sources of degradation.
Contro/ Rod Drive Systems.The Babcock & Wilcox (Il&W) and Combustion Engineering (CE) control. rod drive Significant conclusions of the study (NUREG/CR-5404, (CRD) systems consist of mechanical and electrical com-Vol. 2)are as follows:The present method of testing AFW ponents that position the control rod assemblics in the pumps at the minimum flow condition may lead to degra.
core in response to automatic or manual reactivity cont rol dation of the pu mp and also does not provide an indication signals. Iloth systems are designed to allow rapid gravity of pump condition. 'the report discusses hydraulic insta.
insertion of the control rods upon removal of1he ac power bility at low-flow operation and provides examples from that holds the rods.This study examined the design, mate-the industry of pump degradation. The report also pro.
rials, maintenance, and operation of the system to assess vides head-capacity curves showing that mini-flow tests the potential for age degradation.
are inadequate for assessing pump capability or ability to operate at design basis conditions. An alternative method A detailed operating experience review highlighted of testing is proposed that consists of testing at normal age-related component degradation and failures that operating pressure to climinate degradation and to verify significantly affected plant operation. These effects in-flow at design conditions. Suggested changes in control clude power reductions, reactor shutdowns and scrams, logic testing, check valve monitoring, and emergency ser-and engineered safety feature actuation. While there vice water to A13V system line flow path testing are in-have been no system failures, component failures and cluded.
degradation resulted in increased component stresses and 1-11 NUREG-1266
- l. Reactor Aging unnecessary thermal and pressure cycles that challenged lated MOVs will perform their intended functions over other plant systems.
the life of the nuclear power plants.
The majority of component failures in the CE control Earlier NRC research on MOV performance revealed drive system were caused by the degradation of the con.
deficiencies in the accepted industry formula for calcula-trol system (61%). Failures of the CRD mechanisms ac-ting / predicting required MOV thrusts for controlling counted for 60 percent of reported failures of the B&W flows through the MOVs. These results also contributed control drive system. Aging was the direct failure cause to the issuance of GL 89-10. However, subsequent NRC for 40 percent of the CE power and control system and 55 research resulted in a modified and more accurate formu-percent of the ll&W CRD mechanisms. The operating la for predicting the required MOV thrust values.These and environmental stresses for the system, and the aging latter results and other findings supported the issuance of effects resulting from continued exposure to these other regulatory documents by the NRC. In 1993, addi-stresses, were evaluated for the major system compo-tional effort was expended on refining the modified for-nents. Detailed failure modes and effects analyses were mula for the application of extrapolating low-flow thrust performed for the subsystems. In addition, a survey was requirements to high-flow thnist requirements, which is a made of the current surveillance, inspection, monitoring, process commonly employed by licensees.
f and maintenance practices of utilities.
Since the isauance of GL 89-10, the NRC has been per-Effects of Solar Geomagnetically Induced Currents on Plant forming audits and inspections of licensee MOVs to eval-Electrical Systems. Dansient disturbances in the carth's uate compliance with GL 89-10. This implementation magnetic field caused by auroral currents can induce elec-requires a large effort on the part of both the licensees and the NRC because of the large number of trical potential gradients across the earth's surface.These potentials act like quasi-dc voltage sources impressed be, safe:y-related MOVs in each nuclear power plant. During tween the grounded neutrals of transformers at opposite the implementation and auditing efforts, additional MOV ends of power transmission systems. 'lo study the effects deficiencies in predicting electric motor performance have been identified.'lherefore,in 1993 the NRC started of these geomagnetically induced currents (GICs) on plant equipment, a plant-specific electrical distribution conducting experiments and analyses to resolve these system for a nuclear power plant was modeled using the newest deficiencies. Specifically, technical data bases are Electromagnetic Transient Program (EMTP). The model being developed to determine how the motor outputs are simulated on-line equipment and loads from the station affected by increased loadings and by increased tempera-transformer in the switchyard to the safety busses at 120 tures.This information is important since the entire MOV volts, to which all electronic devices are connected for performance is dependent on the efficiencies of these plant monitoring.
valve power sources.
The licensees are supporting a large program to develop a The EMTP analysis used the half-cycle saturation of the sis for wakaung mpant 6 pMormance wih station transformer (due to GIC) and studied the effects testing the MOVs. During 1993, the NRC exchanged on the voltage harmonic levels that were noted at various technical information with industry researchers and con-electrical busses. The results indicate that the emergency tinued reviewing the applicability and accuracy of their circuits appear to be more susceptible to high harmonics resuhs. In addition, exchanges of tethm, cal findings were due to normally light load conditions. Protective relays m de with agencies in the United Kmgdom and Germany.
(both electromagnetic and solid state relays) without an harmonic filter, which operate purely as a peak detector, The NRC efforts through contractor staff continues to be are vulnerabic to false high readings with GICs as low as one of the leaders in MOV diagnostic research. As a 50 amps present in the system. Based on these results, an result, current and past research results are contributing input side harmon c filter can be used on undervoltage to improving MOV reliability and plant safety in the protective relays, which senses an undervoltage or de-United States and abroad.
graded voltage condition for startmg the onsite diesel generators. The report includes other paramgtric studies 1.2.3.3 Engineering Standards Support on the subject and discusses potential harmome effects on the uninterruptible power system.
The national standards program is coordinated by the American National Standards Institute (ANSI). ANSI Equipment Operability (Mechanical). In the past, MOVs provides procedural guidelines to help ensure that partici-have had a poor performance history. Uccause of this poor pation in the private sector standards development performance, the NRC in 1989 issued Generic Letter process is sufficiently broad-based and that input from (GL) S9-10, " Safety-Related Motor-Operated Valve Tbst-individual Sterests are fairly considcred.The NRC partic-ing and Surveillance." The purpose was to ensure that ipation in.nis process is compatible with Office of Man-licensees take appropriate steps to ensure all safety-re-agement and Budget Circular A-119, dated October 26, NUREG-1266 1-12 j
- 1. Reactor Aging 1982, which provides policies for Federal participation in ance of concrete structures in nuclear power plants has the development and use of voluntary standards.
been good. However, there have been several instances where the capability of concrete structures to meet future The NRC staff is particularly active on ASME codes and functional and performance requirements has been chal-standards writing committees because portions of the lenged because of problems arising from either improper ASME Boiler and Pressure Vessel (11&PV) Code have, material selection, construction and design deficiencies, since 1971, been incorporated by reference into 10 CFR or environmental effects. Examples of some of the poten-50.55a to establish requirements for the construction, in.
tially more serious incidences include post tensioning service inspection, and inservice testing of nuclear power anchor head failures, leaching of concrete in tendon gal-plan t components. Section 50.55a is periodically amended leries, voids under vertical tendon bearing plates, contain-to update the references to include more recent versions ment dome delaminations, corrosion of steel tendons and of the ASME 11&PV Code. In 1993, work continued on rebars, water intrusion through basemat cracks, and leak-rulemaking, initiated in 1992, that not only would update age of corrosion inhibitors from tendon sheaths. Such l
the reference to the ASME Il&PV Code, but would for incidents indicate that there is a need at some nuclear l
the first time incorporate by reference the new ASME power plants for improved surveillance, inspection and Operations & Maintenance (O& M) Code, which provides testing, and maintenance to enhance the technical bases rules for inservice testing of pumps, valves, and snubbers.
for assu rance of continued safe operation of nuclear pow.
The proposed rulemaking would expedite implementa.
er plants.
tion of certain new ASME Il&PV Code requirements for qualification of personnel and equipment used to perform inservice nondestructive ultrasonic examinations on nu-The structural aging (SAG) program is addressing the clear power plant components.
aging management of safety-related concrete structures in nuclear power plants for the purpose of providing im-prov d technical bases for their continued service. The 1.2.3.4 Structural Integrity SAG program objective is to prepare documentation pro-As with all the structural integrity research, timely and viding (1) identification and evaluation of the structural effective transfer of the results to the NRC practitioners is degradation processes; (2) issues to be addressed under a high-priority activity. For the inspection procedures and nuclear power plant continued-service reviews, as well as technologies research program, an inter-office 'Ibchnical criteria and their bases, for resolution of these issues; Advisory Group has been established to provide input to (3) identification and evaluation of relevant inservice in-the research plans and to facilitate transfer of the re.
spection or structural assessmen t programs; and (4) meth-search results.
odologies required to perform current assessments and reliability-based life predictions of safety-related con-l During FY 1993, the research program provided both crete structures. To accomplish this objective, the SAG resource reports and direct support to the regulatory staff.
program has conducted activities under three major tech-Specifically, a document on the fundamentals of comput-nical ta sk areas: (1) materials property data base, (2) struc-er-based ultrasonic systems was prepared to serve as a tural component assessment / repair technologies, and resource on the technology and to describe the methods (3) quantitative methodology for contmued service deter-for characterizing computer based ultrasonic systems, mmations. Nearly 50 reports on these topics have been Additionally, a review was completed that will be pub.
Published smce 1991. Those published m 1993 covered fished as a supplement to the resource document to pro-surveillance and inspection of concrete structures; tests vide a systematic evaluation of the computer-based sys-on age and degraded concrete; data surveys; repairs for tem called P-SCAN.
damage; and reliability-based life assessment. The final seven reports will be completed in 1994.
Direct support was pmvided to the regulatory staff by researchers at PNL who assisted in evaluating inspections Recent experience suggests the possibility that corrosion to detect and characterize cracks that were discovered in effects may significantly degrade the margin that contain-the core shroud of a IlWRJIhis application posed difficul-ments have to accommodate accidents beyond their de-ties because of severe access limitations, and the ultrason-sign basis. Evidence of corrosion has been found both in ic system employed was a modification of a comput-IlWR Mark I containments and in PWR ice condenser er-based system for inservice inspection of piping. The containments. Additionalinformation is necessary to un-f PNL researchers were able to provide the benefit of their derstand the significant factors relating to corrosion, effi-considerable experience in developing and applying cacy of inspection, and capacity reduction to be able to unique inspection systems.
formulate future regulatory requirements that will ensure f
the continued availability of sufficient margins. Research l
Concrete structures play a vital role in the safe operation programs were initiated in 1993 to provide that informa-l of alllight-water reactor plants. In general, the perform-tion.
1-13 NUREG-1266
- 1. Reactor Ar,ing 1.2.3.5 License Rene val Regulatory Standards cern was expressed that the proposed rule would con-strain public comment on environmental issues at the A final rule (10 CFR liu t 51)concerning the environmen-time of license renewal review for an individual nuclear tal review for renewal of a nuclear power plant operating power plant. All comments are being considered in devel-license is under development. The proposed rule was oping the final rule, the generic environmental impact published for public comment in September 1991. Over statement, and other supporting documents. It is ex-120 comments were received on the technical analyses pected that the final rule and supporting documents will and certain pn>cedural aspects of the proposed rule. Con-be published in 1995.
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NUREG-1266 1-14
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- 2. STANDARD REACTOR DESIGNS 2.1 Engineering Issues for Advanced 2.13 Research Accomplishments in FY 1993 Reactor Designs 2.1.3.1 Qualification of Advanced Instrumentation and Control System Hardware The Oak Ridge National Laboratory (ORNL)is conduct.
2.1.1 Statement of Problem ing a study to identify functional and environmentalissues arising from the application of new technologies to the The safety evaluation of standard / advanced reactors in-instrumentation in both present and next-generation nu-clear power plants. Specifically, the program seeks an volves (l) new technologies, (2) new environmental condi-understanding of the technicalissuesinvolved m evaluat-tions, (3) new systems, (4) new design approaches, and (5) new requirements. New technologies include the expan-jng I ng-term properties and performance of advanced ded use of advanced digitalinstrumentation and control instrumentation and control (I&C) systems mtended for P ant upgrades and/or proposed for use m advanced l
systems and novel modular construction methods. New environmental conditions are reflected in the low-light-water reactors (ALWRs). Special emphasis wtll be given to identifytng vulnerabilities and enytronmental pressure operation of check valves for a planned 60-year operational period. New systems are found in the passive limitations that could be imposed on microproces-reactor designs where air-conditioning power is not need-s r-b sed systems m, nuclear plant enytronments. Initial studies have focused on protection systems and the I&C ed for accident response and the safety classifications differ markedly from present standards, thereby affecting employed m engmeered safety systems. The environmen-tal and functional issues studied thus far are reported in a probabilistic risk assessments (PRAs) and the fragility in-formation needed to perform these PRAs. New design draft report (NUREG/CR-5904). In this document evalu-approaches incorporate the use of experience data as a ation templates are presented. They were developed by method of seismic qualification and the climination of the assembimg reasonably complete configurations of safety operating basis earthquake from design. New require-channel components for the maj,or ALWR designs. The ments deal, for example, with NRC goals for containment templates provide for evaluatmg the unpact of environ-performance under severe accidents and the ability to niental stressors affecting these components and inter.
faces. Functional issues considered in the evaluation m-withstand interfacing-system loss-of-coolant accidents, Research is needed to support technical positions in safety clude distributton of functions, sources, supply and evaluation reports prepared by the NRC staff for the Final distribution of electrical power, calibration and testmg Design Approval, the Design Certification, and the Com-capabilities, and attributes of failure predictions. The bined License for standard / advanced reactors dealing application and acceptance of digital computers m reactor with these issues.
protection systems are also reviewed in light of current mdustry standards.
2.1.3.2 lbchnical Basis for Regulatory Guidance on Electromagnetic Interference Issues 2.1.2 Program Strategy ORNL is developing the technical basis for regulatory guidance to address upsets and malfunctions in I&C sys-He objective or strategy of engineering research is to tems caused by electromagnetic and radio-frequency in-verify the safety and quantify the margins in standard /ad-terference (EMI/RFI) and power surges. The concern vanced reactors for those features or criteria substantially stems from EMI/RFI and power surges affecting digital different from currently operating plants and to ensure I&C systems at nuclear power plants and other process that NRC goals and policies are implemented successful-industries.The technical basis centers around good engi-ly. 'Ib deal with these issues, analyses, evaluations, data neering practices to ensure that sufficient levels of elec-collections, and limited testing are planned. In some tromagnetic compatibility (EMC) are maintained cases, research is used to independently confirm or between the nuclear power plant's electronic and electro-modify vendor / utility proposals for design, fabrication, mechanical systems. Good EMC design and installation construction, and inspection. In other cases, particularly practices are recommended to control interference in the passive designs, research serves to demonstrate sources and their impact on nearby circuits and systems in acceptable design basis and severe accident response, for accordance with IEEE Std.1050-1989, " Guide for Instru-example, for structural performance of the AP600 con-mentation and Control Equipment Grounding in Gener-tainment during passive containment cooling. In still oth-ating Stations."These good EMC practices include circuit er cases, research is used to define acceptable standards layouts, terminations, filtering, grounding, bonding, for new features and requirements.
shielding, and adequate physical separation. In addition, a 2-1 NUREG-1266
- 2. Standard Reactor Designs test and validation program is being developed to outline expected that they will be used for the passive designs as the EMI/RFI and surge withstand capability (SWC) test well.
and the associated test methods to be followed to ensure that I&C systems are capable of performing their in-2.1.3.5 Experience-Based Seismic Qualification tended functions.The test and validation program isbeing In its Utility Requirements Document for ALWRs, the developed around the EMI/R FI test critena from Mihtary Electric Power Research Institute (EPRI) has proposed j
Standard (MII-STD)-461, Requirements for the C.on-the use of experience as a method of seismic qualification trol of Electromagnetic Inter erence Emission and Sus' as an applicable substitute on a case-by-case basis for r
ceptibility, the associated EMI/RFI test methods ex-more traditional tests and evaluations. An expert panel tracted from MII-STD-462,
, Measurement of l
established to assess the viability of the experience-based j
Electromagnetic Interference Characteristics, and the method has recommended the use of a graded approach in j
SWC guidelines contained m lEEE Std. ( 62.41-1991' which equipment categories are placed in one of three Recommended Ilractice on Surge Voltages m Low-Vo!-
groups. In the first two groups, the use of experience data r
tage AC Power Circuits. Currently, the electromagnetic is poss ble but seismic capacities are different, and the environment m a typical commercial nuclear power plant attributes for membership in these two groups are also is undefmed; hence, electnc and magnetic spectral receiv-different. In the third group, the use of experience data is ers need to be placed at various nuclear power plants.
not allowed. Equipment categories likely to be placed in Long-term, unattended measurement data wdl be col-the first two groups are horizontal and vertical pumps; l
lected with these spectral receivers and used to provide motor-operated, manual, and check valves; thermal ele-the NRC with a realistic assessment of the probable ambi-ment assemblies; diesel generators; transformers; and ent electrorapnctic environment m nuclear power batteries on racks.
plants.
2.1.3.6 Containment Performance Goals 2.1.3.3 National Codes and Standards In support of the NRC Severe Accident Policy Statement, as it applies to A13VRs, work began on development of Work initiated in FY 1992 on assessing current national puf rm nee requirements for containments under se-codes and standards, such as the ASME Hoiler and Pres-vere accident conditions. Determmistic cntena will be sure Vessel Code, is continuing through FY 1993. The stablished for both steel and concrete contamments em-infortnation derived from this work will be integrated into phastzing standards for local and global strams and defor-the advanced reactor design approval / licensing process.
mations. For these determmistic cntena, probabilistic Also under the guidance of this program, Canadian and m dels wdl be constructed to allow compan, son with the other international quality assurance (Q A) standards are conditional containment failure probability of 0.1 spect-being eva!uated for compliance with Federal QA stan-fied by the Commission for implementing safety goals.
dards*
Particular attention will be given to evaluating the severe accident performance requirements for containments 2.1.3.4 Evaluation of Low-Pressure Piping for Intersys-mentioned in the EPRI Utility Requirements Document.
tem LOCA Development of evaluation criteria for a new design goal 2,2 Systems Performance of Advanced for A13VRs was completed in FY 1993.The intent of the Reactors new design goal is that low-pressure piping attached to the reactor coolant loop withstand reactor coolant pressures 2.2.1 Statement of Problem and temperatures.The condition m question, which could follow from multiple valve failures, is important because The Commission has issued a policy statement on the for certain postulated events it can lead to rapid core regulation of advanced nuclear power plants (51 FR damage and the release of radioactivity outside the con-24643) that states that the NRC will review and comment tainment. The potential event is called an intersystem on new design concepts with special emphasis on vendor loss-of-coolant accident and is being treated as a severe test programs for confirming the performance of novel accident. Because of the low frequency of occurrence of safety systems. It also encourages early interaction with this potential event, the performance goal is to achieve a applicants. As part of this program, the NRC will develop, failure probability in the low-pressure piping of about 10 review, and implement advanced reactor safety and policy percent. A probabilistic methodology to estimate failure issues in the ongoing NRC review of advanced reactor was applied to both carbon and stainless steel piping and concepts. Indepth independent analysis will be performed other piping components-including flanges, valves, as necessary to verify that advanced reactor designs have pumps, and heat exchanger tubes. These results will be the potential for enhanced margins of safety and that used in the Design Certification for the advanced boiling appropriate means will be used to accomplish their safety water reactor (ABWR) and the System 80+, and it is functions.
NUREG-1266 2-2
- 2. Standard Reactor Designs 2.2.2 Program Strategy for handling proprietary information. This screening re-vealed that the best candidate was the Rig of Safety Research programs have been initiated to support the
. Assessment (ROSA) Large-scale Test Facility in Japan certification and licensing of advanced reactor designs Atomic Energy Research Institute (JAERI). lb confirm being developed by the nuclear industry and the Depart-these initial results and to determine the extent of modifi-ment of Energy. Several different designs are being con-cation necessary to simulate the AP600, the Idaho Nation-sidered for certification under the newly formulated al Engineering Laboratory (INEL) was contracted to per-10 CFR Part 52. These designs fall into two groups. The form a comparative study between ROSA and AP600 first group consists of four evolutionary and passive ad-using the RELAP5 code.
vanced LWR types: the advanced boiling water reactor (AllWR), the CE System 80 +, the AP600 advanced pres-A comparison between the existing ROSA facility and the surized water reactor, and the simplified boiling water AP600 design showed that ROSA did not contain the key reactor (SHWR).The second group consists of four future components important for safety response of the AP600.
advanced reactor types: the Process Inherent Ultimate h was not obvious how much hardware modification to the Safety (PIUS) reactor, the Canada Deuterium Uranium ROSA facility would be needed to simulate the AP600.
(CANDU 3) heavy-water-cooled reactor, the Advanced The fidelity of simulation must be balanced against the Liquid-Metal-Cooled Reactor (ALMR), and the Modular associated cost. The fidelity should be high enough to liigh Femperature Gas-Cooled Reactor (MHTGR)-
result in a facility capable of producing data for code i
assessment covering the major AP600 phenomena in the The two evolutionary LWR designs (ABWR and System correct sequence. At the same time, the cost and the 80 + ) have been judged to be similar enough to the cur-schedule have to be affordable. To make an optimum rent generation of LWRs so that little additional research choice, four levels of modification in progressively more is needed to support their certification or licensing; the extensive stages were considered.The first level of modi-four nonconventional reactor design concepts have not f cation was the absolute minimum, and the fourth level reached the point where certification is expected in the was the most inclusive among the four levels.
near term. Consequently, the current emphasis of the advanced reactor research program is on developing the mformation needed to support the certification of the The cornparative analyses of ROSA and AP600 with the RELAP5 code for selected accident scenarios led to the passive AP600 and SilWR reactor designs and making appropriate modifications of existing regulatory assess-conclusion that the optimum choice was the fourth level.
ment methods to accommodate the review of the unique This included two core makeup tanks (CMTh) with appro-features of these reactors.
priate pressure balance lines, a passive residual heat re-moval system with simulated secondary cooling, automat-ic depressurization system with stages 1 through 3 on top 2.2.3 Research Accomplishmenis in FY 1993 of the pressurizer and stage 4 on the hot leg, minimization of the pump loop seal lengths, a properly scaled AP600 2.2.3.1 NRC Confirmatory Safety System 'Itsting in pressurizer with a surge line, installment of appropriate Support of AP600 Design Review upper-head flow paths, and an incontainment refueling Westinghouse Electric Corporation submitted the Ad-water storage tank. Since ROSA has only one cold leg in vanced Passive 600 MWe (AP600) nuclear power plant e ch loop, CMT cold leg pressure balance lines are con-design to the NRC for design certification. RES is pro-nected to the same cold leg for most transients when cceding to conduct confirmatory testing of AP600 safety asymmetry between the two CM'It is not expected, but systems to help the NRC staff evaluate the safety of the connected to a different cold leg for nonsymmetric tran.
AP600 reactor systems. NRC confirmatory safety system sients, such as a break m a pressure balance ime or a break testing is not required for design certification but would in a direct vessel mjection lme.
provide additional technical bases for assessing NRC com-puter codes, which are then used to assist NRC in reach.
The chosen level of facility modification is being implem-ing licensing decisions.
ented by Sumitomo Heavy Industries, which constructed the ROSA facility and has been maintaining and operat-For confirmatory testing, it was determined that the most ing it for the past several years as a contractor to JAERL cost-effective route was to modify an existing full-height, The facility modification will be completed by February j
full-pressure test facility rather than build a new one.
1994 and a series of tests performed later in 1994.
Thus, all the existing integral effects test facilities, both in the United States and abroad, were screened to select the As a confirmatory testing program, the ROSA /AP600 best candidate. The criteria for the initial screening in-testing will cover not only design basis accidents but also cluded the size, facility configuration similarity, availabil-beyond design basis accidents, which vendors may not be ity schedule, willingness to share the cost, and the ability required to address.There will be some counterpart tests to enter into a confidential agreement with Westinghouse among ROSA, SPES-2, and OSU tests. Test data from 2-3 NUREG-1266
- 2. Standard Reactor Designs different scale facilities will help predict what will occur in way that would assist the NitC with its preapplication a full-scale AP600 reactor.
review for design certification. The selection of certain transients and postulated accidents for consideration in The status of the NRC containment response codes (i.e.,
the safety analysis will be crucial in determining what CONTAIN and COMMIX) that will be used for evaluat.
acceptance esiteria have to be demonstrated for CANDU ing the performance of the AP600 passive containment 3; the criteria will be different from those for LWRs. Plant l
cooling system design are discussed in Section 5.2.3.4, systems that are identified as important to safety must be
" Severe Accident Codes."
given early attention to ensure that they receive appropri-ate seismic and environmental qualification. Identifying Scaling analyses of the AP600 core makeup tanks and the need for operator actions to mitigate accidents will passive residual heat removal system were completed and also lead to carly decisions regarding automatic versus findings sent to NRR for their use, manual actions.
An assessment of existing and planned data bases for 2.2.3.2 SilWR Test Facility CANDU 3 was performed by the staff with assistance Purdue University was awarded a 3-year contract on from ORNL, INEL, and Sandia National Laboratories.
July 26,1993, to build a test facility for an advanced reac-
'lhis assessment provides an early identification of areas tor design, the Simphfied lloiting Water Reactor (SIlWR).
requiring long lead times to plan additional test programs
'Ihc objective is to provide data to assess the capabilities of that are needed for completeness. The assessment also the NRC's computer codes to analyze the SilWR. The provides the NRC with a basis for planning its resource Purdue test facility will have all the necessary components requirements for confirmatory research and analysis sup-and systems scaled from the SilWR design. Included are a port for design certification of CANDU 3.
vessel with electrically heated fuel rods, upper and lower drywells, suppression pool, gravity-driven cooling system CANDU reactors are somewhat overmoderated and con-(ODCS), passive containment cooling system (PCCS),
sequently exhibit a positive reactivity coefficient for cool-isolation condenser system (ICS), drywell and wetwell ant voiding, unlike LWRs that have been licensed by the sprays, piping and valves, and instrumentation.
NRC. 'this is a well known characteristic of the design, and the CANDU reactors are equipped with a second,
'lhe facility is a low-pressure, reduced-height facility. It is rapid-acting, independent shutdown system to provide designed for 150 psia, which is adequate for investigating safety margin. Nevertheless, it is the NRC's practice to the integral performance of the GDCS and PCCS. The study postulated accidents of very low probability if they height of the Purdue facility is 1/4 of the SilWR height have the potential for serious consequences. Therefore, a and its volume is 1/400 of the SilWR volume.The aspect series of calculations was performed by the NRC staff and INEL for several transients for which it was assumed that ratio of the facility defined as height scale / diameter scale is 2.5. As a result of the reduced scale, events will occur in neither of the shutdown systems worked. These calcula-the facility on a time scale twice that calculated for the tions provide an indication of the magnitude and time SilWR, The facility is scheduled to be built by December scale of the resulting power excursion that will help the 1994. Approximately 50 tests will be performed by April NRC with decisions about the licensability of this design.
1996 to cover a broad spectrum of loss-of-coolant acci-2.2.3.4 Ilurnan Reliability dents and transients. At least 10 of the tests will be com.
pleted prior to the completion of the final safety evalua-A study is ur'Jer way to develop methods for assessing the tion report on SilWR, which is scheduled for July 1995.
impact risk of changes in human performance a
promp'.cd by the introduction of advanced digital displays 2.2.3.3 Support for CANDU 3 Preapplication Licens-and. controls. Research to establish a technical basis for mg Review n> cessary shift staffing to support safe operations for ad-vanced control room designs is being planned for initia-A review of the regulation of CANDU reactors in Canada ti n in FY 1994.The research will be based on workload was performed by the Argonne National Laboratory.That and task analysis studies.
review summarized Canadian regulations as they are applied to CANDU. It described safety systems in the CAND U designs and their relationship to Canadian regu-2.3 Advanced Reactor Risk Analysis lations.The review also identified major NRC regulatory requirements that have a bearing on unique features of 2.3.1 Statement of Problem the CANDU design.
Probabilistic risk analysis (PRA) has been shown to be a A systems analysis of the CANDU 3 design was per-systematic and comprehensive method for identifying and formed by ORNL The purpose of this work was to classify evaluating the effectiveness of safety improvements pro-event sequences, plant systems, and operator actions in a posed to reduce the likelihood and consequences of nu-NUREG-1266 2-4
l L
- 2. Standard Reactor Designs 1
clear power plant accidents. PRAs are required as part of
'lhe most limiting design basis accident evaluated is re-the license applications of all advanced reactor designs.
quired by 10 CFR Part 100 and is derived from the 1%2 RES supports the Office of Nuclear Reactor Regulation report TID-14844, " Calculation of Distance Factors for (NRR) review of these applications by performing inde-Power and 1bst Reactor Sites," which postulates the re-pendent assessments of accident-initiating-event fre-lease of the entire core inventory of noble gases and 50 quencies and the reliability of key systems in these percent of the core inventory of iodine fission products designs.
from the core into the containment. This flD source term"is used for evaluating the suitability of the reactor design as well as the site. Present regulatory guidance, 2.3.2 Program Strategy reflected in Regulatory Guides 1.3 and 1.4, stipulates that this source term is instantaneously available for release 1he research effor! u this area is focused on the assess-from the containment to the environment and specifies ment of the frequencies of challenges to an advanced that the iodine would primarily be in elemental form.The reactor design and the ability (reliability) of mitigating TID-14844 source term, originally intended for site evalu-systems to cope with such challenges and prevent damage ation purposes, has also been applied to many aspects of to the reactor core. The program accomplishes this by plant design.Thesc include requirements for fission prod-i defining the set of potential challenges, estimating the uct cleanup systems, such as sprays and filters, allowable I
frequencies of these challenges, and estimating the reti.
containment leak rate, control room habitability, equip-ability of systems included in the design in response to the ment qualification, and others.
j challenges.
(
This source term is associated with a severe reactor acci-dent, since only major core degradation and melting could
)
2.3.3 Research Accomplishments in FY 1993 result in the release of such large quantities of fission products. However, the present formulation, while pro-The advanced passive reactors have engineered safeguard viding a high level of protection for plant systems, is now systems that maximize the use of passive devices such as recognized to be incompatible with present research find-nitrogen-powered accumulators, natural circulation flow, ings. As a result, a rigid application of theTID source term and gravity-driven safety injection. They do not rely on may not permit the best engineering solutions on some active systems such as ac-powered equipment, although aspects of future plant design.
certain valves may require stored energy (e.g., battery power) to change state. These passive designs are ex-pec*cd by the designers to both mcrease safety and de-2.4.2 Program Strategy crease cost as a result of their simplified design. However, On May 25,1988, the staff presented to the Commissica because of the lack of actual working experience with the an " Integration Plan for Closure of Severe Accident Is-design and because of uncertainties in the modeling of sues," S ECY-88-147. This plan discussed major elements i
processes such as natural circulation, there are uncertain-relating to closure of severe accident issues, including ties in th: performance of the engineered safeguard sys-severe accident research efforts and related activitics in tems.
siting, emergency planning, and potential changes to ex.
isting regulations as a result of severe accident research The passi te reliability project, presently focusing on the findings. On October 4,1990, the staff presented to the Westinghouse AP600 design, is developing a candidate Commission a " Staff Study on Source Ibrm Update and methodology for quantifying the uncertainty distribution Decoupling Siting from Design," SECY-90-341. This in the core damage frequency arising from uncertainty in plan discussed the staff proposal to decouple reactor sit-the modeling of the natural processes. The project will be ing from source term and dose calculations and to prepare completed in FY 1994.
an update and revision of the source term given in TID-14844. On April 10,1992, the staff presented to the Commission its " Revised Accident Source 1brms for 14 Regulatory Applicalion of New Light Water Nuclear Power Plants," SECY-92-127. This Source Terms p pu pnwided a draft f a revised soum ynn. On June 12,1992, the staff presented to the Commission its proposed " Revision to 10 CFR Part 100, Revisions to 10 2.4.1 Statement of Problem CFR Part 50, New Appendix B to 10 CFR Part 100 and New Appendix S to 10 CFR Part 50," SECY-92-215.This Potential accidents are evaluated during reactor licensing paper presented proposed rules to revise the reactor sit.
as part of the Commission's defense-in-depth policy. Cer-ing criteria. On July 28,1992, the Commission announced tain accidents, referred to as " design basis accidents," are the availability for comment (57 FR 33374) of a draft postulated to occur and their consequences must be report on " Accident Source 'Ibrms for Light Water Nu-shown to be acceptable.
clear Power Plants," NUREG-1465. On September 28, 2-5 NUREO-1266
=_. __ _
- 2. Standard Reactor Designs 1992, the Commission issued an advance notice of pro-quences Predicted by the MELCOR Code," dated posed rulemaking on " Acceptability of Plant Performance September 1993.
for Severe Accidents: Scope of Consideration in Safety Regulations" (57 FR 44513) Revised accident source NUREG/CR-5901,"A Simplified Model of Aerosol e
terms, as well as other severe accident considerations, are Scrubbing by a Water Pool Overlying Core Debris l
to be incorporated into this rulemaking effort.
Interacting with Concrete," dated November 1993.
2,4.3 Research Accomplishments in FY 1993 2.4.3.1 Update of Siting Regulations
'the Commission's reactor site criteria (10 CFR Part 100)
In FY 1993, staff efforts continued on updating 10 CFR l
require that an accidental fission product release from the Part 100, " Reactor Site Criteria." A proposed rule to re-l core into the containment should be assumed to occur and vise Part 100 was issued in the Federal Register (57 FR that its radiological consequences should be evaluated.
47802) on October 20,1992, for a 120-day comment peri.
The criteria for the release into the containment are od. The rule proposed to decouple calculation of the derived from the 1%2 report, TID-14844, which assumed exclusion area distance from source term and dose calcu-an instantaneous release of fission products. Although lations by specifying a minimum acceptable exclusion area l
this source term has long been included in the Commis-distance and by stating population density criteria as well.
sion's regulations for siting, it has traditionally had a In addition, an update of seismic considerations proposed greater effect on plant design than on siting.
to incorporate probabilistic as well as deterministic meth-ods. The comment period, extended twice, expired on l
Since 1%2 the NRC has attained a better understanding June 1,1993. Extensive comments both domestic and of the timing and nature of the fission product release and foreign, were received. The staff is analyzing comments has identified a number of areas subject to regulation that received and is examining further options in this regard.
may benefit from changes introduced as a result of source term and severe accident research. In FY 1993, work 2.4.3.2 Emergency Planning Regulations cont, ued on a replacement toTID-14844. The comment m
i period for a draft report, " Accident Source '1brms for In May 1993, a proposed rule was published in the Federal Light-Water Nuclear Power Plants" (NUREG-1465),
Register (58 FR 29795) on the emergency planning licens-which was issued in July 1992, expired in December 1992.
ing requirements for independent spent fuel storage faci-Comments were solicited and received from an mterna-lities and monitored retrievable storage facilities.
tionally recognized group of experts as well as from the public. Afinalversionof NUREG-1465isinpreparation.
In J une 1993, a proposed rule was published in the Federal In addition, in support of this effort, the following docu-Register (58 FR 34539) on revised emergency plannmg that ments were issued:
would update and clarify ambigu ties that have surfaced m NUREG/CR-5950," Iodine Evolution and pli Con.
the implementation of the Commission's emergency e
trol," dated December 1992.
plannmg exercise requtrements.
NUREG/CR-5966, "A Simplified Model of Aerosol In March 1993, a petition for proposed rulemaking by Removal by Containment Sprays," dated June 1993.
VEPCO was published in the federalRegister. In this peti.
tion VEPCO proposed that the NRC amend its regula-NUREG/CR-5942," Severe Accident Source'Itrm tions to change the frequency of emergency exercises at Characteristics for Selected Peach Ilottom Se-nuclear power plants from annual to biennial.
NUREG-1266 2-6 1
PART 2--NUCLEAR SAFETY RESEARCII-REACTOR REGULATION SUPPORT
1
- 3. PLANT PERFORMANCE 3.1 Statement of Problem tor. Third are integral system tests that are used to evalu-ate the code predictions of a complete reactor.The results A wide range of reactor plant design variations exists in of these tests provide feedback to correct the code and our l
the United States, and the safety of these plants must be understanding of the transients.
ensured over a wide range of normal and abnormal oper-ating conditions. He NRC is required to independently li assess licensees safety analyses and performance m de-3*3 Research AccomE shments in FY signing, constructing, and operating a reactor with respect 1993 to the safety of the public for the complete spectrum of credible operating conditions and events.
33.1 Reactor Safety Experiments NRC's task is difficult because straightforward testing of Experiments are being conducted on decay heat removal all transients in all plant design variations would not be by natural circulation at North Carolina State University technicMiy and economically feasible. On the other hand, in a facility that is a scaled model of a Westinghouse straightforward and exact theoretical analyses of a reac, pressurized water reactor. It is 1:9 in height with a 1:700 tor's thermal hydraulic behavior is not possibic because volume scale and uses freon as the working fluid. The 2
mass, energy, and momentum exchanges take place over facility includes a secondary system to provide representa-complicated interfaces between reactor components, wa.
tive plant behavior. Parts of the facility are constructed ter, and steam and because of the moving mechanical with plexiglass to allow visual observation ofinternal con.
interfaces in pumps and the extensive baffle arrange.
ditions. Facility construction was underwritten by several ments of steam generators in the primary loops, utilities and is used extensively for the training of utility personnel and university students. The NRC is conduct.
As a result, the NRC rnust use available experimental ing decay heat removal experiments under conditions of data to validate analytical models for evaluating design reduced coolant inventory m the primary system, using basis accidents, the safety implications of actual events in the visual observation features of the experimental facil.
operating reactors, and hypothetical transient scenarios ity.
s determined to be major contributors to risk as a result of probabilistic nsk assessment studies.
Experiments are also being performed at the University of Maryland in a scaled experimental facility that simulates a Babcock and Wilcox reactor and is 1:4 in height with a 3.2 Program Strategy 1:500 volume scale.This facility was originally constructed under an NRC contract to study small-break loss-of cool-A dual analytical and experimental approach is used to ant accidents and, following successful completion of this achieve a firm technical understandingof the thermal-hy.
program, its mission was shifted to the current study of the draulic behavior of the reactor.The NRC starts by simu.
natural circulation of steam under severe accident condi.
lating the actual reactor's continuous flow of heat and tions. For these tests, sulfur hexafluoride is being used as fluids with a computerized model consisting of many dis-the working fluid.The current program will be completed crete cells exchanging mass, energy, and momentum at in FY 1994, each small but finite time step. Physical laws are used when possible to calculate all these exchanges. Empirical.
33.2 Safety Code Development and Mainte-ly derived formulas that are obtamed from expenments "U"CU are used as necessary to account for such complex effects as friction between vapor and liquid.The calculations are The third semi-annual international thermal-hydraulic made for each time step and for each cell. The reactor code applications and maintenance program (CAMP) models interact in a tightly coupled manner at each time meeting was held during October 20-22,1993,in Santa Fe, step.
New Mexico. There are now 14 member countries in CAMP, each of which had representatives at this meeting, Our reliance on the computer codes to provide predic-in addition to representatives from the NRC and its code tions of reactor response with acceptable uncertainties development contractors.
depends on three levels of experiments and comparis(ms of experimental results with code predictions. First are Cash contributions made by, or on behalf of, these 14 basic experiments to derive empirical formulas for deter-member countries in the amount of $390,000 annually are mining phenomena within each cell. Second are used to supplement the code development and assess-separate-effect experiments to test the code's predictions ment programs funded by the NRC. As part of their for a single, complex component such as a steam genera-agreements, the members provide code assessment bl NUREG-1266 i
- 3. Plant Performance studies or other noncash contributions to assist NRC's tic steam separator, spherical accumulator model, and assessment of code applicability, scalability, and uncer-improved IRWST model, tainty when applied to nuclear power plant safety. He j
codes covered by the agreement are REL.AP5/ MOD 3, Improvements were also made to the RAMONA code to TRAC /PWR, and TRAC /BWR. Papers presented at the perform stability analyses in SBWRs. These improve-meeting discussed work on these programs funded by ments include modeling of the isolation condenser, chim-NRC and CAMP member contributions, CAMP member ney component, boron transport and reactivity feedback, activities, code assessment performed by CAMP mem-standby liquid control system, and flow-dependent loss bers, and code activities of United States code users.
coefficients for low flows. In addition, the modeling of all j
balance-of-plant components is completed. This capabili-ty is needed to analyze stability in both SBWRs and exist-Improvements were made to RELAPS in support of the ing BWRs.The RAMONA code was made operational on
)
AP600 and SBWR design reviews. The following set was a workstation used at the NRC.
completed in FY 1993: energy balance out the break, j
condensation in the presence of noncondensables, con-The manual for the'IT(AC-P code that is being used to tainment modeling (coupling RELAP with CONTAIN analyze large-break LOCA scenarios in the AP600 design code), code speedup to analyze long-term transients, level has been prepared for publication.The TRAC-P code has tracking, boron transport (moving wave front), mechanis-been made operational on workstations used at the NRC.
NUREG-1266 3-2
l
- 4. HUMAN RELIABILITY 1
4.1 Statement of Problem plants. Measures and supporting documentation for a training effectiveness evaluation method will be included A large fraction of all safety-related events reported at in a report to be issu ed in FY 1994. Data analyses are being nuclear power plants and among nuclear materials licens-incorporated into a final report (NUREG/CR-6122) for a ces continue to involve human performance. Methods project on the factors that are considered when making and data are needed to identify, systematically set priori-decisions on operations staffing and on how staffing re-ties for, and suggest solutions to human performance lates to safe startup, shutdown, and operation of nuclear issues in nuclear operation and maintenance during nor-power plants.
mal, transient, and emergency situations.
A study to establish a technical basis for minimum shift t
staffing for operational crews at nuclear power plants 4.2 Program Strategy based on workload and task analysis was initiated. A hand-book on the effects of environmental factors on human The human reliability research program has three objec-performance was completed and will be issued as tives: (1) to broaden NRC's unders'tanding of human per-NUREG/CR-5680 in 1994 for use by nuclear power plant formance and to identify causes of human error; (2) to inspectors. Two reports concerning training for respond-accurately measure human performance for enhancing ing to accidents are being prepared. These reports de-safer operations and precluding critical errors; and (3) to scribe decisionmaking and stress coping skills that may be develop the technical basis for requirements, recommen-needed to respond to an accident situation, as well as dations, and guidance related to human performance.
potential training approaches for developing those skills.
The human reliability research program was divided into An effort to identify measures for better characterizing four interrelated program elements: (1) personnel per.
the quality of personnel performance in an operational formance, (2) human-system interfaces, (3) organization.
setting was initiated. These measures and methods could al factors, and (4) reliability assessment. The purpose of be used for evaluating the effects on safety of changes in the personnel performance element is to develop en.
system interfaces, particularly changes from analog to hanced methods for collecting and managing personnel digital interfaces.
performance data and to improve understanding of the effects of personnel performance on the safety of nuclear 4.3.2 Human-System Interfaces operations and maintenance. In addition, personnel per-formance research will broaden the understanding of 11um n-system m. terface research m. eludes NRC partici-such factors as staffing, qualifications, and training that p tion m the Organization for Economic Cooperation and influence human performance in the nuclear system and Development (OECD) Halden Reactor Proj,ect, which is will develop information necessary to reduce the negative a multifaceted program that includes verification and vali-impact of these influences on nuclear safety. Research in dation of digital systems, man-machme mteraction, and the human-system interface element will provide the surveillance and support systems for advanced control technical basis for guidelines and criteria to evaluate the ro ms. Specific NRC research needs were identified m interface between the system and the human user from the form of lessons-learned reports on (1) methods and the perspective of safe operations and maintenance. Or-tools for the development, verification, and validation ganizational factors research will result in the develop-of safety related software, (2) experience with develop-ment of tools for evaluating organizational issues within ment and quality assurance of software systems at the the nuclear industry. And, lastly, the reliability assess-Halden Project, (3) man-machme mteraction with ment element includes multidisciplinary research that computer-based systems, (4) test and evaluation activities will integrate human and hardware considerations for n man-machme mteraction with computer-based sys-evaluating reliability and risk in NRC licensing, inspec-tems, (5) recommendations on advanced control rooms tion, and regulatory decisions.
based upon the Halden Proj,ect's experience,(6) coordina-tion and integration of computer-based operator support systems, and (7) reports containing information related to 4.3 Research Accomplishments in FY human reliability. These reports will serve as part of the 1993 technical bases for NRC guidelines. The lessons-learned report on test and evaluation activities was completed; it summarized over 10 years of experience with various acti-4.3.1 Personnel Performance vities on computer-based eperator support systems.
Work continued on the development of a method toassess Research continued to evaluate the positive and negative the effectiveness of training programs at nuclear power attributes of standards and computer-aided-software 4-1 NUREG-1266
- 4. Iluman Reliabbty engineering tools for use in the certification of high-433 Organizational Factors integrity sof tware for nuclear power plant safety systems.
Research also continued to develop and test a comput-Research on organizational factors developed draft,inte-cr-aided software engineering tool for assessing the de.
grated research products that could be considered for use gree of functional diversity in software that performs safe.
in regulatory applications such as risk-based inspection ty functions. A project, cosponsored by EPRI, to develop and diagnostic evaluation. Additional research continued guidelines for the verification and validation of expert to develop alternative quantification methods for incor-systems is nearing completion. Research continued on a porating the influence of organizational factors into PRA.
project, also cosponsored by EPRI, on verification and It was discovered that an analysis of organizational factors validation guidelines and quality metrics for digital can identify dependent failures across systems in a se-high-integrity systems.
lected work process at a plant.
'the NRC is reassessing the necd for continuing research
'ihe project to independently evaluate, test, and improve rganizational factors, takm, g tnto consideration the n
upon verification and validation guidelines for use in the remaining obstacles to incorporating organizational fac-audit of computer-based safety systems completed a field t rs mto risk.
test of the verification guidelines at a developer's site.
'lhe NRC, assisted by the NationalInstitute of Standards 43.4 Reliability Assessment and 'li'chnology, sp<msored a Digital Systems Reliability Work is nearly completed on collecting, cataloguing, and and Nuclear Safety Workshop. 'Ihe purposes achieved by st ring in a computerized library estimates of probabilities the workshop were to (1) provide feedback to the NRC f perator error and hardware failure. Because one of from outside experts regarding potential safety issues, the largest contributors to risk is operator cognitive error, proposed regulatory positions, and research associated research continued, m order to gather data on cognitive with application of digital systems in nuclear power plants performance, on an effgrt to validate a computer simula-and (2) continue the indepth exposure of the NRC staff to tion model of human cognitive tasks during accident se-digital system design issues related to nuclear safety by quences. The data were gathered from operating crews discussions with experts in the state of the art and practice responding to two simulated accident scenarios on tram-of digital systems. Proceedings of the workshop will be ing simulators.
issued early in 1994.
Research was initiated to analyze information from the Research was initiated to ensure the comE eteness of the simulator portion of the NRC-admmistered operator re-l techm. cal bases for regulatory requirements intended to qualification examinations. Estimates from this source ensure the integrity of safety-related software. These may provide empirical error rates for use in a nuclear technical bases wdl be used to develop guidelines that may power plant PRA. Research continued to evaluate the risk be used m developmg regulatory positions on software.
impact of replacing existing nuclear analog systems with l
digital systems.
Draft guidelines prepared for advanced control room de-sign reviews were subjected to internal and peer review.
For several years the NRC has been developing reliability
'Ihe draft guidelines are built on the validation of guide-analysis tools for use with PRA to improve the technical lines available from other industries, including the aero-bases of selected requirements in technical specifications.
space and defense industries. The draft guidelines were
'lhese tools can evaluate risk technical specification prepared in paper form and computerized (interactive) changes such as (l) surveillance test intervals, (2) allowed form. The paper form will be issued as NUREG/
outage times, both during operation and during shut-CR-5M)8. 'lhe interactive version of the guidelines was down, (3) action statements that require shutdown, demonstrated to the NRC staff. The final version is being (4) technical specification defenses against dependent developed for an application through Windows software, failures, and (5) scheduling preventive maintenance. Al-most all the methods are completed, and both detailed Wo meetings with subject matter crperts were held to technical reports and a handbook to guide NRC reviewers assess the feasibility of conducting NRC human factors through the use of these methods are in preparation. A regulatory research at available and potentially available Research Information Letter will be issued early in FY facilities.
1994.
NUREG-1266 4-2
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c r f
- 5. REACTOR ACCIDENT ANAINSIS 5.1 Reactor Risk Analysis was initiated in 1989 to examine the potential risks of accidents initiated d u ring low-power and shutdown modes 5.1.1 Statement of Problem of operation. Phase 1, completed at the end of 1991, was a coarse screening analysis of all operational modes (other Probabilistic risk analysis (PRA) has been shown to be a than fun [mn) for one HWR and one PWR to provide systematic and comprehensive method for identifying and suppon for the Office of Nuclear Reactor Regulation's evaluating the effectiveness of safety improvements pro _
(NRR) regulatory analysis and to guide the Phase 2 effort.
posed to reduce the likelihood and consequences of nu.
A significant finding was that the traditional concept of clear power plant accidents. PRA is used by the MtC staff technical specification modes of operation does not ade-in a number of ways, including for evaluating the level of qu tely delineate plant operating boundary conditions safety at selected operating plants; for assessing the mar.
(states) needed for risk analyses, lhe Phase 2 effort has gins of safety in current requirements in light of the Com.
concentrated on a specific operating state for each of the mission's safety goal policy; for monitor N plant perform-two plants, selecting the potentially highest risk operating ance; and for identifying potential impmvements in state based on the Phase I results. In addition, a simplified equipment or operator reliability.
an lym of potentialin plant and olfsite accident progres-sion and health consequences of such accidents has been performed and provided to NRR in support of their regu-5.1,2 Program Strategy latory activities, as documented in NUREG-1449. The results of Phase 2 will be published as NUREG/CR re-The reactor accident risk analysis research effort is ports early in FY 1994, applied in four ways:(1)providing expert review of severe accident PRAs to assess, for example, the risk implica.
South Texas Project Risk Analysis. In 1992, the staff com-tions of accident management strategies in order to mini-pleted a review of the South Texas Project risk analysisand mize the release of radioactive material to the environ-documented the results and findings (NLIREG/
ment during severe reactor accidents: (2) developing, CR4606).The licensee estimated the overall mean core verifying, demonstrating, and maintaining methods for damage frequency to be 2 E-4 per reactor year, which is analy/ing the consequences of in-plant and offsite severe found to be within the range of core damage frequency accident physical processes for use in risk assessment and estimates provided for similar Westinghouse PWR facili.
developing and demonstrating methods for quantifying ties. The licensee has subsequently requested modifica-the uncertainty in risk estimates and the relative contribu.
tions to its plant technical specifications based, in part, on tions of specific issue uncertainty to the overall uncertain.
its risk analysis. The RES staff is now working with NRR ty; (3) reassessing periodically the frequencies, conse-on the acceptability of the requested modifications and quences, and risk of severe accidents in nuclear power expects to complete this work early in FY 1994, plants and performing peer review of methods used and results obtained; and (4) developing risk based manage.
5.1.3.2 Methods Development Projects ment tools capable of determining the incremental risk SAPHIRE Computer Tools. A suite of computer codes reduction associated with proposed plant design and op-called SAPHIRE (System Analysis Programs for Hands-erational modifications and assisting in the setting of prio-on Integrated Reliability Evaluation) has been updated to rities for efforts in licensing and research activities.
Version 5.0. This set of codes is for use in performing PRA and to permit an analyst to perform many of the functions 5.1.3 Research Accomplishments in FY 1993
".ecessary to create, quantify, and evaluate the accident nsks of nuclear power plants. 'Ihe codes are being used Probabilistic risk analysis is used by the NRC staff to extensively to perform the low-power and shutdown risk support the resolution of a wide spectrum of reactor regu.
analyses described above, to perform analysis for resolv-latory issues. In FY 1993, these applications encompassed ing generic safety issues, and to set priorities for inspec-both specific issue-oriented projects and more general tion activities. During 1993, PRA data from three more work, including development and demonstration of risk licensed nuclear power plants were added to the SA-analysis methods and development of risk related train.
PHIRE data base, and most of the data from previous ing and guidance for the NRC staff.
plant k) ads was updated to Version 5.0. This brings the data base total to 13 plants, two of which are advanced 5.1.3.1 Issue-Oriented Projects concept plants that have been added to support the agency's design certification reviews. Courses continued Analysis ofLow-Power and Shutdown Accident Risks. As a to be provided to both the NRC staff and contractors on result of the Chernobyl accident and other precursor the use of these codes. When the documentation for Ver-events around the world, an extensive two-phased project sion 5.0 is completed (scheduled for January 1994), the 5-1 NUREG-1266
- 5. Reactor Accident Analysis o
new codes and user manuals will be sent to the Energy 5.1.3.3 Risk Related'Iraining and Guidance Develop.
Science and 'Ibchnology Software Center at Oak Ridge ment National I aboratog for general distribution.
Guidance for Staff Use of Risk Analysis. In a July 1991 letter, the NRC's Advison Con mittee on Reactor Safe-Consequence Code Benchmark. The NRC is working with guards (ACRS) identified a number of concerns with the the Commission of the European Communities and the staff's uses of risk analysis. In response, t he N,RC's Execu-tive Director for Operations formed a working group of Organization for Economie Cooperation and Develop.
staff management to " consider what improvements in ment to perform an intercomparison exercise on probabil, methods and data analysis are possible and needed, the istic accident consequence codes. The six codes being role of uncertainty analysis in different staff uses of evaluated are MACCS (U.S.), COSYMA (Germany),
CONDOR (UK), OSCAAR (Japan), LENA (Finland),
PSA,. "This workmg group was organized in early 1992 with the following objectives:
and ARANO (Sweden).The comparison exercise uses six radioattive accident source terms and calculates dose
,Ib develop guidance on consistent and appropriate e
consequences for such measures as whole body dose and uses of PRA within the NRC:
fatal cancers. The results of these comparisons will be obtained early in FY 1994 and will provide a data base to
,Ib identtfy skills and experience necessary for each e
judge the performance of new codes to predict reactor category of staff use; and accident consequences.
'Ib identify improvements in PRA techniques and e
associaled data necessary for each category of staff Sunty and baluation ofAging Risk Assessment Methods. A use.
survey and evaluation of aging risk assessment methods and applications is being performed. A draf t NU REG /CR In October 1993 the group completed a draft final report on the work has been received from the contractor and the that included initial guidance to the staff on the use of final NUREG/CR will be published early in FY 1994 PRA in screenin1; and analyving reactor operational Major findings of the review include:
events and on basic terms and methods used in PRA.The report also contains a number of recommendations for The issue of aging in nuclear power plants c:mnot be additional guidance development, improvements to the e
addressed by models that are based solely on the NRC's PRA training program, and improvements in PRA current PRA structure and failure rates. Structures, tools and data bases used by the staff. It is expected that systems, and components (SSCs) neglected in the the final report will be published early in 19%
basic PRA model may become important because of aging; the basic degradation mechanisms, such as Reactor Safety Taining Course. In response to a reque.st fatigue, embrittlement, and etosion-corrosion, must from the Office of Analysis and Evaluation of Operation-al Data (AEOD) RES has developed a new course that is be considered.
intended to treat reactor safety in a broad sense. 'fbpics include a historical overview, design basis accidents, bey-Probabilistic models for degradation mechanisms ond-design-basis accidents, accident progression in the would allow the effective use of information regard-reactor vessel, accident progression in the containment, ing the aging of SSCs. Failure rate-based models radiological releases and consequences, and emergency cannot accommodate this type of information, which response. The intended audience includes new agency typically does not include a significant number of employees and other NRC staff not generally familiar failures. The use of these probabilistic models for with these topics. 'lWo presentations of the course have the degradation mechanisms would decrease the un-been offered at N RC'sTechnical Training Center in Chat-certainties present in models that are exclusively tanooga in 1993. With the developmental work on this failure rate based.
course completed, responsibility for its conduct will be turned over to AEOD in FY 1994.
Probabilistic models for the degradation mecha-e nisms would also allow the development of effective 5.2 Containment Performanee risk management strategies.
5.2.1 Stalement of Problem The development of a methodology that includes Core melt accidents that proceed to vessel failure have o
aging mechanisms can build on existing PRA models, the potential to produce high pressures and temperatures appropriately modified, as current external event that might challenge containment integrity. It is known analyses do.
from previous risk studies, and from the experiences at N UREG-1266 5-2
- 5. Reactor Accident Analysis i
l Chernobyl and Three Mile Island, that containment sur-5.2,3 Research Accomplishments in FY 1993 vivat or even delayed failure has an all-important e ffect on minimizing the release of radioactivity to the environment in order to ensure that existing regulations adequately in the event of a core melt accident. If realistic assess-protect the public from the consequences of severe acci-ments of the consequences of core melt accidents, which dents, the N RC conducts research in several areas, among so strongly depend on whether or when containments them source term release and transport, core melt pro-might fail in the course of the accident, are to be made, gression, fuel-coolant interactions, melt-concrete inter-then an understanding of the phenomena that occur in actions, direct containment heating, and hydrogen com-containment in the latter stages of the accident that could bustionJihe overall goals of this research are to develop lead to containment failure is imperative.
technical bases for assessing containment performance I
over the range of risk-significant core melt events, to develop an improved understanding of the range of phe-nomena expected during severe reactor accidents, and to t
5.2.2 Program Strategy develop improved methods for assessing fission product behavior. With these kinds of data, the NRC is better able NRC's research efforts in this program element focus to confirm the adequacy of its requirements for the design i
directly on the phenomena believed to be most likely to and reliability of the systems that may be used for mitigat-produce high pressures and temperatures that might chal-ing the effects of severe accidents.
lenge the containment integrity: high-pressure ejection from the reactor vessel of finely divided particles of mol-5.23.1 Melt-Concrete Interactions and Debris Cool-ten core debris: generation of noncondensable and flam.
ability mahle gases from the decomposition of concrete by hot in those severe accident scenarios in which the reactor core debns; the direct thermal and chemical attack by molten core debris of structures and engineered safety vessel fails, high-temperhture core debris may fall into the features; and the burning or detonation of hydrogen and reactor cavity where it can the rmally and che mically inter-other gases pnxluced in the course of the accident.
act with structural concrete. The consequences of these melt-concrete interactions can have a significant effect on containment loading, the potential modes of contaimacnt NRC's research program dealing with reactor contain-failure, and the radiological source terms.The major ar-ment safety consists of five areas of research. These five cas of concern associated with melt-concrete interactions research activities address: (1) the interaction of molten during a severe accident are the penetration of the base-core debris with structural concrete, including the abla-mat and failure of the liner, the generation of radioactive tion of concrete structures, heat transfer to structures in acrosols and gases, including combustible gases, and the the containment and to overlying water, the generation of overpressurization of the containment.
noncondensable and flammable gases, and fission prod-ucts containing acrosols; (2) direct containment heating The NRC has conducted an extensive program of analyti-by molten debris particles ejected from the vessel at high cal and experimental research to obtain improved under-pressure and hydrogen pnxtuction resulting from steam standing of melt-concrete interactions. The experimental oxidation of t he metallic component of that debris; (3) the research is focused on conducting experiments simulating combustion of hydrogen in the containment, including the a wide variety of concretes used in nuclear power plants in potential for detonation; (4) the development, validation, the United States and consideration of the diverse acci-maintenance, and application of various computer codes dent scenarios that may lead to melt-concrete interac-that are capable of describing the multiple phenomena tions. The analytical research focused on the develop-that occur in severe accident sequences of interest; and ment of models for st udying phenomenological aspects of (5) assessment of severe accident phenomena, includmg melt-concrete interactions and included a reassessment containment performance, for advanced reactors (see of models used to predict aerosolgeneration and radionu.
Section 2.2).
clide release.
Early experiments on melt-concrete interactions were The overall goals of the research are to develop technical conducted without the presence of an overlying water bases for assessing containment performance over the pool. The experimental data base generated from these range of risk significant core melt events, to develop an experiments is extensive and spans a broad range of melt improved understanding of the range of phenomena ex-release conditions as well as concrete types. No additional pected during severe ret.etor accidents, and to develop melt-concrete interaction experiments are planned. More improved methods for assessing fission pnxluct behavior.
recent experiments on melt-concrete interactions were With nese binds of data, the NRC is better able to con-conducted in the presence of an overlying water pool. In firm 6he adequacy of its requirements for the design and I'Y 1991, the NRC initiated the WETCOR program, also anability of the systems that may be used for mitigating called the debris coolability program, to address two spe-the effects of severe accidents.
cific issues: (1) the comparative coolability of oxidic and 5-3 NUREG-1266
- 5. Reactor Accident Analysis i
metallic debris and (2) the effects of boundary conditions ejected from the reactor cavity into surrounding contain-on coolability,i.e., crust formation and stability. A report ment volumes in the form of fine particles, thermal ener-describing the WETCOR-1 test (NUREG/CR-5907), the gy could be quickly transferred to the containment atmo-only integral test conducted under this program, was pub-sphere. The metallic components of the ejected core hshed in FY 1993, debris could further oxidize in air or in steam and could generate a large quantity of hydrogen and chemical ener.
'Ihc second experimental program on debris cootability, gy that would further pressurize the containment. 'lhis called the Melt Attack and Coolability Experiments process is called direct containment heating (DCil).
(M ACli) program, was developed as an extension of the Advanced Containment Experiments (ACE) program un-
'Ib help develop a data base by which to estimate the risk der the sponsorship of NRC, EPRI, and other largely associated with high-pressure core melt accidents, the governmental agencies in several countries. The M ACE NRC has, in FY 1993, completed four DCII integral ef-program is intended to determine the ability of water to fccts tests fora containment configuration simulating that cool prototypie core debris (urania-zirconia composition).
of the Surry PWR plant, three in the 1/6th. scale Contain.
Three tests were conducted under the M ACE program in ment Technology Testing facility and one in the FYs 1992 and 1993. Data from these tests werc analyzed in 1/10th-scale SURl'SEY facility at the Sandia National FY 1993. Except for the last test that was terminated Laboratories. Analysis of the test results is near comple-prematurely, the results from the M ACE tests generally tion. In addition,in the 1/40th-scale COREXITfacility at support the concept of crust formation at the melt.
the Argonne National bboratory, three integral effects coolant interface with periodic access of water to the melt tests were conducted using reactor materials (urania-and partial melt cooling. A fourth MACE test at a scale based melt) instead of an iron-alumina melt simulant.
larger than the previous tests is planned in FY 1994 to Separate-effects testing is being performed at Purdue study the effect of scale on crust formation, stability, and University in a 1/101h scale model for a containment con-debris coolability.
figuration simulating that of the Zion PWR plant to study the DCll phenomena in detail and will be completed in A topic related to melt concrete interactions, particularly FY 19%
in connection with the HWR Mark I containments, is that of melt-structure interactions, leading to early contain.
As part of the DCll issue resolution plan for PWRs, a ment failure attributable to liner (shell) meltthrough.'the study was completed and documented in " Integrated lle-NRC research over the past few years has adJressed key port on DCil issue llesolution for PWRs"(N U REG /CR-phenomena associated with the liner meltthrough issue.
6109).This report outlines the DCil issue resolution pro-Integration of the research into the assessment of the cess, demonstrates the process for two specific plants, conditional probability of liner failure-both with and Zion and Surry, and describes the approach for resolution without an overlying water pool in the drywell-given a for the remaining PWRs. Supporting documentation and core melt accident that proceeds to vessel failure, was studies related to this report have also been completed, completed in FY 1991 and documented in NUREG/
One of these supporting documents is a study of DCil for CR-5423. The overall conclusion from the assessment the Zion plant,"'Ihe Probability of Containment Failure 1
was that the presence of water on top of the core debris by Direct Containment licating in Zion" (NUREG/
would prevent containment shell failure. The report was CR4075). Hoth reports are now undergoing peer review, peer reviewed, and, as a result of the review process, four and it is expected that they will be completed in FY 1994.
areas were identified that needed further resolution-lin-er failure criteria, melt spreading, melt concrete interac-5,2.3.3 Ilydrogen Combustion tions, and melt release conditions. Additional work was SI "iIIcant informat. ion exists on hydrogen combustion, 8.
performed in these areas in FYs 1992 and 1993. The re-sults from the additional research confirm the conclusions and it is sufficient to assess the possible threat to contam-in the original study. NUREG/CR4025, which included
- C"I and safety-related equipment. Some res, dual issues i
rentam that are related to understanding the likelihood of updates from the additional research efforts, was issued in various modes of combust on at high temperature and in py 1993, the presence of large quantities of steam, i.e., deflagra-tions, diffusion flames, accelerated flames, transition 5.2.3.2 liigh Pressure Melt Ejection-Direct Contain.
from deflagration to detonations (DDI.J.and detonations.
ment lleating In certain reactor accidents, degradation of the reactor The largest current program in this topical area comes out core can take place while the reactor coolant system is of a joint agreement between the NRCand the Ministryof projected to remain pressurized. A molten core, if left International Rade and Industry (MITI) of Japan (man-unmitigated, will slump and collect at the bottom of the aged by the Nuclear Power Engineering Corporation).
reactor vessel. If a breach occurs, the core melt will be Under this agreement, a high temperature hydrogen ejected under pressure, and if the material should be combustion program related to high-speed combustion l
l l
NUR EG-1266 5-4 1
?
- 5. Reactor Accident Analysis f
modes, i.e., detonations and Dlyl; has been developed na-including reactor coolant system and containment and is under way at llrookhaven National Laboratory. A thermal hydraulic response, core heatup, degradation small scale developmental apparatus (SSDA) was con-and relocation, and fission product release and trans-l structed. The SSDA has prmided a preliminary set of port-is treated in MELCOR in a unified framework for experimental data to characterize the effect of high tem-both IlWRs and PWRs.
peratures on the ability of hydrogen-air-steam mixtu r es to undergo detonation. Equally important, the SSDA was MELCOR has been applied to the analyses of various used to support the design of the larger-scale high-plant accident transients, and assessment efforts have temperature combustion facility (1ITCF)by providing so-been completed by a number of United States and inter-lutions to a number of design and operational problems at national user organizations. The focus of the develop-f high temperatures.The construction of the Ifl CFis com-ment cfforts in FY 1993 has been to improve capabilities plcted, and high-temperature experimen ts will begin dur-to h indle the phenomena of natural circulation, external
(
ing FY 1994. As a result of the cooperative agreement heat transfer, and lower-head failure and to model a few with Japan, the NRC has access to ongoing hydrogen specific features of the advanced light-water reactor i
mixing and distribution testing in the'lhdotsu facility and (ALWR). The efforts have also addressed a number of the combustion testing in the 'thkasoga facility. This re-suggestions for improvement of the code made by an search provides a greatly expanded and improved data independent peer review group convened at the NRC's base for the validation of analytical tools.
request. Further, a MELCOR Cooperative Assessment Program effort, started last year, has continued in FY A hydrogen research program is also under way to investi-1993. The goal of this effort is to create an international gate diffusion flame behavior in low-speed hydrogen com-forum for information exchange on the applicability,limi-bustion. Experiments were performed in a small-scale tations, and operational experience of MELCOR while facility to examine the influence of ignition source providing detailed MELCOR assessment calculations strength on lean flammability limits of hydrogen-air mix-and reports to the NRC and to the MELCOR user com-tures at a temperature of 300K and pressure of I bar. A munity.
report was issued describing the facility and the first series of experimen t Se experimental facility will be modified CONTAIN is a detailed code for the integrated analysis of to climinate jet interference with the s! ewalls and to containment phenomena. The code provides the capabili-d establish stable jets. The results from this research prP ty to predict the physical, chemical, and radiological con-gram will be used to help resolve outstanding issues in ditions inside a nuclear reactor containment in the event severe accidents, i.e, hydrogen combustion aspects of of a severe accident. The code also provides the capability DCil,1iigh-temperature combustion phenomena, and to predict fission product releases to the environment in detonation tmtiation by high-temperature steam-the event of containment failure. Among the models in-hydrogen-particle Jets.
cluded in CONTAIN are heat and mass transfer, acrosols n
np et Masr, DammaNe gammWon, 5.2.3.4 Severe Accident Codes core-concrete mteractions, and direct containment heat-Hecause of the difficulty in performing prototypic experi.
ing. The code is capable of analyzing a wide variety of ments for a variety of severe accident scenarios, substan.
LWR plants, including their engineered safety systems tial reliance must be placed on the development, verifica.
and many kinds of accident scenarios, tion, and validation of system-level computer codes for analyzing severe accident phenomena. A number of codes One issue currently under investigation is DCII and pres-(e.g., MEL.COR, CONTAIN, SCDAP/RELAP5) have surization of the reactor containment atmosphere by mol-been developed for various stages in severe accidents, ten core materials ejected following the lower-head fail-both in-vessel and ex-vessel, for both HWRs and PWRs.
ure of the vessel under pressure. A program to Additional codes such as CORCON, VICTORI A, COM-incorporate selected DCil modets into the CONTAIN MIX, llMS, and IFCI are being developed and main-code, including the assessment against available exper-tained to perform specific functions that require detailed imental data, will be completed in FY 1994. Also, plant modeling and will be used to benchmark the system-level cases were run with the updated CONTAIN code to de-codes discussed above.
termine the impact of DCII on the containment. Another development cffort is related to containment analyses for M ELCOR is an integrated computer code that models the ALWR designs. The industry is developing containment progression of severe accidents in LWR power plants.The designs for ALWRs that incorporate passive cooling and l
tode can be used to evaluate the progression of severe decay heat removal features for protection against reactor accidents from initiation through containment long-term containment ovcipressure in accident situa-l failure and to estimate severe accident source terms and tions. The CONTAIN code was modified in selected areas their sensitivities and uncertainties in a variety of applica-in this regard; it is planned to use the code to evaluate the tions. The entire spectrum of severe accident phenome-experimental data generated by the industry's research.
5-5 NUREO-1266 1
-- \\
- 5. Itcactor Accident Analysis i
i SCDAP/ItELAP5 is a computer code that has the capabili.
completed, COMMIX c;m serve as a benchmarking tool ty to perform detailed analyses of in. vessel core melt for portions of the CONTAIN code.
progression phenomena during various severe accident conditions. 'lhe code has been used for severe accident The COltCON code was developed as a best estimate analyses and for the evaluation of the effect of different computational tool to calculate the thermal hydraulics accident management strategies, including natural circu-and chemistry involving the progression of high-lation studies and the analysis of lower-plenu m debris and temperature core debris as it enxies concrete in the reac-lower. head heatup. The systematic assessment of tor cavity. A significant update of the code, designated SCDAP/ItELAP5 that began in 1991 identified several COllCON410D3 was completed in FY 1993. 'this up-areas of modeling improvements. Work on these improve-date involves improved axial and ndial heat transfer mod-ments that was completed in FY 1993 included: (1) im-cis; inclusion of a condensed phase chemistry model for provement of code modeling and climination of numeri-oxide-metal reactions; improved coolant heat transfer cal errors (such as converting SCDAP fixed input format models, including the effects of subcooling and gas injec-to free format, expanding input error checking and diag-tion on film boiling; addition of models for interphase i
nostic printout, identifying and resolving programming mixing and stratification, improvement of models for errors causing the code to fail, and correcting the code bubble behavior (e.g., bubble size, bubble rise velocity, failure due to axial nodalization limit); (2) enhancement and void fraction); and incorporation of the VANESA of code reliability (such as correcting numerical oscilla.
model for acrosol generation and radionuclide release. A tions, codc failures, excessive run time, code failure or topical report has been prepared to describe the pheno-results changed upon restart); and (3) improvement in menological models and correlations incorporated in the late. phase molten pool formation and melt slumping code and to identify acceptable limits of validity for the modeling (such as modifying changes in junction and vol-models and correlations. Extensive validation of the code urne flow area to account for actual debris-bed height, was performed in FY 1993 to determine its capabilities independent of node size, and adding user input to allow and limitations. The COllCON-MOD 3 is now incorpo-radial spreading of liquefied material into adjacent chan.
rated into LONTAIN and MELCOlt.
nels). 'Ihese modeling improvements have significantly reduced uncertainties in the code calculation cf core melt VICTOIIIA is a computer code designed to analyze fission progression.
product behavior within the reactor coolant system (llCS) during a severe accident.The code provides detailed pre-dictions of the fission product release from the fuel and Other SCDAP/RELAP5 research activities accompliehed transport in the RCS of radionuclides and nonradioactive in FY 1993 include the completion of an independent peer m tenals dunng core degradation. During FY 1993, review of the code and initiation of model extensions to ssessment and validation of models used in the VICI'O-the code in order to address ALWR issues. Further model q computa cale against existing data bases and impiovement and code documentation based upon the ag mst new data from vanous expenmental test facilities recommendations of the peer review committee are being (e.g., FALCON VI, ST) were carned out.
considered. 'Ib ensure that SCDAP/RELAP5 meets de-sign objectives and targeted applications, model assess-Ilattelle Columbus Laboratory is performing (for NRC) ment and validation efforts will contmue.
additional experiments and analyses on the deposition and revaporization of certain radionuclides. The purpose COMMIX is a three-dimensional transient single-phase of these experiments is to identify thermodynamic oroper-computer code for thermal-hydraulic analysis of single ties of the radionuclides in the primary circuit during a and multicomponent engineering systeras. The code severe accident and advance the chemsorption models.
solvcs a system of time-dependent and multidimensional lloric acid is known to react with both cesium hydroxide conservation of mass, momentum, energy, and transport and with cesium iodide to form the less volatile species equations. A number of phenomena encountered in pos-cesium borate. The chemistry of boric acid is complex tulated severe accidents in ALWRs are inherently multi-because it decomposes to a variety of different species, dimensional in nature. The COMMIX code is being de-depending on the environmental conditions. The reac-veloped to address issues such as natural circulation, flow tions under investigation will enhance the models rircady stratification,and the effect of noncondensable gas distri-in the VICIORI A code.
bution on hical condensation and evaporation for the AP600 plant. Code upgrades that were completed in FYs IIMS is a best estimate, transient, three-dimensional 1992 and 1993 include the implementation of multicom-code for analyzing the transport, mixing, and burning of ponent capability, development of the liquid film tracking hydrogen. The code can model geometrically complex model, and incorporation of heat and mass transfer mod-structures with multiple compartments and can simulate els. Code validation effort has been initiated using the the effects of condensation, heat transfer to walls and 1/8th-scale test results for the Westinghouse AP600 pas-internal structures, chemical kinetics, and fluid turbu-sive containment c(xiling system. After the validation is lence. During FY 1993, the assessment work of IIMS NUREG-1266 5-6
.m,
- 5. Reactor Accident Analysis against three test problems was completed, and a report experiment at high fuel temperatures, licre the core ge-was issued documenting the governing physical equations ometty is changing and fission product chemistry and its and computational model of IIMS. During FY 1994, the effect on the retention of fission products within the RCS IIMS user's manual will be developed to provide the basic are significant.The FCI work is focused on the develop-information for setting up and running problems with the ment and validation of appropriate phenomenological i
code. Also, the 11MS will be converted from a main frame and analytical models addressing the fundamental aspects computer code to a workstation environment code.
of FCI, namely, melt quenching and FCI energetics.
)
IFCI is an Integrated Fuel-Coolant interaction code that 533 Research Accomplishments in FY 1993 was formulated as an mdependent stand-alone version in 1993 for workstation application. De capabilities of the The phenomenological research areas for improved un-code are described below in the section on Fuel. Coolant derstanding of accidents include the quantification of Interactions, source terrns, core melt progression, primary system fail-ure from severe accidents, and fuel-coolant interactions.
5.3 Severe Accident Phenomenology 5.3.3.1 Source Terms
" Source Terms" refers to the magnitudes of the radioac-53.1 Statement of Problem tive materials released from a nuclear reactor core to the containment atmosphere, taking into account the timing Major unceitainties in estimating the probability of early of the postulated releases and other mformation needed containment failure, and the associated radioactive re-lease,in the event of a severe accident appear to be signifi-to calculate offsite consequences following a hypothetical severe accident. NRC research m this area helps update cantly related to uncertainties in the in-vessel progression
,11D-14844, which has been m use for three decades, m of the accident while the fuel material remains in the connection with plant siting assessments. The update to reactor pressure vessel. Iletter understanding is being gained of the entire sequence of severe accident phenom;
,11D-14844 was published in d raft " Accident Source Terms for Light Water Nuclear Power Plants"(NUREG-1465).
ena, including core melt progression, fission product re-lease, fuel-coolant interactions, hydrogen generation, The NRC has also entered into an agreement with the and response of the RCS to fuel melting and rek> cation.
Comm.issariat UEnergie Atomique of France (CEA) to Containment failure probabilities and related source terms can now be estimated with less conservatism than in p rticipate m the PIIEllUS-FP program. The program, sponsored by the CEA and the Commission of the Euro-previous analysis to ensure adequate margin.
pean Commumties, is aimed at studying in an in-pile facility, under sufficiently prototypical conditions, those 5.3.2 Program Strategy phenomena that govern the transport, retention, and chemistry of fission products during LWR severe accident In order to better understand just what happens during a conditions. Phenomena to be studied are those occurring core melt accident, and thereby reduce the uncertainties in the core, the primary reactor coolant circuit, and the in both accident behavior and the potential release of containment. The agreement is of significant benefit to radioactivity, the NRC is pursuing a program of research the NRC because, at a relatively modest cost, the NRC addressing (1) the heatup and meltdown of the core; (2) can participate in the PilEBUS FP project over the life of hydrogen generation; (3) fission pnxluct release, trans-the project.The NRC will be able to obtain integral exper-port, and deposition within the RCS; (4) energetic imental data to further validate its analytical models for fuel-coolant interactions (FCis) that may occur as molten fission product transport in the reactor coolant systera and debris falls into the water-filled lower head or as water is containment and for iodine chemistry in the containment.
added to molten debris; (5) the mass composition and Information on core melt progression will also be ob-temperature of the core debris at the time of vessel (or tained to supplement data obtained under the NRC RCS) failure; and (6) the mode of vessel failure, The Cooperative Severe Accident Research Program.The ex-overall program is divided into three main activities: (1) perimental data from PIIEBUS-FP is confirmatory in the behavior and chemistry of fission products released nature and will be used to assess the revised source term during core melt, (2)in. vessel core melt progression, and assumptions used in NUREG-1465.
(3) fuel-coolant interactions. The in-vessel core melt prm gression and hydrogen generation work has included The PilEBUS-FP facility has received a license to refuel in. reactor experiments, out-of-reactor experiments, ex-and start up. Final preparation for the first test, Fl'ID, is amination of specimens from TMI-2, and analytical mod-near completion, and the test is scheduled for the first el development.ne research on the amount, the chemi-quarter of FY 1994.The test matrix consists of six tests, cal form, and the behavior of the fission pmducts released with testing at a rate of one per year.The test matrix was from the fuel in the course of a severe accident requires revised recently to include fission product release tests 5-7 NUREG-1266
- 5. Reactor Accident Analysis under shutdown conditions for a degrading core in an air On the issue of blockage of the core by metallic melt, environment and for late-phase core geometry.
TMI-2 and the results of the experiments cited above have indicated that, for " wet core" conditions (with water in the bottom of the core), the relocating molten metallic 5.3.3.2 Core Melt Progression Zircaloy in the core freezes to block the lower core. All but one of the previous experiments for both PWRs and "In-vessel core melt progression" describes the state of an BWRs were performed for these wet core conditions, and LWR reactor core from core uncevery up to reactor vessel this one experiment did not address the blockage or drain-f meltthrough m unrecovered accidents or through tem-age question. The emergency operating procedures for
(
perature stabilization m accidents recovered by core re-U.S. IlWRs, however, call for reactor depressurization, fkxxhng. Melt progression provides the imtial conditions which lowers the water level below the reactor core so for assessing the loads that may threaten the integrity of that core heatup occurs with very low steam flow through the reactor containment. Sigmficant results of melt pro-a " dry core." Analysis of this case indicates that the mol-gression are the melt mass, composition, temperature ten core metal (and later molten ceramic fuel) might drain (superheat), and the rate of release of the melt from the from the core rather than block the core as at TMI-2.
core and later from the reactor vessel if vessel failure does Drainage would produce a major difference in the mass occur. Melt progression retcarch also provides mforma-and other characteristics of the melt released from the tion about the m vessel hydrogen generation, the condi-core and later from the vessel at meltthrough.
tions that govern the m-vescel release of f,ssion products i
and aerosols and their transport and retention in the In FY 1993, a series of new laboratory experiments were primary system, and the core conditions for assessing acci-prepared to address whether metallic melt drainage, core dent management strategies.
blockage, or cc e plate blockage occurs under BWR dry core accident conditions, and the initial two tests in a simplified development system were performed success-Much has been learned about the processes involved in core degradation and in the early phase of melt progres-fully, The experiment test assembhes m the follow-on tests will be a mockup at full radial scale of a cross section sion from integral tests in the PDF, ACRR, NRU, Pl!E-BUS, and NSR R reactors, from the LOFT-FP2 test, from of the lower quarter of a BWR core (and core plate re-gion) where blockages might occur. The experiment will tests in the German CORA ex reactor fuel-damage test facility, and from separate-effects experiments on signif,
have prototypic reactor materials, heat capacities, geome-tries, and temperature distributions. In these experi-cant phenomena. Most of the currently available informa-ments, melts of metallic Zircaloy that also contam con-tion on late-phase melt progression has come from the tr 1-blade materials are poured into a test assembly at post-accident examination of thcThree Mile lsland Unit 2 prototypic rates (dribbles), and the melt relocation and (FMI-2) core and the MP-1 and M P-2 experiments at the bh>ckage behavior are observed.
Sandia National laboratories. In the TMI-2 core, a debris-supporting metallic blockage formed across the Late-phase melt progression experiment MP-2 was per-lower core from the rek) cation and freezing of metalh,c formed in the Annular Core Research Reactor (ACRR) melt (mostly cladding) during coolant boildown. A grow-in FY 1993 to complete the currently planned program of ing pool of mostly ceramic uranium dioxide fuel melt was experiments on late-phase melt progression, which in-formed later by decay heating m the particulate ceramic volves ceramic (fuel) melting and relocation. Along with (fuel) debris bed above the metallic core bk)ckage. Even i
the earlier MP lexperiment MP-2hasprovided unique though the core had been previously refknied by the information on the governing processes in the growth and high. pressure mj,ection system, the growing pool melted meltthrough of a ceramic melt pool in a particulate ce-out the side of the core and partially dramed into the ramic debris bed in bk)cked core accidents. These are the water-filled lower plenum of the reactor vessel.
conditions of the TMI-2 accident. With the results and interpretation of MP-1 and MP-2 in hand, an expert peer Current NRC research on melt progression is focused on review group will be convened in FY 1994 to review the two major issucsJlhe first issue is determining if there are status of the current knowleds viicte-phase melt pro-any accident conditions for HWRs (and possibly also for gression, the significance of the remcaning late-phase PWRs)in which a metallic core bk)ckage similar to that at melt progression uncertaintics, and the need for, efficacy TMI-2 would not be formed so the metallic melt, and of, and nature of any further research in this area.
later the ceramic (fuel) melt, would drain when formed 5.3.3.3 Reactor Vessel Integrity from the core mto the lower-plenum water. 'Ihc second issue concerns the conditions for meltthrough of the In 1988, the NRC-in cooperation with 10 foreign coun-growing pool of ceramic melt above the metallic bk>ckage.
tries under the auspices of the Organization for Economic The meltthrough threshold and kication determine the Cooperation and Development's (OECD) Nuclear Ener-mass of the melt released from the core and later from the gy Agency (NEA)-undertook a follow-on program to the reactor vessel.
TMI-2 core examination conducted by the U.S. Depart-NUR EG-1266 5-8
- 5. Reactor Accident Analysis ment of Energy. Under this program, called the Th11-2 Results of the lower-head fadure analysis are presented in Vessel Investigation Project (VIP), test specimens from NUREG/CR-5642 in terms of key dimensionless parame-the lower head of the TMl-2 reactor vessel were removed ters to provide " failure maps" that indicate the relative in 1990, and examinations of the specimens were con-potential for failure of the lower head in various failure ducted to obtain information on the melt attack on the modes. 'l hus, the methodology in the report should pro-lower head during the accident. The United States and vide the most likely moue ia the initial quantitative fail-the foreign countries participating in the OECD/NEA ure analysis.
pmject performed metallurgical and mechanical exami-nations of the'IMI-2 test specimens. Results of metallur-Creep-rupture data and high-temperature material prop-gical examinations of the vessel steel samples have pro-erty data were required that were not previously available.
vided estimates of temperature histories of the hiaterial data in the literature apply to design conditions, lower-head samples. These specimens indicated that whereas these failure analyses require data in the vicinity some regions of the lower head reached temperatures of failure conditions, for example, creep-rupture times of during the accident that exceeded the transformation I to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. Thus, the required data were obtained temperature of the steel (727'C). In fact, a hot spot was as part of the lower-head failure program for both pres-found in an elliptical region on the lower head (about I m sure vessel steel and penetration materials (inconel, stain-by 0.8 m)in which the inner surface of the reactor vessel less steel, and SA105 steel). The new high-temperature steel reached temperatures as high as 1100*C and re.
data agree well with extrapolation of previous lower-mained at that temperature for approximately 30 minutes.
temperature measurements and should be adequate for Results of examinations of instrument tube nozzles ex.
failure analyses.
tracted from the lower head showed that some nozzles were melted off by interaction with molten core debris, Tb demonstrate that heat generated in the debris inside whereas other nozzles were not affected at all. In general, the reactor lower head can be effectively dissipated by it was found that the nozzles with the greatest damage boiling on the external surface of the vessel, a contract were located in the vicinity of the hot spot. Data from was awarded to Penn State. This program involves heat examinations of the reactor vessel specimens performed transfer measurement on a downward-facing surface (i.e.,
under the VIP, as well as other data from the TMI-2 hemispherical and toroidal surfaces) to obtain the data accident, were used to perform calculations on potential base for critical heat flux and to develop a comprehensive reactor vessel failure modes. Results of these calculations model to predict the heat transfer from this surface.
showed that instrument penetration tube failures, such as 5.3.3.4 Fuel-Coolant Interactions tube ejection or tube heatup and ex-vessel failure, had large margins-to-failure during the TMI-2 accident. Re-Since the quantification of a steam explosion-induced suits also showed that creep rupture of the vessel, either missile as a possible mode of containment failure (alpha via global creep rupture or local creep rupture resulting mode) in the reactor safety study called WASH-1400, from the hot spot, were the more likely failure modes for significant progress has been made in understanding the the TMI-2 reactor vessel. All the technical work for the limitations of damage from such potential missiles. For TMI-2 VIP was completed in FY 1993, with the final example, in NUREG-1150, alpha-mode failure is not a reports for this project issued in October 1993.
dominant contributor to early containment failure. The emphasis prior to NUREG-1150 on fuelcoolant interac-tions (FCis) was directed at the conditions for in vessel molten fuel pouring into a coolant pool and its likelihood A more general study of reactor lower-head failure for of causing containment failure by energetic interactions, both PWRs and HWRs has also been completed in FY The current emphasis of FCI research is to provide the 1993.The mode and timing of lower-head failure resulting appropriate phenomenological and analytical tools to ad-from in vessel melt progression have controlling effects dress those aspects of FCI that are germane to three on the subsequent containment loads during a postulated specific issues: (1) FCI energetics, (2) fuel melt quenching severe accident. A final report, " Light Water Reactor in water pools, and (3) water added to a degraded core.
Lower ilead Failure Analysis"(NUREG/CR-5642, Octo-ber 1993), documents the results of this research. The Complementary to the experimental programs on FCI, an report presents the results of potential failure mode anal-Integrated Fuel-Coolant Interaction (IFCI) code was de-yses for a range of debris conditions, lower-head designs, veloped by the Sandia National Laboratories. Work was and accident scenarios. The failure modes include global completed in FY 1993 to modify the IFCI modules and creep-rupture of lower head, penetration tube melt-produce a stand-alone code for workstation computers.
through, tube ejection, and ablation byjet impingement of An operational report with examples of runs using the molten core material. In addition, an analysis of a limited stand-alone version and a code manual were completed in vessel wall area that may be heated to a high temperature, FY 1993. Validation and peer-review of the code will as occurred in the TMI-2 accident, was also developed.
be the major efforts for the next two years. The code 5-9 NUREG-1266
- 5. Reactor Accident Analysis calculates the four phases of the FCI process: mixing, ing in the PCCS and acrosol modeling in the contain-triggermg, detonating and propagating, and hydrodynam-ment. Work has begun on an in-house code analysis ic expansion.
effort using MEL.COR 1.8 2 to review the new bottom-head model developed at the Oak Ridge Na-The NRC and the Safety'R chnology Institute of the Joint tional Laboratory.
Research Center of the Commission of the European 4.
A CONTAIN input deck has been prepared to model Communities at Ispra, Italy, have entered into a technical exchange arrangement to perform a series of FCI experi.
the SURFSEY facility at the Sandia National Lalw ments at the FARO facility k)cated in Ispra. At the FARO ratories.This deck is being prepared to gain experi-facility, large masses of prototypic reactor core materials ence with creating a CONTAIN input deck before can be melted and can interact with different depths of attempting to create a larger, more complex CON-coolant at different temperatures and pressures. At least TAIN deck modeling the ISP-35 (NUPEC mixing five molten FCI experiments will be conducted to obtain experiments).
data prototypical of reactor conditions in the United 5.
'the IFCI code (fuel-coolant interactions) was re-States.
ceived, installed, and run on the in-house worksta.
5.3.3.5 In-liouse Severe Accident Analysis Capability tion. Also, the IIMS code (gas flow, hydrogen) was transferred via internet from Los Alamos National Growth in the capability of workstation level computers Laboratory to the in-house workstation. The HMS provides the opportunity to run severe accident codes on code is currently being installed.
other than main frame computers. In FY 1993, RES pur-chased workstations to enhance the in-house analysis ca-5.4 Reactor Containment Structural pability at the NRC. These workstations are scheduled to be upgraded by March 1994. The upgrades will increase Integrity processing speed and electronic storage space. Reactor plant descriptions, or decks, for analyses using the MEL-5.4.1 Statement of Problem COR, SCDAP/RELAP, CORCON, CONTAIN, and The major source of risk to the public from the operation VILTORIA codes have been installed on the worksta-of nuclear power plants stems from accidents that lead to tions.'lypical use of this new in house capability has been a containment failure.The regulatory concern is that the to review input decks developed by NRC contractors and f ilure modes and associated load levels for contamment to use these decks to extend previous analyses. In addi-structures cannot be predicted with any real confidence by tion, in-house analyses have been used to check new mod.
the methods used for design. Ihis ts especially so if the els in the codes and to do bounding calculations to deter-contemplated failure mode is localized leakage. Iloth mine the appropriateness of the new modcIs. The ssessments of the risk posed by loads outside the design following describes examples of some of the work done in basts and estimates of the effectiveness of proposed miti-this area in 1993:
gative steps require an ability to predict the way m which a c ntainment will fail.
1.
A CORCON AllWR deck was developed, and sensi-tivity analyses were performed to investigate the effects of melt temperature, melt mass, melt compo-5.4,2 Program Strategy sition, concrete composition, and other code param-Research on containment failure modes is based on the cters that influence chemistry and heat transfer observation that excessive leakage can occur, basically, modeling. Results of this work were provided to from the following sources:
a gg g, Failure of the shell, either the containment shcIl e
2.
An AP600 MELCOR input deck, develoI3ed by the itself (in the case of steel containments) or the liner Brookhaven National Laboratory, was run and eva-(in the case of concrete containments);
luated. Several problems regarding control function activation levels were discovered, and NRR users I cakage at large penetrations as a result of the in-e were notified of these problems.
elastic deformations and/or degradation of seals and 3.
The in-house effort has included an application of Leakage at electrical penetrations due to degrada-CONTAIN to the AP600. Sandia National Laborato-e rics developed the input deck and studied heat trans-tion of materials under the high temperatures asso-fer modeling uncertainties and the uncertainties in ciated with accident scenarios; and passive containment coolant system (PCCS) shell Leakage through valves due to pressure and temper-wetting 'lhe in. house effort extended these studies e
to h>ok at uncertainties in thermal radiation model-ature effects.
l NU REG-12th 5-10 l
l
- 5. l<cactor Accident Analysis flescarch related to shell failure or deformations of pene.
5.5 Severe Acci(lent l'olicy trations rests on analyses of and expcriments on model II11plCI11CithlfiOli tests of actual containment designs. Ihese tests involve pressurization up to failure levels under ambient temper-atures. Since seal and gasket materials are adversely af.
5.5.1 Statement of Problem fected by the temperatures associated with severe acci-dents, separate tests focusing on the development of A severe accident in a nuclear power plant is an event in leakage are performed under pressure and temperature which the core is damaged and there is a potential for conditions, usually at full scale. Examining the possibility release of large amounts of fission products. Significant 01 developing leakage through electrical penetration as-research has been performed on the likelihood, progres-semblies and valves also requires experiments under tem-sion, and consequences of a severe accident as discussed perature and pressure conditions at full scale.
earlier. Much of this work has concentrated on the per-p formance of the containment during a severe accident, i
including potential containment failure mechanisms and the ability of the containment to mitigate the conse-quences of a severe accident, 5.4.3 Research Accomplishments in FY 1993 In the Commission policy statement on severe accidents in nuclear power plants issued on August 8,1985, the The major undertaking in this program for Ihe next few Commission concluded that existing plants pose no unduc years will be a cooperative one with the Ministry of Inter-risk to the public health and safety and that there is no national Trade and Indust ry (M f n)of Japan. Two areas of immediate need for generic rulemaking related to severe cooperation have been identified-one dealing with steel accidents, liowever, based on NI(C and industry experi-containments used in both the United States and Japan ence with plant-specific probabilistic risk assessments, the for ilWit designs, the other relating to prestressed con-Commission recognized the need for a systematie exami-crete containments. The er.rren. generation of Japanese nation of each existing plant to identify any plant-specific PWit containments are of a prestressed concrete design, vulnerabilities to severe accidents. The policy statement indicated the intent of the Commission to take all reason-able steps to reduce the probability of a severe accident
{
and, should a severe accident occur, to mitigate its conse-In FY 1993, agreement was reached on the concept and quences to the extent possible. As part of theimplementa-preliminary design of a two phase test of a model of a steel tion of the Commission's Severe Accident Policy State-IlWh conlainmentJihe model will be fabricated in Japan ment, the staff has requiredindividual plant examinalions and erected at the Sandia National Laboratories in Albu.
(IPEs) of all existing plants to identify any plant specific querque, New MexicoJlhe model will be at a scale of 1:10 vulnerabilities to severe accidents.
to a typical llWR steel containment.The first phase of the test will not have a representation of a shield building and will simulate response to loading scenarios for which Much of the work performed to implement the Severe there is no contact between the containment and the Accident Policy Statement has focused on research into phenomena that would occur during severe accidents and shield building. That phase is planned for 1996. The sec" methods to systematically discover vulnerabilities for se-ond phase willinvolve the construction of a structure t vere accidents. This work has shown that the causes and
)
represent the effects of a shield building and a test to consequences of severe accidents can be greatly in-simulate the response to loading scenarios for w hich con-fluenced by nuclear power plant operators and that many tact is expected. That phase is planned for 1997, vulnerabilities to severe accidents can potentially be clim-inated by proper operator actions. The TMI-2 accident and other abnormal occurrences in nuclear power plants have shown that operators do not stand idle but actively In FY 1993, agreement was also reached on the prelimi-intervene in attempts to control the event. If operators nary design for a model of a prestressed concrete contain-are provided with proper guidance and training to take ment, representative of those used in PWR designs.The beneficial actions when needed and, most importantly, model will be at a scale of 1:4 to a typical large, dry PWR refrain from actions that can have adverse effects, the containment. The model structure will be designed by consequences of a severe accident can potentially be sig-Japanese firms and constructed at the Sandia National nifictmtly reduced. Since many accident management laboratories. The test, currently scheduled for 1998, will strategies do not involve significant plant design changes, consist of a single phase in which the model will be pres-substantial safety benefits can be quickly achieved by en-suriced up to failure.
suring proper operator actions. Thus, the initiation of 5-11 N UREG-1266
- 5. Itcactor Accident Analysis accident management programs at operating plants is a tions, human factor considerations, and probabilistic risk logical result of the IPE process.
assessments. In August 1985, the Commission issued a Severe Accident Policy Statement, which concluded that This program clement provides for the implementation of existing plants posed no undue risk to public health and the Commission's Severe Accident Policy Statement and safety, Ilowever, the Commission recognized that system-the application of the results of severe accident research atic examinations of existing plants could identify directly to the regulatory process. Modification of the plant-specific vulnerabilitics to severe accidents for which Commission's rules or policies regarding siting, cmergen-further safety improvements could be justified.
cy planning, and containment design are examples of ar-cas in which the results of severe accident research may In connection with the implementation of the Commis-l affect future changes.
sion's Severe Accident Policy Statement, the staff has required IPEs of all existing plants to identify any plan t-specific vulnerabilitics to severe accidents. The IPE 5.5.2 Program Strategy process involves two different efforts. The first is an ex-l amination of existing plants for vulnerabilities to severe RES has been given the responsibility for the i.mplemen-accidents resulting from events occurrmg within the plant i
tation of the IPE. This implementation has involved de-b8" equipment failures, pipe breaks). Ihe second effort veloping guidance for performance of the IPE, preparing is to consider severe accident vulnerabilitics from exter-a generic letter to plant operators requesting the IPE. and n
z g, carthquakes, fh>ods, winds), this activity developing review plans and reviewing the results of the e inMual plam exanunauon fom IPE submittals.The requirement to correct any identified is refened to aspEE).
ternal cents @
plant-specific vulnerabilitics not voluntarily corrected will be determined by the backfit rule. Accident management
'lkenty six new submittals for IPEs were received from is not required as part of the IPE process but was high-licensecs in FY 1993, making an overall total of 63 submit-lighted in the IPE generic letter as a future requirement tals received to date. Staff evaluations were issued for that will make use of the results of the IPE process. Sc-
'Ibrkey Point 3 and 4, Oconce 1,2, and 3, licaver Valley 2, verc accident vulnerat ilitics duc to external hazards (c.g.,
and Diablo Canyon 1 and 2 and draft staff evaluations f
carthquakes, fk>ods, fires)are being considered under the were completed for FitzPatrick, Surry 1 and 2, Millstone 1, IPE for external events (IPEEE) program, and Monticello. It is expected that all IPE submittals will be received and reviewed by the end of calendar year 1995.
I 5.5.3 Research Accomplishments in FY 1993 The approach for review of the IPEEE will follow closely In the 14 years since the TMI-2 accident, the NitC has that developed for review of the internal-cvent IPE sub-sponsored an active program in research on severe nu-mittals. The staff initiated the procurement process to clear power plant accidents as part of a multifaceted ap-obtain contractual assistance for the IPEEE reviews. Four proach to the assurance of safety in this context. Other IPEEE submittals have been received, with two currently elements of this approach include improved plant opera-in the review process.
f i
NUllEG-12%
5-12
- 6. SAFETY ISSUE RESOLUTION AND REGULATION IMPROVEMENTS 6.1 Earth Sciences proving estimates of carthquake hazards by identifying potential earthquake sources and determining the propa-g g g, p.gg gation of seismic energy with distance, (2) estimating the i
possible range and likelihood of scismic ground motions Earthquakes are among the most severe of the natural at nuclear plant sites, and (3) assessing the effect of these hazards faced by nuclear power plants. Very large carth, ground motions on soil, structures, equipment, and sys-1 quakes would simultaneously challenge the ability of all tcms of the plants.*lhe m, tegrated results of this research plant safety systems to function and, coupled with the will be used to quantify the nsk to nuclear plants from likely loss of offsite power and dependent safety systems, carthquakes, to assess the seismic safety margins mherent could pose a unique threat to public safety. As wit'h many in current or future plant design, and to help identify and I
potentially severe conditions, there is much uncertainty set pnontics for what improvements are needed m plant associated with the design and evaluation of nuclear designs or what parts of seismic design cntena may be plants for carthquakes. Seismic hazard in the Eastern and relaxed.
Central United States remains an issue that is not likely to be easily resolved.'Ihese regions contain the highest per-A major focus of the NRC research programs in geology, i
centage of nuclear power plants in the United States.
scismology, and geophysics continues to be identifying and defining potential earthquake sources or source Historically, the largest earthquake in the United States mnes in the Eastern United States and using that infor-has occurred at New Madrid, Missouri. The geology of the mati n in assessing seismic hazards with respect to nu-central and eastern regions makes it difficult to estimate clear power plants. Many unknowns exist regarding these carthquake magnitudes or scismic parameters for specific issues, meluding a strong basis for seismic zonation, locations or to ensure a proper design basis for individual s urce mechanisms, charactenstics of ground motions, power plants.
and site-specific response. Ihc NRC is addressmg these uncertaintres through research that encompasses sus-
'lhe publication of seismic hazard curves in 1989 by both t ined scismic monitoring, geologic and tectonic studies, the NRC (NUREG/CR-5250) and the Electric Power ne tectom,e m, ycstigations, explonng the carth,s crust at Research Institute (EPRI)(NP-6395) marks the end of hypocentral depths, and conducting ground motion stu-dies.
major efforts to characterize the seismic hazard at U.S.
nuclear reactor sites. Although the best information and procedures available were used, they revealed that large
,Ihe backbone of the NRC program in the Eastern United uncertainties still remain in scismic hazard estimates.
States has been the seismographic networks deployed Also, recent full. scope probabilistic nsk assessmen ts, per.
throughout the Eastern and Central United States.'lhe formed as part of the NUREG-1150 effort, continue t NRC is currently funding seismographic networks in the show that setsmic hazard uncertaint,es contribute signtit.
following regions: Northeastern United States; Virginia; i
cantly to the overall uncertainty m nuclear reactor nsk Charleston, South Carolina; the southern Appalachian estimates. l'hese large uncertainties make it difficult to region; the New Madrid (Missouri) region; Ohio and In-place the contributton of seismic nsk into its proper per-diana; eastern Kansas; and Oklahoma. An agreement was spect,ve, e.g., m the development of individual plant ex-reached in 1986 between the United States Geological i
ammation guidelmes.
Survey (USGS) and the NRC to jointly support the estab-I shment of the castern portion of a national seismograph.
ic network.The eastern portion of the national network is Recent successes m. the geolog.ical, geophysical, and sets-mological studies sponsored by RES show that it is possi-now fully in place.
ble to answer the basic scientific questions that underlie these seismic hazard uncertainties. It is the goal of the 6.1.3 Research Accomplishments in FY 1993 NRC carth science research program to significantly re.
duce the uncertainty m seismic hazard estimation m the The objective of NRC research in earth sciences, as re-next decade through emphastzmg this type of research, lated to reactor regulation, is to define potential earth-quake ground motions at nuclear power plant sites and in the regions surrounding them. This informetion provides 6.1.2 Program Stralegy a basis for evaluating the effects of earthquakes on the plants and their safety systems.
The strategy to resolve the scismic problem involves re-search to develop the methods and data that will support Seismic hazards contribute a sizable proportion to its the necessary seismic criteria development and provide overall plant hazards and, because ofinherent difficulties the evaluation tools. The research is focused on (1)im-in defining them, they form an even more significant 6-1 NUREG-1266
I
- 6. Safety Issue Resolution I
m portion of the overall uncertainty in estimating plant haz-areas of northeastern Massachusetts and southeastern ards. In order to reduce these uncertainties, research into New llampshire that were affected by the 1727 Newbury-the causes and distribution of seismicity is continuing.
port and the 1755 Cape Ann earthquakes. Both areas are Research is also progressing on improved methods of sites of current seismic activity, applying earth science information to estimates of ground motion levels for use in plant design.
In the Wonalancet/Ferneroft area, the investigations per-formed consisted of trenching, a ground penetrating radar survey, geotechnical engineering tests, landslide /
6.1.3.1 Seismographic Networks rockfall reconnaissance, and geophysics. No evidence was For about a decade and a half, the NRC has supported found that indicated the occurrence of prehistoric carth.
regional seismographic networks, primarily in the Eastern quakes larger than the 1940 event. Samples were taken for and Central United States where most of the nuclear radiocarbon age-dating to determine the length of time plants are h>cated and where scismicity is less well defined during which a large carthquake has not occurred in this than in the Western United States.'Ihese networks have area.
provided essential carthquake data to better describe the scismicity in this region and to compare the seismicity with In the Newburyport and Cape Ann region, many thou-geologic and tectonic information in order to gain insight sands of linear feet of exposure were observed along into structures in the carth's crust that may create a po.
marshes, estuaries, and rivers. Although liquefaction.
tential for carthquakes. 'Ihe NRC discontinued most of susceptible soils were found, no seismically induced pa-these networks in September 1992.Three networks, those leoliquefaction features were identified along these expo-operated by Weston Observatory, Massachusetts Institute sures. Samples for carbon-14 age-dating were obtained to of 'Ibchnology, and Virginia Polytechnic Institute and constrain the time period during which a large earthquake State University, were continued through this reporting has not occurred.
period until September 1993. The function formerly During FY 1993, an investigation was begun in the epicen-served by the regional networks has been taken over by tral area of the 1944 Cornwall-Messena carthquake (mag-the new National Seismographic Network (NSN), estab.
lished through a cooperative agreement with the USGS.
nitude 5.5) to determine whether there was evidence for prehistoric moderate-to-large carthquakes. The 1944 event induced numerous liquefaction features, and the The NSN was officially dedicated in April 199L At pres-strategy is to identify similar features that pre-date the ent, the network consists of 16 network stations and 13 1944 features and suggest earlier occurrences of similar cooperating stations operated by the IRIS consortium and carthquakes in the late lick >cene.
various universities. A number of regional seismographic network stations have also been integrated into the net-Another palcoscismic investigat. ion that is under way work, forming a national seismic system and providing long the Atlantic coast of North America is a study of more detailed coverage for special regions, such as the tsunami deposits left by the 1927 Grand Banks carth-New Madrid, Missouri arca.The NSN operates with high-quake (magmtude 7), the development of critena to d,s-i quality three-component stations and satellite telemetry,
}inguish between these depositsand stonn-gencrated sed; thus providing data on significant earthquakes within min-ment, and a preliminary scarch for earlier tsunami deposits that were the result of prehistoric Grand utes' 13 nks sized carthquakes. Like the previously descril ed Near the end of this reporting period, a broad agency studies, this one is an attempt to extend the relatisle announcement was made public with the purpose of es-short historic seismic record back into time, tablishing research contracts for analyzing NSN data and other available seismological, geological, and geophysical 6.1.3.3 Faulting in Giles County Seismic Zone in Vir-data.This research will continue the type of mvestigations ginia previously carried out by the universities operating re-In June 1992, two faults were discovered at a borrow pit gional networks. It is anticipated that the high-quality, site in Pembroke, Virginia, near the epicenter of the 1897 broadband, and three-component data of the NSN will Giles County Modified Mercalli Intensity Vill (magni-lead to new insights into the causes and distribution of tude 5.8) carthquake. The faults displace high level ter-seismicity and on the ground motion propagation charac-race soils of the New River, which consist of bedded silts teristics of the earth's crust, particularly in the Eastern and gravels estimated to be of Quaternary age (less than 2 and Central United States, million years old). One fault strikes north 64* cast, dips 60* to the northwest, and displaces the terrace strata 6.1.3.2 Northeastern Neotectonics more than 3 meters.The second fault strikes north 70*
During FY 1993, investigations were conducted at Wona-east, dips 80* to the northwest, and offsets the soils about lancet /Ferneroft in central-cast New flampshire, near the 1 meter. Minor tension cracks and slip surfaces are also epicenter of the 1940 Ossipee carthquake; and in the present in the outcrop. Further examination revealed that NUREG-1266 6-2
.1
f
- 6. Safety Issue Resolution these faults formed the eastern margin of a small graben the Reelfoot Rift; and the llootheel fault zone, which is (a narrow depression bounded by faults on either side) located in the middle of the rift zone. 'Irenching will be within the castern limb of a small northerly trending anti.
carried out at specific sites pending the results of the cline. 'lhree possible origins for the faults are being con-geophysical studies.
sidered: landslide, karst, and tectonic.
Ilased on limited evidence, it is hypothesized that a recur-i Additional investigations during FY 1993 revealed that rence interval for the 1811-1812 New Madrid earthquakes the faults bounding the graben (the two faults described (magnitude 8) ranges from 550 to 1000 years. Ilowever, above on the cast boundary and another fault forming the this idea was brought into question following the more west boundary) experienced approximately 11 meters of recent work of Wesnousky, who examined the banks of normal displacement. Other faults were mapped in the numerous Corps of Engineers drainage ditches that ex-excavation, which eventually exposed up to 50 feet of posed llolocene soils, which were strongly affected by Ihe vertical face 'the other faults consisted of small normal 1811-1812 carthquakes but showed no indication of defor-faults with 30 centimeters or less of offset and reverse mation by prehistoric events. The focus of another re-faults with apparent displacements up to 1 meter. Prelimi, search project in this region is to determine whether geo-nary geophysical investigations squth of the exposure logic evidence supports a recurrence of earthquakes like have been inconclusive as to the origin of the faults, but the 1811-1812 magnitude 8 events and to attempt to de-some methods show promise of helping to resolve this termine the ages of those events and the regional extent, issue. Pending the availability of funding, core borings, if they exist, and to develop criteria for identifying them, trenching, age-dating of soils, and geophysical profiling Ilreliminary results indicate that there is paleoliquefac-are planned for FY 1994.
tion evidence for at least one such prehistoric event.
6.1.3.6 Pacific Northwest 6.1.3.4 Palcoscismicity of Southern Illinois and in-diana The Pacific Northwest, from sou thwestern liritish Colum-Investigation began m. FY 1991 to identify and analyze bia to northern California, is underlain by the Cascadia palcoseismic evidence along the Wabash River valley and subduction zone, into which three minor oceanic plates-valleys of its major tributaries 'Ib date, hundreds of pla-the Explorer, Juan de Fuca, nad Gorda plates-are being 1
nar, nearly vertical, sand and gravel. filled dikes-caused subducted beneath the North American plate. Although by carthquake-mduced h,quefaction-have been discov' geological and geophysical evidence indicates active sub-cred m these valleys m Indiana and Illmois. 'lhe dikes duction, there have been no historic large-thrust carth-range in width from a few centimeters to as much as 2.5 quakes along the plate interface-the type of earthquake that characterizes other active subduction zones around meters; the largest of them being found around Vm-the rim of the Pacific Ocean, cennes, Indiana, and decreasmg m size and abundance to the north and south of this area. Studies indicate that most of these features were caused by a large earthquake (esti-The USGS has completed a major 5-year study of the mated magnitude of about 7.5) that occurred in the Vin' geology and tectonics of the Pacific Northwest and contin-cennes area between 2,500 and 7,500 years ago.
ucs to sponsor more limited research in this area. The NRC is partially funding several projects under this pro-gram in western Washington and Oregon. These efforts Investigations during FY 1993 were carried out in south-are continuations of investigations that revealed geologi-crn Indiana and Illinois. The results so far confirm the cal evidence suggesting the occurrence of several large occurrence of a very large carthquake (magnitude 7.5) prehistoric earthquakes during the past several thousand about 6,000-7,000 years ago, centered near Vmcennes, years. This evidence consists of several cycles of normal and indicate that t em was a liquefaction-producing, stratigmphic deposition of shallow marine sediments i
moderate earthqtu 9out 4,(C' y cars ago and a strong overlain by marsh deposits, each of which has been earthquake centered in southcentral Indiana 4,000-6,000 abruptly terminated by a catastrophic subsidence event years ago.
and a new cycle has begun. These events are interpreted to be related to the occurrence of large subduction zone i
6.1.3.5 New Madrid Seismic Zone earthquakes. Along the coast, geologic and radiocarbon data indicate that the most recent of these events oc-Several sites are being investigated in the New Madrid curred about 300 years ago, affecting lowland soils at the seismic zone to define faults associated with the scismicity Copatis River and at Willapa llay about 65 kilometers there. The sites are at the intersection of Crowely Ridge apart. A 300-year-old event is also represented in north-and the west side of the Rectioot Rift, Marston, Missouri, ern California about 610 kilometers to the south. 'l\\vo of where waterfalls formed in the Mississippi River during the research projects have been concentrating on deter-the 1811-1812 New Madrid carthquakes; at the Critten-mining whether these widespread deformations were den County fault zone, which is the kical cast margin of caused by a single magnitude 9 carthquake or by several 6-3 NUREG-1266 1
N
- 6. Safety issue Resolution smaller events of magnitude 8 or less.The data available These data-along with other geological evidence gath-so far supports either hypothesis.
cred by other researchers in the Puget Sound region (such as submarine slides in Lake Washington, uplift at Ilestora-Investigations along the coast of north central Oregon tion Point on Hainbridge Island, geophysical and strati-confirmed the regional subduction zone subsidence graphic evidence for a large east west striking fault in events, but also identified geological evidence for kical south Seattle-Seattle fault)-suggest the occurrence of a prehistoric carthquakes and subsidenec-like evidence large (magnitude 7) carthquake on the Seattle fault about j
that may have been related to nonseismic phenomena 1,100 years ago.
such as storm surges or flooding due to the damming of Field studies in FY 1993 found additional evidence of a estuaries by sand barriers.
tsunami generated by the 1,100-year-ago event in a cove n Whidby Island in Puget Sound in the form of a buried In conjunction with these studies, a study is under way to sheet of sand that underlies the cove and laps up on its j
identify and define seismically induced paleoliquefaction features in the region to determine whether strong shak-D @s e gave is considered to be favorably onented to receive a seismically generated sea wave from an earth-ing occurred during these subsidence events. lleconnais-qu ke on the Seattle fault.
sance investigations in the Chehalis River valley and oth-er drainages in southwestern Washington did not identify 6.1.3.7 Fault Segmentation Studies such features, even though there are long stretches of exposures of liquefaction-susceptible soils along the river It is well known that faults do riot usually rupture over banks.
their entire length during a single earthquake. Numerous structural and palcoscismic studies and investigations of historical earthquakes indicate that there are physical
'Ihe first positive evidence for seismic shaking that can be controls within a fault zone that define the extent of attributed to a subduction zonc earthquake in the Pacific Northwest was found in the Columbia River estuary.The rupture and divide a fault into segments and that these evidence consisted of seismically induced paleoliquefac-segments can persist through many carthquake cycles.
tion features (sand dikes and sills) on islands within the This project is being carried out to establish a basis for estuary 'the features range from up to 0.3 meter in size, recognizing and identifying geometrical and structural features that constram or control rupture propagation which are very numerous in the vicinity of Astoria, Ore-within a fault zone.
gon, to fewer in number and smaller in size upstream ranging from 7.5 to 10 centimeters about 30 to 40 kilome-Eivaluation of the segmentation for selected faults was ters away, and 2.5 to 5 centimeters wide about 60 kilome-begun in FY 1991 using palcoscismic recurrence data and ters mland. The dikes and sills are estimated to be about information on slip-per event and slip rate. Studies in FY 300 years olgl, based on the estimated age of soils cut by the 1992 continued on these faults, including the Rodgers dikes (specifically, a 1,482. year-old layer of pumice),
Creek Ilayward fault zone, the segment of the San An-younger undisturbed soils, and the age of the oldest living dreas fault that ruptured during the 1989 Loma Prieta trees (240 years) unaffected by the event. Ihe evidence carthquake, the Wasatch fault zone, and the Calaveras, for shakmg is correlated with the 300 year-old subsidence Superstition Ilills, Imperial, White Wolf, Lost River, Red event m, southwestern Washington.
Canyon Hebgen,Dixic Wiley-Stillwater, Pleasant Valley, North Anatolia (lbrkey). Pitagcachi(Mexico), Oued Fod-Field studies in FY 1993 showed that paleoliquefaction da (Algeria), Marriot Creek,'Ibnnant Creek (Australia),
features extend at least an additional 30 kilometers up' and Landers faults, stream in the Columbia River for a total distance of 90 kilometers from the coast Preliminary geotechnical in-Work during FY 1992 on the Rodgers Creek fault pro-
+
vestigations of the liquefaction susceptibility of soils on vided the first estimates of the timing of individual pa-Wallace Island in the Columbia River estuary suggest that leocarthquakes with events at about 1000 A.D., between the shaking that accompanied the 300-year-ago event was 1200 and 1400 A.D., and between 1650 and 1808. Addi-probably less than that which would be expected from a tional evidence supporting a 6-kilometer-wide step be-great subduction zone earthquake.
tween this fault and the llayward fault was found. Studies at Grinly Flat on the San Andreas fault revealed Geological evidence from excavations at West Point, evidence for the last two large surface faulting events, one Washington, 10 kilometers northwest of downtown after 1800 A.D. (probably 1906), and the other before Seattle, is interpreted to indicate that tsunami-like surges 1636-1660 A.D. Along with evidence gathered by other of s mdy water from Puget Sound covered a tidal marsh researchers farther north along the fault, these findings that subsided at least 1/2 meter about 1,100 years ago.
indicate a recurrence interval along this part of the San Estuarian mud about 1/2 meter in thickness overlies the Andreas fault of about 250 years. Initial data on fault sand and marsh deposits. Radiocarbon age dates of plants geometry, lithology, and rupture directivity collected buried beneath the mud range from 900 to 1,300 years.
for the Coyote I ake, Morgan Ilill, and Alum Rock NU REG-1266 6-4
- 6. Safety Issue Resolution earthquakes on the Calaveras fault indicate a south-to-In FY 1992, more than 250 earthquakes were recorded, north progression of events. Ilowever, a north to-south the largest of which was the April 23 Joshua 'Iree carth-rupture propagation during each event was indicated.
quake at a distance of 45 kilometers from the array and a depth of 13 kilometers. Maximum acceleration rec cded 2
fr m this event was 89 cm/s at grpund surface. Record-Studies were begun late in FY 1992 on the complex 85-kilometer-long surface rupture of the 1992 Landers earth-mgs were also obtained from the foreshock and the after-quake (magnitude 7.5) to determine its implications for shock. Unfortunately, the data acquisition system was not segmentation modeling. This rupture was characterized working on June 25 durmg the Landers earthquake. Am by strike-slip faulting containing at least three major geo-plification charactenstics for ground motion of FY 199.
metric segments with echelon steps up to 2.5 kilometers carthquakes are bemg analyzed.
across.
Because of the relative lack of near-field recordings of large intraplate carthquakes, such as those in the Eastern In FY 1993, six trenches were dug across the 1992 rupture and Central United States, the prediction of strong along the llomestead Valley fault where it truncated an ground motions radiated by these types of earthquakes is alluvial fan and farther to the south where the rupture severely hindered. Tb compensate for this lack of cuts across a playa. Four events have been identified: the near-field recordings, an analytic method was developed 1992 event, an event about 4,000 years ago, one 8,000 years by Ihe USGS to correct teleseismic recordings of the ago but with a very large error band, and an event about Global Digital Seismic Network for focal mechanisms, 14,000 years ago with an even larger error band. The next interference of the depth phases, and the teleseismic at-stepis to excavate trenches across a segment of the Emer-tenuation in order to estimate the acceleration source son fault that did not rupture in 1992 and continue to try to spectrum of the carthquake in the frequency band from 50 correlate events from fault segment to fault segment to seconds to 2 lie. Many large intraplate carthquakes have test the fault segmentation model and the characteristic been analyzed to estimate the acceleration spectrallevel carthquake model.
expected for near field strong ground motion in north-eastern North America. In FY 1992 the extensive 6.1.3.8 Strong Ground Motion Studies near-field and regional accelerograph recordings from the 1989 Loma Prieta earthquake were analyzed with a In 1989, in cooperation with the French Commissariat a view to applying the results to predicting strong ground l'Energie Atomique, a seismic experiment was undertak-motions in eastern North Amenca.
en at Garner Valley, California, to measure in situ ampli-fication and attenuation of seismic waves that propagat During FY 1993, by studying the S-wave fIains from 97 through a soil column from bedrock to ground surface.
earth 9uakes recorded by the Eastern Can da Network, The original con tract was for the design, construction, and meluding the Saguenay, Mt. Laurier, Miramichi, Good-depk>yment of five downhole accelerometers and a field now, Gaza, and Painesville earthquakes, a model for attenuat,on of ground motions was developed, and infor-operable data acquisition system. In 1990, EPRI funded i
the installation of an additional downhole accelerometer m tion was obtamed about propagation and source char-and four surface accelerometers, along with additional acteristics m the Eastern United States.
data acquisition capability for the extra accelerometers.
As presently deployed, the system is comprised of five One of the objectives in the USGS strong ground motions surface accelerometers in a Im, car array spannmg 310 me-program is to use the stochastic model to predict ground ters and five accelerometers at depths from 6 meters to motions irom earthquakes in castern North America. In 220 meters, l'he network is located 7 kilometers from the FY 1992, an extension of the Boore and Atkinson (1987)
San Jacinto fault, at the northern end of the Anza seismic ground motion predictions to deep soil sites was com-gap on this fault, where a magnitude 6.5 or greater carth-pleted, representmg an mitial step m generahzmg the quake can be expected, and 35 kilometers from the Indio P.rediction methodology to account for local variations m segment of the San Andreas fault.
site geology. During FY 1992, much of the mitial develop-ment of a strong-motion data base, including selection of those carthquake records that meet established quality Since its operation began, the downhole seismic array has control criteria, was completed, recorded numerous earthquakes ranging in magnitudes from 6.1 to approximately 1.0. Analyses of the data The updated work in FY 1993 resulted in a division of sites through FY 1991 indicated that the spectral amplitudes into four building code-like classes based on shear wave recorded at ground surface are amplified on average by a velocities: soil A - > 750 m/s, soil B - 750 to 360 m/s, soil C factor of 10 over the spectral amplitudes at 220 meters
- 360 to 180 m/s, and soil D - < 180 m/s.
depth. Resonance peaks have spectral ratios (surface spectral amplitudes divided by those at 220 meters) of Another ground motion study that was made during FY about 40 for frequencies near 1.7,3.0, and 12.0 Hz.
1993 was in regard to rupture histories of eastern North 6-5 NUREG-1266 w
w 7
wT-v-
- 6. Safety issue Resolution American earthquakes. Sophisticated methods were used 6.1.3.10 Probabilistic Scismic IIaiard Assessments in the Western United States (inverting teleseismic and strongmotion recordingsforspace timeslipdistributions)
Probabilistic seismic hazard assessments (PSHAs) were to analyze large Eastern United States earthquakes such begun about a decade ago, and they have become an as Miramichi, two of the Nahanni events, Ungava, and increasingly important aspect of site evaluations for nu.
Saguenay. The Saguenay event exhibited a concentrated clear power plants and other facilities. 'the revision to source rupture pattern with an initial high-st ress drop that Appendix A, " Seismic and Geologic Siting Criteria for spread over a broader area.The slip concentrations of the Nuclear Power Plants," to 10 CFR Part 100, " Reactor Site two Nahanni carthquakes were spatially complementary.
Criteri4" still in progress, will put substantial emphasis
'the Ungava rupture took place within the upper 3 kilome-on PSHAs as part of the investigation required for pro-ters of the crust.
posed nuclear pohr plant sites. PSHAs are of particular interest in the Eastern and Central United States where At the Savannah River site, a seismic array has been uncertainties created by a lack of detailed knowledge of installed in a borehole. Four events in South Carolina the scismicity make it difficult, by a daterministic evalua-were recorded, the largest of which was a magnitude 4 at tion, to arrive at a dependable estimate of seismic hazards.
Summerville.The data are still being analyzed, but initial results indicate that stress drops increase strongly with
'IWo large-scale PSH A studies are available for the East-increasing moment.
ern and Central United States. One was performed by the Lawrence Livermore National Laboratog (LLNL) and Digital aftershock data from the 1992 Petrolia, California, sponsored by the NRC (NUREG/CR-5250); the other carthquake sequence were analyed in an attempt to de-was performed by EPRI and sponsored by utilities in the termine the reasons for the high accelerations recorded at Seismicity Owners Group. The two studies used similar several of the stations.'lhe results indicated that the high methodologies and produced hazard curven with similar ground motions at the Cape Mendicino Station and the characteristics; they also produced consistent relative Petrolia General Store sites were most likely caused by hazard rankings for pant sites in this region. A serious site responses. The results of the analytic technique problem arises, however, from the fact that, at certain applied to other sites with anomalous readings indicated sites, absolute hazard levels may differ significantly, that the causes were attributable to either poor instru-ment calibration or to wave propagation characteristics.
Results from both studies are used by NRC staff for regu.
I t ry decisions, but, for future nuclear plant design and
- 6. L3.9 Crustal Strain Measurements licensing, more consistent hazard values will be needed.
At the end of FY 1992, an effort was begun to analyze During FY 1993, the crustal strain network for the East.
differences between the LLNL and EPRI seismic hazard ern and Central United States was measured for the third time since 1987. After this strain network was established,
. methodologies and to arrive at a more unified methodolo-it became the backbone of a new geodetic network for the g that will produce more reliable absolute hazard levels.
United States based on Global Positioning System (GPS)
From previous analyses, it was known that methods of measurements. In addition, high-precision G PS networks clicitmg expert opimons and certain other factors-such have been established for many states; within the next few as scismic parameters and ground motion models-cause some of the observed differences. The computer pro-
)
years all of the United States will be covered with detailed high-precision GPS networks for surveying purposes. He.
grams used for th,e LLNL and EPRI methods, although J
cimse of this, many stations are now available for strain different, are designed to solve the same basic equation determinations, in addition to the original 45 stations of and do not seem to be a cause of discrepancy.
the crustal strain network. These additional stations will also be periodically resurveyed and, in many h> cations, A new study will be conducted cooperatively by the NRC permanent GPS stations have been established that will and the Department of Energy (DOE), both because its provide a continuous record of measurements.
cost will be t elatively high and because the DOE also has an interest in PSIIA methods for assessing the numercus J
Because the intraplate strain rates in this region are ex-critical facilities it operates. EPRI will also make a signifi.
pected to be low, many years may be needed to arrive at cant contribution to the research through the DOE.The meaningful strain determinations. Ilowever, with the NRC is sponsoring a peer review by a panel formed by the large number of high-precision GPS stations now avail-Committee on Seismology of the National Academy of able, it should eventually be possible to get a veg detailed Sciences / National Research Council. The peer review picture of strain distribution. Detailed information on panel will provide an independent, scientific review of the deformations in the crust and their temporal rates will project and thus ensure the impartiality and objectivity of then provide a basis for refinements in seismic hazard the study. It is expected that the study will be completed in i
determinations.
1994.
NUREG-1266 6-6
- 6. Safety Issue Resolution 6.2 Plant Response to Seismic and the second time (58 FR 16377)in response to a public Oflier Exterritil Everits request. The comment period expired on June 1,1993.
Responses were received from approximately 47 domestic 6.2.1 Statement of Problem and foreign commenters.'lhe domestic organizations pro-viding comments included State geological surveys, the in the 1970's and before, our interest in nuclear plant USGS, the Association of Engineering Geologists, and seismic design was mainly limited to response at design industry representatives. Nine foreign countries either levels (e.g., OllE and SSE) and our knowledge of this was individually or as a group provided comments. The staffis primarily based on analytical techniques and assumptions, reviewing all the comments and will revise the regulations in the 1980's, a considerable effort was made to better and guidance documents as appropriate during FY 1994, predict the potential response of nuclear plants to carth-quakes greater than those considered in design. Our un.
Revisions of the geologic, seismic, and carthquake engi-derstanding has been increased greatly by the testing to neering criteria are being performed in conjunction with failure of equipment and structures, by the gathering and the revision of the reactor site criteria,10 CFR Part 100.
synthesis of earthquake experience data from non-nu-6.2.3.2 Seismic Testing of Relays clear facilitics, and by the large number of seismic PRAs that have been made.
Seismic testing of relays to determine the influence of relay chatter on circuit breaker tripping, among other This research has generally found that the seismic capac-things, was completed in FY 1993. 'lhe research initially ity of important nuclear plant structures and equipment was intended to support the resolution of US! A-46, (when properly anchored)is high. Ilut there remain spe.
"Scismic Qualification of Equipment in Operating cific capacity concerns that need to be resolved, such as Plants," but will also serve the needs of IPEEEs and how to address the potentially harmful effects of relay seismic PRAs for advanced light water reactors chatter. The importance of plant-specific walkdown re-(ALWRs). Results obtained in FY 1993 indicate that relay views to find nongeneric vulnerabilities has been noted in chatter may or may not be acceptable m spectfic circuits recent seismic margin studies, depending on the circuit parameters. Thus, the two milli-second chatter criterion in IEEE codes may not be appro-priate in all cases. Further evaluations are needed to 6.2.2 Program Strategy decide on the appropriate course of action.
In recent years, the NRC has supported seismic testing 6.2.3.3 Hurricane Andrew and the collection of earthquake experience data in order On August 24, 1992, liurricane Andrew, a Category 4 to improve and gain confidence in the use of seismic PRAs hurricane, struck the "Ibrkey Point nuclear power plant and scismic margin studies.These data are also being used with sustained winds of 145 mph (233 km/h). During FY to support proposed improvements to seismic design crt-1993, a combined NRC and Institute of Nuclear Power teria. The earthquake resistance of structures, equip-Operations (INPO) team investigated the impact of the ment, and piping has been found, in general, to be higher hurricane on the'Ibrkey Point Units 3 and 4 nuclear power than previously thought. Major efforts in this area were plants.The emphasis of the investigation was on identify.
completed in 1990, and the results are being successfully ing those areas, events, or conditions that were problem-used in licensing actions. Relay chatter is the one remain-atic for the facility and Florida Power and Light (FPL) ing seismic capacity issue that will require additional test-staff as well as those special preparations or actions that ing to resolve.
had a positive effect on the course and consequences of events relevant to plant safety. A RES staff memberinter-Upcoming individual plant examinations and USI A-46 viewed FPL staff about the performance of structures seismic reviews will use the recent results of NRC seismic associated with the nuclear units, the chimneys and research.
Ilunker C oil tank associated with the fossil units, and an earlier systematic evaluation they had performed-the 6,2.3 Research Accomplishments m, Fi,1993 individua! plant examination to identify severe accident vulnerabilities because it included the results of the wind 6.2.3.1 Revision of Appendix A to 10 CFR Part 100 and fire external-event analyses.The team report,"Effect of Ilurricane Andrew on the 'Ibrkey Point Nuclear Gen.
On October 20,1992, the NRC published for public com.
erating Station from August 20-30, 1992" ment (57 FR 47802) the proposed revision of Appendix A (NUREG-1474), was published in March 1993.
to 10 CFR litrt 100. The public comment period was 6.2.3.4 Earthquake Response extended twice-the first time (58 FR 271) so that the expiration date would be consistent with the expiration Following 1992 earthquakes in the Cape Mendocino and date of the supporting regulatory guides (57 FR 55601);
I;inders/Ilig llear areas of California, investigations were 6-7 NUREG-1266
- 6. Safety Issue Resolution j
1 conducted to document the impact that the seismic events is a clearer understanding of the likely failure modes of had on several selected non-nuclear industrial facilities, piping systems unde r carthquake loadings. Design criteria "Ihe facilhies selected included a fossil. fueled power need to address actual seismic failure modes in piping and plant, cogeneration power plant, lumber mill, and a cc-need to be revised to climinate excessive conservatisms ment plant because they had steel framing, piping, and that do not add to safety and may hinder plant operation in some equipment similar to that used in the nuclear power the long term, plant facilities.The effects of an earthquake on facilities of this type is perceived to be more severe than at a Therefore, the N RC has initiated a program in this report-nuclear power plant because, in general, they do not have ing period whose objectives are: (1) to assist the NRC staff the stringent design requirements associated with nuclear in developing regulatory changes on the subject of seismic facilities.1he findings supplemented the existing experi-analysis of piping systems and perform supporting re-ence data base and provided additional insights into the scarch activities as needed; and (2) to evaluate the cumu.
performance of nuclear power plant structures and equip-lative impact of proposed changes on the overall safety ment during an earthquake. Proposals have been made to margins of the piping systems.
use an experienced-based approach for the seismic quali-fication of selected equipment in ALWR designs.
This program will be completed in 1995, allowing the staff to develop its position on the piping design requirements.
6.2.3.5 Shear Wall Ultimate Drift Limits 6.2.3.7 Cooperative International Seismic Programs
'Ihe ultimate drift limit is defined as the lateral displace-The NRC's participation m mternational seismic test pro-ment at the top of the wall relative to its base normalized grams is beneficial both for the sharing of research re-by the height of the wall. When performing seismic PRAs s urces and for gammg different perspectives on seismic and seismic margin assessments (SMAs), the ultimate design issues.The pooling of resources allows the devel-drift limit is necessary to estimate the seismic capacity of pment of larger-scale tests, an important element in the concrete nuclear power plant structures, in many cases, y lidation of methods for predicting the seismic response loss of equipment function has been considered to occur behavior of nuclear plant systems.
when the ultimate drift limits are reached; hence, the ultimate drif t limit is a failure parameter in these studies.
The Large-Scale SeismicTest(UST) facility s one of the i
Seismic 1 ras and SM As have been identified as accept.
1 rgest in the world for soil-structure interaction (SSI) able methods for performing the scismic portion of the research.The construction of a 1/4-scale model of a rem-IPEEEs for severe accident vulnerabilities. A research f reed concrete containment-10.5 meters m diameter program was started this fiscal year with the objectives of and 16.5 meters high (11.1 meters above the ground)-
establishing appropriate values of ultimate drift limit and w s completed in March 1993. All instrumentation was obtaining the statistics to define this parameter in a proba-completed by April 1993, and a formal dedication ceremo-bilistic sense, it is anticipated that the technical report will ny w s held in Huahen,'thtwan.
be published in the second quarter of FY 1994.
The UST program was initiated in January 1990 and is 6.2.3.6 Seismic Analysis of Piping expected to continue for 5 years.The goal of this program is to collect real earthquake. induced SSI data in order to The ASME Iloiler and Pressure Vessel Code, Section 111, evaluate computer codes used in SSI analysis of nuclear Nuclear Power Plant Components, Division i provides power plant structures. In the program, observations will rules for the design of piping systems m nuclear power be made on the motions of the reactorbuilding modeland plants. In general, the design rules have been proved over the surrounding ground during large-scale carthquakes.
the years to result in a design that affords reasonably The expectation is that the test model will be shaken by certain protection of life and property and provides a numerous earthquakes in this seismically active area of margin for deterioration in service so as to give a reason-Thiwan. Instrumentation h>cated on the scale model and ably long safe period of usefulness, n the field along a three-dimensional strong ground mo-tion array will record any observed data. The LSST pro-Recent developments in the nuclear industry have re-gram at Hualien,'lhiwan,is a follow-on to the SSI experi-sulted in proposed changes to the design philosophy of ments at Lotung,'Ihiwan, piping systems for the ALWRs. Both the industry and the NRC staff have gained knowledge from piping tests con-EPRI has organized the Hualien LSST experiment and ducted in the mid-1980's under NRC and industry spon-coordinated participation with the'lhiwan Power Compa-sorship and from the failure data base obtained from ny(thipower), the NRC, the Central Research Institute of actual seismic events at various facilities. The inherent Electric Power Industry (CR1 EPI), the 'Ibkyo Electric ability of welded piping systems to withstand extremely Power Company (PEPCO), the Commissariat a PEnergie large seismic inct tia kiadings is now recognized, and there Atomique (CEA). Electricite de France (EdF),
- 6. Safety issue Resolution Framatome, the Korea Power Engineering Co. (KOPEC).
6.3.3 Researcli Accomplisitmenis in FY 1993 and Korea Electric Power Corp.
6.3.3.1 Priorities of Generic Safety issues During this reporting period, a collaborative effort involv-The NRC revised the general methodology set out in the ing exchange of technical information was established 1982 NRC Armual Report for determining the priority of with the Ministry of International'Irade and Industry and GSis. An updated methodology (SECY-93-10S), Revised Nuclear Power Engineering Corporation (NUPEC) of Guidelines for Prioritization of Generic Safety Issues, was Japan. In this effort, NUPlic is carrying out a seismic approved by the Commission and incorporated in proving test program for a main steamline typical of the NUREG4033, "A Prioritization of Generic Safety Is-PWR plants and a feedwater system typical of the llWR sues." In December 1983, a comprehensive list of the plants. These tests will be conducted at the shake table of issues was published in NUREG4N33, and this list has
'lhdotsu Engineering 12boratory and will begin in late been updated semi-annually voith supplements in Junc 1994 and continue in 1995. Tests will be conducted for and December. The list of issues includes TMI Action several levels of seismic excitation and will use energy Plan (NUREG4660) items. The results of the NRC's absorber supports for the piping systems.The NRC in this continuing effort to identify significant unresolved GSis collaborative effort will carry out pre-and post-test analy-will be included in future supplements to NUREG-0933.
ses to assess the applicability of currently available analyt-ical models. In addition, data will also be obtained from During FY 1993, the NRC identified 5 new genericissues, NUPEC for scismic proving tests of a computer system established priorities for 12 issues (thble 6.1). and re-and a reactor shutdown cooling system.
solved 10 GSIs (Ihble 6.2). 'ihble 6.3 contains the sched.
ules for resolution of all unresolved GSis.
6.3 Generic Safety Issue Resolution h3.3.2 Resolution of Human Factor Generic Safety issues 6.3.1 Statement of Problem GSI IIF4.4 on procedures other than emergency operat-ing procedures was resolved while an effort contmued to develop a report describing work that relates to the use of in order to ensure the timely resolution of important safety concerns raised by the staff and outside sources, the procedures for low. power and shutdown operations. GSI HF5.1, on hical control stations, was resolved and a les-Commission directed the NRL staff to prepare a priority list of all generic safety issues, including iMI.related sons learned report on existing industry practices will be issues. The list was to be based on the potential safety produced. Activity continued on the development of an significance and cost of implementation of each issue. In advanced control room design review guide. The interac-tive version of the guide was demonstrated to the NRC December 1983, the origmal listing and procedures were approved by the Commission. Ihis guidance is reflected in staff. The final version is being developed for an applica-tion through Windows software. GSI IIF5.2 on annuncia-the NRC Poh,ey and Planning Guidance, the NRC Strate-gic Plan, and the NRC Fivedear Plan.
tors was resolved, and guidelines for the review of annun-ciators and alarm systems are being prepared for incorporation in t he advanced control room design review 8"
6.3.2 Program Strategy A generic safety issue (GSI) is one that involves a safety 6.4 Reactor Regulatory Standards concern that may affect the design, construction, or oper-ation of all, several, or a class of reactors or facilities and 6.4.1 Statement of Problem may have a potential for safety improvements and is-suance of new or revised requirements or guidance. Time-RES has the primary responsibility to manage, coordinate ly resolution of these issues is a major NRC concern. A reviews of, and control all NRC reactor related (materi-prioritization and management process has been estab-als-related for Section 7.1.1) rulemaking activities and to lished for identifying important issues for immediate monitor schedul?ng of such rulemaking to ensure that action, for eliminating non. safety.related or non-cost.
rules are developvi in a timely manner, in addition, RES effective and duplicate issues from further consideration, provides support for pieparation of the regulatoryimpact and for keeping the Commission and the public informed analyses (RI As) that accompany all rulemaking through of the resolution of these issues. Strategies for this pro-the development of generic methodology and guidance.
gram are to provide timely prioritization of proposed new Technical reviews of all RIAs are performed upon re-GSIs, eliminate the backlog of proposed issues (as re-quest. The NRC Regulatory Agenda Report and other sources permit), and issue periodic updates on the status management information systems associated with rule-and progress toward resolution of GSIs.
making activities are maintained.
6-9 NUREG-1266
- 6. Safety Issue Resolution Table 6.1 Generie Safety issues Prioritized in FY 1993 Number Title Priority 146 Support Flexibility of Equipment and Components RESOLVED LOW 149 Adequacy of Fire llarriers l
152 Design Basis for Valves'that hiight lie Subjected LOW to Significant Illowdown Loads 155.3 Improve Design Requirements for Nuclear Facilities DROP LOW l
156.3.6.2 Emergency DC lbwer 159 Qualification of Safety-Related Pumps While Running DROP on hiinimum Flow 160 Spurious Actions of Instrumentation Upon Restoration DROP of Ibwer i
161 Use of Non-Safety Related Power Supplies in Safety-DROP Related Circuits 162 Inadequate'Ibchnical Specifications for Shared DROP Systems at hiultiplant Sites When One Unit is Shut Down 164 Neutron Fluence in Reactor Vessel DROP 166 Adequacy of Patigue Life of Metal Components NEARIX REhuif/ED 168 Environmental Qualification of Electrical Equipment NEARIX RESOLVED Table 6.2 Generic Safety Issues Resolved in FY 1993 Number Title 105 Interfacing Systems LOCA at LWRs 120 On-Line'Ibstability of Protection Systems 14 2 Leakage Through Electrical isolators 143 Availability of Chilled Water Systems and Room Cooling 153 Loss of Essential Service Water in LWRs 11 - 5 6 Diesel Reliability llF4.4 Guide!Mes for Upgrading Other Procedures llF5.1 Local Control Stations llF5.2 Review Criteria for lluman Factors Aspects of Advanced Controls and Instrumentation I.D.3 Safety System Status hionitoring NUREG-1266 6-10
- 6. Safety Issue Resolution l
Table 6.3 Generic Safety Issues Scheduled for Resolution Scheduled lesue Resolution Number Title Priority Date 15 Radiation Effects on Reactor Vessel Sup;mrts IIIGII 03/96 23 Reactor Coolant Pump Seal Failures 111G11 12/94
!!.11.2 Obtain Technical Data on the Conditions inside the I!!Gli TilD TMI-2 Containment Structure 24 Automatic Emergency Core Cooling System Switch to MEDIUM 08/94 Recirculation 57 Effects of Fire Protection System Actuation on MEDIUM 12/93 Safety-Related Equipment 78 Monitoring of Fatigue'Itansient Limits for Reactor MEDIUM TilD Coolant System 106 Piping and Use of liighly Combustible Gases in MEDIUM 10/93 Vital Areas 11-1 7 Criteria for Safety-Related Operator Actions MEDIUM 09/94 11-5 5 Improve Reliability of'lhrget Rock Safety Relief Valves MEDIUM TBD 11-6 1 Allowable ECCS Equipment Outage Periods MEDIUM 12/94 83 Control Room Habitability NEARLY RESOLVED 12/93 145 Improve Surveillance and Startup'Ibsting Programs NEARLY RESOLVED 01/94 155.1 More Realistic Source 'Ierm Assumptions NEARLY RESOLVED 01/94 166 Adequacy of Fatigue Life of Metal Components NEARLY RESOINED TBD 168 Environmental Qualification of Electrical Equipment NEARIX RESOLVED TliD 11-6 4 Decommissioning of Nuclear Reactors NEARIX RESOLVED 10/93 1.D.5(3)
On-Line Reactor Surveillance Systems NEARLY RESOLVED 10/93 Needed reactor-related (materials.related for Section of this program are to (1) review the effectiveness of LWR 7.1.1) regulatory products, e.g., regulations and regulatory regulatory requirements and guidance and make recom-guides, are developed. Rulemaking is proposed or initi-mendations for revisions; (2) develop screening method-ated, as appropriate, and complex rulemakings that span ology to systematically review requirements and guid-the technical or organizational responsibilities of several ance; (3) coordinate and review proposed changes to the groups or that involve novel or complex questions of regu-IAEA safety standards; (4) develop or assist the develop-latory policy are managed. Petitions for rulemaking are ment of rules and regulatory guides; and (5) continue to investigated.
develop and maintain management information systems for rulemaking.
6.4.2 Program Strategy 6.4.3 Research Accomplishments in FY 1993 6.4.3.1 i ination of Requirements Marginal to
'Ihe purpose of the NRC nuclear regulatory program is to ensure that nuclear reactor (materials for Section 7.1.2) facilities are designed, constructed, and operated in a safe The NRC has instituted a program to eliminate require-manner. 'therefore, a continuing need exists to revise ments that are marginal to safety.The basic objective is to rules and guides and to develop new ones. The strategies avoid dilution of safety efforts by reducing resource f>-11 NUREG-1266
- 6. Safety Issue Resolution application to marginal safety issues.This improvement in
'Ihe Commission issued a final rulemaking on Jane 23, efficiency is expected to result in a net beneficial effect on 1993 (58 FR 33993),10 CFR 50.65, on monitoring the
- safety, effectiveness of maintenance at nuclear power plants.
- lhe proposed rule was published on March 22,1993 (58 As part of this program (57 FR 55156) to climinate re-FR 26938). The rule requires that the licensee conduct quirements that are marginal to safety and yet impose a maintenance activities once every refueling cycle but not regulatory burden that more directly enhances safety, the exceeding a period of 24 months. Ilecause of the quality NRC conducted a public workshop on April 27 and 28, and quantity of data, this will provide a greater assurance 1993, in Bethesda, Maryland. The purpose of the work-t hat the nuclear power plant will operate safely, shop was to provide information on the NRC program, solicit comments from the public and regulated industry The Commission issued a proposed rule on May 20,1993 on the program, and discuss a number of specific (58 FR 20336),10 CFR Part 55, on requalification require-initiatives being considered. The NRC encouraged the ments for licensed operators for renewal of licenses.The public and the regulated industry to attend the workshop proposed amendment would delete the requirement that and provide input to the NRC in the early stages of the cach licensed operator pass a comprehensive requalifica-program. In order to facilitate discussions at the work, tion written examination and an operating test conducted shop, advanced material on a framework for a by the NRC during the term of the operator's 6-year performance-based regulatory approach and applications license as a prerequisite for license renewal Forty-two to three specific rulemakings were published (58 FR 6196) comments were received, the majority of which supported prior to the workshop, the proposed amendments. It is expected that the final rule will be published in FY 1994.
Over 320 people attended the 2-day workshop, including 6.4.3.3 Regulatory Analys.is representatives from 44 utilities,5 industry groups,8 ven-dors,34 engineering and consulting firms,4 public inter-
'lhe Commission issued the proposed regulatory analysis est groups, and 6 State, Federal, and international govern-guidelines for public comment (58 FR 47159) on Septem-ment agencies. Representatives from an international ber 7,1993 (NUREG/BR-0058, Revision 2) The pro-union, law firms, and academia were also present. The posed guidelines represent the NRC's policy-setting doc-discussions at the workshap have been documented in ument with respect to RIAs. The document contains a
" Proceedings of the Workshop on Program for Elimina-number of policy decisions for the preparation of an RIA tion of Requirements Marginal to Safety" (NUREG/
performed to support NRC actions affecting reactor and CP-0129), dated September 1993.
non reactorlicensees.
1b implement this program, the NRC is currently taking Along with the guidelines, the NRC issued a draft report, action on requirements related to containment testing,
" Regulatory Analysis 'R'chnical Evaluation Handbook" fire protection, and quality asst.rance programs. The re.
(NUREG/BR-0184). The purpose of the handbook is to quirements in these areas will be modified to be less provide guidance to regulatory analysts, to promote prep-prescriptive and more performance-based to allow cost.
aration of high-quality RI As, and to implement the poli-effective implementation of regulatory safety objectives cies of the guidelines.The handbook expands upon the with marginal impact on safety. The NRC plans to use policy concepts included in the guidelines and translates probabilistic risk analysis technology and its safety goals in the six steps necessary to prepare an RI A into implement-reformulating requirements in these areas, able methodologies for the analysts.The guidelines and handbook establish the guidance and structure of the 6.4.3.2 Other Rulemaking existing operating procedures, the better to integrate backfit analysis requirements and safety goal policy con-1he Commission issued a final rulemaking on April 26, siderations.
1993 (58 FR 21904),10 CFR Part 50, on training and qualification of nuclear power plant personnel.The final Also to aid NRC analysts in preparing RIAs, the NRC rule amends the Commission's regulations to require published " Replacement Energy Costs for Nuclear each applicant and holder of a license to operate a nuclear Electricity Generating Units in the United States; power plant to establish, implement, and maintain pro-1992-1996" (NUREG/CR-4012, Volume 3), which up-grams that consider all modes of operation for the training dates replacement energy costs associated with of nuclear power plant personnel.The rule requires that short-term outages. 'Ihese estimates can be useful in the training programs be derived from a systems approach quantifying the overall impact of proposed regulatory ac-to training, as defined in 10 CFR Part 55.The objectives of tions when these requirements would necessitate retrofit-the rule are to codify existing industry practices related to ting and short-term outages at nuclear power reactors, personnel training and qualification and to meet the di-The NRC will continue to develop these methodologies in rectives contained in Section 306 of the Nuclear Waste an effort to facilitate NRC decisionmaking in evaluating Policy Act of 1982 (Public law 97-425).
the need and effectiveness of the regulatory actions.
NUREG-1266 6-12
- 6. Safety Issue Resolution During this report period, about 15 safety related RIAs ically consider the effects of electrical transients, includ-were completed or initiated to justify specific regulatory ing lightning strikes, on all electrical systems of nuclear actions for ructor and non-reactor licensees, power plants (PRM-50-56). On February 9,1993 (58 FR 7757), a notice of denial for rulemaking was issued.
6.4.3.4 hiainte. nance Rule and Regulatory Guide 6.4.3.5 Summary of Rulemaking Actions The purpose of the maintenance rule is to require com.
mercial nuclear power plant licensees to monitor the ef-During FY 1993,94 rulemaking actions were processed of fectiveness of mtintenance activities for safety-related which 22 rules were formally published,10 were termina-and certain non-safety-related plant equipment, as de-ted/ withdrawn, and 62 are ongoing (see 'Thble 6.4). Be-fined in 10 CFR 50.65, in order to minimize the likelihood sides the 62 ongoing rulemaking actions, there are 30 of failures and events caused by the lack of ef fcctive main.
potential rulemaking actions, and it is estimated that in tenance. The rule seguires that licensees monitor the FY 1994 there will be approximately 15-to-20 new rule-performance or condition of certain structures, systems, making requests requiring RES review and approval by and components (SSCs) against licensee-established the Executive Director of Operations.
goals in a manner sufficient to provide reasonable assur-ance that those SSCs will be capable of performing their intended functions. Such monitoring would take into ac.
Table 6.4 count industrywide operating experience. Where moni-Rulemaking Actions Processed During FY 1993 toring proves unnecessary, licensees would be permitted the option of relying upon an appropriate preventive Rulemaking Activities Number maintenance program.
Final Rulemakings Published 22 The following chronology outlines the completion of the Rulemakings Terminated / Withdrawn 10 process to issue regulatory guidance to implement the maintenance rule.
Ongoing Final Rulcmaking Actions 22 Ongoing Proposed Rulemaking Actions 37 In November 1992, the draft regulatory guide and o
Rulemakings on Hold 3
regulatory analysis for endorsement of the industry guidance document N UM A RC 93-01 forimplemen-tation of the maintenance rule was issued for pubhc
,g g 94 comment (FR 57 552S6). Eleven responses to the request for public comments were received. By the end of January 1993, the NRC staff had reviewed and resolved all public comments.
6.5 Radiation Protection and Health Effects o
in February 1993, the NRC staff made revisions to the regulatory guide in response to public comments.
6.5.1 Statement of Problem in April 1993, the NRC staff made presentations to the Advisory Committee on Reactor Safeguards and The N RC must provide reactor-related (materials-related the Committee to Review Generic Requirements on for Section 7.1.1) radiation protection standards and guid-the final regulatory guide.
ance that ensures that workers and members of the gener-al public are adequately protected from the adverse con-o On April 28, 1993, the NRC sent a letter to NU-sequences of exposure to ionizing radiation from licensed MARC to suggest improvements in their guidance activities. R ES reactor-related (materials-related for Sec-document resulting from the staff review of the pub-tion 7.1.1) activities needed to support the program in-lic comments on the draf t regulatory guide. By letter clude developing radiation protection standards; develop-dated May 13, 1993, NUMARC transmitted a re-ing guidelines for implementing these standards; and vised version of NUM ARC 93-01, dated May 1993 in pla n ning, developing, and directing safety research to pro-
- response, vide the information necessary for licensing decisions, inspection and enforcement activities, and the standards in June 1993, Regulatory Guide 1.160, " Monitoring development process. This includes analyzing available o
the Effectiveness of Maintenance at Nuclear Power scientific evidence to evaluate the relationship between Plants," which endorses NUM ARC 93-01, dated human exposure to ionizing radiation and radioactive ma.
May 1993, was issued.
terial and the potential occurrence of both late and early radiogenic health effects, including the radiation risk to A petition for rulemaking was received requesting that workers and the public, and estimates of the probabilityof the regulations be amended to require licensees to specif-increased incidence of cancer and genetic effects.These 6-13 NUREG-1266
- 6. Safety Issue Resolution analyses are used to provide bases for severe accident 6.5.3 Research Accomplishments in FY 1993 consequence analysis, probabilistic risk assessment (PRA), the development of safety goals and emergency The NRC maintains a program of research and standards plans, the identification of radiation protection problems, developraent in radiation protection and health effects the allocation of priorities for regulatory action, and envi-intended to ensure continued protection of workers and ronmental impact assessments. Recommendations of members of the public from radiation and radioactive such organizations as the International Commission on materials in connection with reactor licensed activities.
Radiological Protection (ICRP) and the National Council The program is currently focused on improvements in on Radiation Protection and Measurements (NCRP),
health physics measurements, identification and dissemi-Presidential guidance to Federal agencies, consensus nation of cost-effective dose reduction techniqu es, assess-standards, licensee performance indicators, cost and fea-ing health effects consequences of postulated reactor ac-sibility da+a, and available technical information also pro.
cidents, and monitoring health effects research.
t vide bases for developing regulatory and technical docu-ments related to radiation protection for workers and the 6.5.3.1 Revision of Part 20 Radiation Protection Stan-dards public.
Staff efforts in support of the implementation of the new 10 CFR Part 20 rule continued in FY 1993. These efforts 6.5.2 Program Strategy included development of training courses, publication of questions and answers on I art 20, and publication of regu-
'the Commission's egulatory proecss requires that safety
!atory guidance. Also, several minor corrective rulemak-enhancements to reactor (materials for Section 7.1.2) ings were completed.
rules and guidance be systematically screened to ensure that there is substantial inercase in public protection and Three new regulatory guides needed to implement the revised 10 CFR Part 20 were published.These guides are:
that based on analysis the costs are justified. Realistic values of the dollar-per-person-rem criterion are needed 1.
Regulatory Guide 8.9, Revision 1, " Acceptable Con-for analysis to justify changes, but gaps in knowledge asso, ciated with radiation health effects cause uncertainties in cepts, Models, Equations, and Assumptions for a these analyses.The strategies of this program are to iden-Ilicassay Program," was published in J uly 1993. The tify and compensate for uncertainties in radiation risk guide describes practical methods acceptable to the NRC staff for estimating intake of radionuclides us-coefficients used for health effect estimates in PRAs and ing bioassay measurement techniques.
regulatory decisions.
2.
Regulatory Guide 8.38," Control of Access to liigh When the Commission approved the whole body dosime-and Very High Radiation Areas in Nuclear Power try accreditation rule, they directed the NRC staff to Plants," was published in June 1993. This guide de-extend the rulemaking to include extremity dosimetry.
scribes a framework of graded radiation protection Therefore, the strategies of this program are to (1)im-procedures recommended to ensure that control for prove regulatory performance for radiation protection by access to high and very high radiation areas are ap-establishing measurement performance criteria and ac-propriate to the radiation hazard present in those creditation programs in the areas of extremity dosimetry, areas.
bioassay, and air sampling; (2) investigate effective new measurement techniques for these areas:(3) establish the 3.
Regulatory Guide 8.37, "ALARA Levels for Efflu-data base required for regulations; and (4) monitor specif-ents from Materials Facilities," was published in July ic indicators to detcet improving and declining licensee 1993. This guide provides guidance for materials li-performance.
censees only and is addressed later in this chapter.
6.5.3.2 rookhaven National Laboratory ALARA Federal guidance was approved by the President on occu-pational radiation protection. Further, the ICRP has pub-lished new recommendations for radiological protection.
The llrookhaven National Laboratory (IlNL) ALARA As a result of this new guidance, NRC reactor (materials Center, funded by the NRC, continued its surveillance for Section 7.1.2) regulations and regulatory guides will and dissemination of DOE and industry dose reduction have to be revised.The strategies of this program are to and ALARA research. BNL continued publication of the (1) modify radiation protection guidance and standards to series that abstracts national and international publica-be consistent with Presidential guidance on radiation pro-tions discussing dose reduction in areas such as plant tection requirements and (2) continue to monitor licensee chemistry, stress corrosion cracking, steam generator re-performance indicators by using the Radiation Exposure pair and replacement, robotics, and decontamination loformation Reporting System program.
(NUREG/CR-3469, Volume 7, July 1993). HNL also NUREG-1266 6-14 i
- 6. Safety Issue Resolution continued publication of the newsletter, "ALARA centration of the vanadous ion during chemical decon-Notes," on about a quarterly schedule. In 1993, llNL tamination of nuclear power plants.
focused on making the data base more easily accessible, adding information from overseas contacts, making final The NRC published " Enhanced Removal of Radioactive plans for an international conference on dose reduction, ihrticles by Fluorocarbon Surfactant Solutions" and continuing development of an ALARA handbook.
(NUREG/CR-6081, August 1993). The report provides
'lhe center provided information and advice on dose re-test results for the radiation stability and the application duction to NRC staff and licensees.
of environmentally compatible liquids to the nondestruc-tive decontamination of nucIcar equipment using ultra-sonics.
6.5.3.3 New Skin Dose Computer Code A revised computer code (VARSKIN II) for calculating 6.5.3.6 Performance Testing of Extremity Dosimeters dose to the skin from radioactive materials on the skin was
-Pilot l'st e
published (NUREG/CR-5873. December 1992). The re-The NRC published " Performance Testing of Extremity vised code is more flexible than earlier versions, allowing Dosimeters-Pilot 1bst" (NUREG/CR-5989, July 1993).
consideration of factors such as self-absorption, particle This report is the third of a series of tests run against the shape, and partic!cs on clothing.
draft performance standard for personnel extremity do.
simeters, ANSI N13.32, in order to establish the appropri-6.5.3.4 Occupational Exposure Data Systems ateness of the standard for use in dosimeter processing certification. The NRC presently requires licensees to The NRC continued to collect and process data in the become accredited or to use dosimeter processors accred-computerized data system called the Radiation Exposure ited under the National Voluntary Laboratory Accredita-Information Reporting System (REIRS). REIRS provides tion Program (NVLAP) operated at the Nationalinstitute a permanent record of worker exposures for reactors and of Standards and Tbchnology (NIST). At present, the several other categories of licensees. A report on 1991 NVLAP accredits whole body dosimeter processors and exposures, " Occupational Radiation Exposure at Com-will add accreditation of extremity dosimeter processors mercial Nucicar Power Reactors and Other Facilities, as soon as the standard is jointly approved for use by the 1991"(NUREG-0713, Volume 13, July 1993), was issued.
NRC and NIST Compilation of the statistical reports indicated that ap-proximately 200,000 individuals were monitored and half 6.5.3.7 National Institute of Standards'Ibchnology received a measurable dose. The average measurable dose dropped from 0.36 rem in 1990 to 0.31 in 1991.The Interagency Agreement, RES-93-01, between the NRC collective dose obtained from summing all the individual and NIST involves an ongoing study aimed at establishing doses dropped from the 1990 value by 20 percent to about traceability between NISTand the Pacific Northwest Lab.
32,000 person rems.The data base also includes exposure ratories (PNL) for neutron irradiations. PNL provides data on individuals who have terminated employment the neutron irradiation to NIST/NVLAP as part of its with certain licensees. Data on some 687,000 persons are duties as the testing laboratory for dosimeter processor accreditation run under the NVLAP.
in the system, most of whom worked in nuclear power plants. NRC continued to respond to requests for individ.
6.5.3.8 Electronic Personnel Dosimeters ual exposure data from the system. The data also assist m Ihe examination of the doses incurred by transient work-PNL is presently involved in developing a set of perform-ers as they move from plant to plant (about 2,900 in 1991).
ance tests and implementing procedures that would per-mit c!cctronic personnel dosimeters (EPDs) to be used in 6.5.3.5 Water Chemistry and Decontamination place of film or thermoluminescent dosimeters to estab-lish radiation doses for radiation workers.The product of Advanced Process 7bchnology determined the effects of this effort is to be a report that could be used by the NRC hydrogen water chemistry on radiation buildup in IlWRs to evaluate EPDs until such time as an appropriate ANSI and identified the most promising mitigating techniques.
standard for EPDs becomes available.This report would be used as the basis for a possible future certification The Idaho National Engineering Laboratory obtained in.
program to qualify EPDs for use in radiation meast.re-formation on out-of-core PWR power stations that will be
- ments, useful to NRR in evaluating system contamination (radio-nuclide surface concentrations and exposure rates)in ad-6.5.3.9 Gamma Dose Spectrometer vanced reactor designs.
Work is being carried out under a Small Business Innova.
tive Research Phase !! contract that involves the develop-OMNI' Itch International has developed an on-line UV-ment of a gamma-ray dosimeter / spectrometer that will Vis spectrometer that will be used to determine the con-measure the gamma ray spectrum over a wide range of 6-15 NUREG-1266 l
- 6. Safety Issue Resolution energies. From this information and the electronic signal a personal computer. It includes provisions for various retrieved from the dosimeter, it will be possible to calcu-dose calculations and can produce NRC Forms 4 and 5 in late, through the use of appropriate algorithms, the dose paper and electronic format. In addition, REMir can delivered to the skin. the eye, and the whole body.'Ib date, import and export data from ASCil and data base files, an Active Differential Absorption Spectrometer has been designed, developed, and tested.
6.6 Small Business Innovation 6.5.3.10 Spent Fuel Ifcat Removal Research The Oak Ridge National Laboratory is also continuing to Pursuant to the Small ilusiness Research and Develop-improve the data base m the guide for llWR and PWR ment Enhancement Act of 1992, Public Law 102-564, the fuel decay heat generation by meluding analysts of recent NRC supports the Small Business Innovation Research data to provide a basis for evaluating the adequacy of the (SBIR) program, which stimulates technological innova-storage system heat removal capability to limit fuel rod tion by small businesses, strengthens the role of small temperature.
business in meeting Federal research and development needs, increases the commercial application of NRC-6.5.3.11 Radiation Exposure Monitoring and Informa-supported research results, and improves the return on i
tion ' transmittal (REMrF) System mvestment from Federally funded research for economic A new software package, REMrr, for electronically re-and social benefits to the nation. The NRC has partici-porting radiation exposure measurements to the NRC was pated in the program since its inception in FY 1982, pro-made available (58 FR 41526; August 4,1993). REMir is moting high quality, " cutting edge" research of relevance designed to assist NRC licensees in meeting the reporting and potential inportance to the NRC mission. One goal of requirements of 10 CFR 20.1001 through 20.2401, as out-the program is to couple this research with follow-on lined in Revision I to Regulatory Guide 8.7," Instructions private funding, pursuant to possible commercial applica-for Recording and Reporting Occupational Radiation Ex-tion. As of FY 1993, the NRC was supporting 20 Sil!R posure Data." REMil is a menu-driven system for use on projects-in progress.
(>-16
PART 3--NUCLEAR MATERIALS LICENSING AND REGULATION SUPPORT l
l l
l 1
l L_.___
- 7. NUCLEAR MATERIALS 7.1 MMerials Regulatory Researcli The proposed action would improve the public health and 4
i safety by reducing the likelihood of unnecessary radiation 7.1.1 Statement of Problem exposur s rmn m active materials by ensuring that generally licensed devices are accounted for and disposed (See Sections 6.4.1 and 6.5.1.)
f pr perly, it is expected that the final rulemaking willbe completed in FY 1995.
7.1.2 Program Strategy A final rule (Appendix 11 to 10 CFR Ibrt 73) on day-firing (of firearms) qualifications for security personnel at cate-(See Sections 6.4.2 and 6.5.2.)
gory I fuel cycle facilities was published on August 31,1993 (58 FR 45781).This rule was needed to provide assurance 7.1.3 Research Accomplishments in FY 1993 that security force personnel maintain required weapon-handling and marksmanship skills by annual performance 7.1.3,1 Materials Licensee Performance testing.The rule is applicable to the specific security force personnel at facilities authorized to possess formula
'through its human factors regulatory research program, quantities of strategic special nuclear material.
the NRC seeks to improve its understar. ding and to main.
tain its requirements concerning the effect of human per-A final rule and a proposed rule (10 CFR 72.214) adding formance on the safety procedures myolving the medical casks VSC-24 and'IN-24 to the list of approved spent fuel and industrial use of nuclear materials.
storage casks were published on April 7,1993 (58 FR 17948)and on April 16,1993 (58 FR 19786) respectively.
Function and task analyses of the systems involved in The rules would increase the number of spent fuel storage teletherapy and remote after-loading brachytherapy were casks hn wM We holders of power reactor operatmg completed as a first step in better understanding the root lic nses can choose to store spent fuel under a general heense.
causes of human error associated with these systems. In-1 depth studies of procedures, training, human system in.
A final rule (10 CFR lbrts 26,70, and 73) on fitness for terface, and organizational policies and practices were also completed.13eports are being prepared on setting duty for category I facilities and shipments was published priorities of function and task performance problems re-in June 1993 (58 FR 31467).The rule amends the regula-lated to human enors m terms of their safety significance tions for the possession, use, or transport of strategic and an evaluation of alternative appmaches for resolving special nuclear material (SSNM). This action was neces-safety-sigmficant problems. Specific work on the human sary to ensure that specific employees of licensees who factors evaluation of the indostrial radiography system possess, use, or transport SSNM do not have a drug or was discontinued and is being reconsidered in view of the alcohol problem.
rule change to 10 CFR Part 34 established in 1993.
A final rule (10 CFR lbrt 73) on physical protection re-quirements at fixed sites was published in March 1993 (58 7.1.3.2 Materials Regulatory Standards CFR 13699). The rule clarifies the Commission's regula-The Commission issued a proposed rule (10 CFR Ibrts 30, tory intent that protection against both radiological sabo-40,50,-70, and 72) on January 11,1993 (58 FR 3515) that tage and theft of special nuclear material is not required would allow self-guarantee as an additional mechamsm at all facilities.This final rule also adds a requirement that for financial assurance. This prop 6 sed rule is in response protection be provided against radiological sabotage at to a petition for rulemaking (PRM-30-59) submitted by nonpower reactor licensees who operate at or above 2 the General Electric Company and Westinghouse Elec.
megawatts thermal, where deemed necessary,
- tric Corporation.'the rule would allow certain financially A proposed rule (10 CFR Ibrt 73) that would require a strong, non-electnc utilitylicensees to use self-guarantee as fmancial assurance for decommissionmg funding. It physical fitness program for secu rity personnel at category I facilities was published for public comment on Octo-would not apply to electrie utility licensees. It is expected ber 6,1993 (58 FR 52035). The rule would add new re-L that the final rule will be completed early in FY 1994-quirements for a physical fitness program and annual performance testing or a quarterly site specific content-A final rule (10 CFR lbrts 31 and 32)is under preparation based performance test. It is expected that the final rule-on requirements for the possession of industrial devices making will be completed in FY 1994.
containing byproduct material. This rule would require licensees to provide the NRC with specific information A proposed rule (10 CFR Fart 72) on reporting events at about radioactive material used under a general license.
Independent Spent Fuel Storage Installations (ISFSis) 7-1 NUREG-1266
l
- 7. Nuclear Materials l
and the Monitored Retrievable Storage (MRS)installa-when the source is shielded)and underwater irradiators in tion was published on September 14,1993 (58 FR 48004).
which the source always remains shielded under water
'this rule would ensure that significant events such as and the product is irradiated under water 'lhe rule does contamination events, personal mjuries, fires, and explo-not cover self contained dry. source storagc irradiator de-I sions at these facilities were promptly reported so that the vices, medical uses of scaled sources (such as teletherapy),
Commission could evaluate whether the licensee has tak-or nondestructive testing (such as industrial radiography).
en appropriate actions and whether prompt NRC action is The effective date of the rule was July 1,1993, necessary. The proposed rule would improve public health and safety by reducing the likelihood of unneces.
The N RC is now in the process of publishing for comment l
sary radiation ex;wures from these events. It is expected a draft regulatory guide, " Guide for the Preparation of that the final rulemaking will be cornpleted in FY 1994.
Applications for Licenses for Non-Self-Contained Irra-diators." which is related to the irradiator rulemaking.
A proposed rule (10 CFR Parts 30,32, and 35) on the The guide describes the information that an applicant medical use of byproduct material was published in July should submit for a new or renewed license application.
1993 (58 FR 333%). This action, taken in response to a issuance is scheduled for November 1993.
petition for rulemaking (PRM-35-9), is intended to pro-vide greater flexibility by allowing properly qualified nu.
Air Sampling. In September 1993, the NRC published clear pharmacists and authorized users who are physicians
" Air Sampling in the Workplace" (NUREG-1400). The greater discretion to prepare radioactive drugs containing report provides technical information on air sampling that j
bypraluct material for medical use. The proposed rule willbe useful for facilities following the recommendations would also allow research involving human subjects using in the NRC's Regulatory Guide 8.25, Revision 1, " Air byproduct material and the medical use of radio-labeled Sampling in the Workplace." This guide addresses air biologics. It is expected that the final rulemaking will be sampling to meet the requirements in NRC's regulations completed in FY 1994.
(10 CFR Part 20) on radiation protection. The report describes how to determine the need for air sampling A proposed rule (10 CFR Parts 30 and 35) cxtending the based on the amount of material in process modified by expiration date of the Interim Final Rule related to the the type of material, release potential, and confinement preparation and therapeutic use of radiopharmaceuticals of the materialJihe purposes of air sampling and how the was published in May 1993 (58 FR 26938), and a final rule purposes affect the types of air sampling provided are was subsequently published in July 1993 (58 FR 39130),
discussed.The report discusses how to locate air samplers This action allows licensees to continue to use byproduct to accurately determine the concentrations of airborne radioactive materials to which werkers will be exposed.
material under the provisions of the Interim Final Rule until the NRC completes a related rulemaking to address The need for and the methods of performing airflow pat-broader issues for the medical use of byproduct material tern studies to improve the accuracy of air sampling re-(including those issues addressed by the Interim Final suits are included. The report prescats and gives exam.
Rule). It is expected that the broader rule will be issued as pies of several techniques that can be used to evaluate a final rule in FY 1994. This extension of the expiration whether the airborne concentrations of material are rep-date was necessary to maintain the relief provided by the resentative of the air inhaled by workers. Methods for Interim Final Rule until the broader rule is issued.
adjusting derived air concentrations for particle size and methods for calibrating for volume of air sampled and estimating the uncertainty in the volume of air sampled
'the petitioner, Amersham Corporation, requested that are described. Statistical tests for determining minimum petition PRM-35-8 (add iridium 192 wire for the treat, detectable concentrations are presented. f low to perform ment of intestinal cancer) be withdrawn. The withdrawal anannualevaluation of theadequacyof theairsamplingis notice was published on August 23,1993 (58 FR 44466).
also discussed.
7.1.3.3 f aterials Radiation Protection and llcalth Ef-In April 1993, the NRC published "DEPOSU10N: Soft-ware to Calculate Ibrticle Penetration through Acrosol Irradiator Rulemaking. On February 9,1993, the NRC
'Iransport Systems"(NU REG /G R@06). DEPOSn10N published a final rule on licenses and radiation safety is user friendly software to calculate particle losses in requirements for irradiators (58 FR 7715)flhe rule estab-aerosol transport systems. Revision 1 to Regulatory Guide lished a new Ibrt 36 to specify radiation safety require-8.25 states that use of the DEPOSITION software is an ments and licensing requirements for the use of licensed acceptable method to calculate particle loss in acrosol radioactive materials in irradiators. Irradiators use gam-transport systems.This software was developed at'Ibxas A ma radiation to irradiate products to change their charac-
& M University under an NRC grant. Research will con-teristics in some way. The safety requirements apply to tinue there in FY 19940n the design of sampling probes to panoramic irradiators (those in which the material being minimize loss of particles at inlets and on where to place irradiated is in air in a room that is accessible to personnel sampling probes in ducts relative to bends and contrac-NUREG-1266 7-2
- 7. Nuclear hiaterials lions in duct diameter in order to obtain a representative health effects models and risk coefficients intended for
- sample, use in severe accident analyses, probabilistic risk assess.
ments, emergency responsa planning, and safety goal and Solubility of Particles in the Lung, in August 1993, a cost / benefit analyses. An addendum, "hfodification of research contract was awarded to the Inhalation'Ibxicolo.
Models 1(esulting from 1(ccent Reports of Ilealth Effects gy Research Institute on " Methods for Determining the of lonizing Radiation," was published in 1991. 'the reports Solubility of Itadioactive Materials in Grder to imple.
that led to modification of the models presented in ment 10 CFR Part 20."'Ihe new 10 CFR Part 20 lists NUREG/CR-4214 are the reports of the United States derived air concentrations (DACs) and annual limits on Scientific Committee on the Effects of Atomic Radiation intakes (ALIs) for inhalation of radioactive materials ac.
(UNSCEAR,1988), the National Academy of Sciences /
cording to their solubility in the lung. Compounds are National Research Council HEIR V Committee (NAS/
classified as "D" (soluble), "W"(moderately insol uble), or NRC, 1990), and the revised recommendations of "Y"(highlyinsoluble)basedon theirclearance half times.
ICRP-60 (ICRP,1991). A second addendum, "Modifica-This roughly translates to the time needed to dissolve in tion of Models Resulting from Addition of Effects of lung fluid, i.e., days (clearance half time less than 10 days)
Exposure to Alpha Emitters," was published in FY 1993.
for D class compounds, weeks (clearance half-time be.
Revision 2 of NUREG/CR-4214, Fnt 1, " Introduction, tween 10 and 100 days) for W class compounds, and years Integration and Summary," which incorporates the new (clearance half-time greater than 100 days) for Y class information presented in the two addenda was also com-compounds. The objective of the research is to identify pleted in FY 1993 and will be published in early FY 1994, methods that the NRC can accept for determining the This project is completed.
solubility in the lung of radioactive materials that may be inhaled so that licensecs can determinc internal radiation Embryo / Fetal Dose from Maternal intake. A study to doses to meet the requirements in 10 CFR Part 20.
improve understanding of the contribution of maternal radionuclide burdens to prenatal radiation exposure was ALARA 1.esels for Efuuents from Materials Facilities.
continued in FY 1993, with significant progress. Revision Regulatory Guide 8.37, "ALARA Levels for Efnuents 1 to NUREG/CR-5631, " Contribution of Maternal Ra-from Materials Facilities," was issued in J uly 1993. Section dionuclide Hurdens to Prenatal Radiation Dose"(March 20.1302(b) of 10 C FR Part 20 requires licensees to demon-1992), provides a method for calculatmg internal doses to strate compliance with the annual dose limit for members the embno/ fetus with an expanded data base of radionu.
of the public in 20.1301. In addition,10 CFR 20.1101(b) clides. Research that will permit inclusion of additional requires that licensees use procedures and controls to radionuclides, such as technetium, molybdenum, and ad.
achieve doses to members of the public that are as low as is ditional transuranic elements began in FY 1993 and will mnunue in W 1994.
reasonably achievable (ALARA). This document pro-vides guidance for materials licensees, such as medical and academic mstitutions, on acceptable methods of dem' An addendum," Relationships Between Annual Limitson tr it mpliance with this new mandatory ALARA intake and Prenatal Doses," to NUREG/CR-5631 was prepared in FY 1993 and will be published early in FY 1994.The methods and data developed under this project have been used by the NRC in preparing Regulatory Patient Release Criteria. A proposed tule to amend Guide 8.36, " Radiation Dose to Embryo / Fetus," which 10 CFR thrts 20 and 35 concernmg the entena for the release of patients admmistered radioactive matenal, as describes acceptable methods of compliance with 10 CFR well as a regulatory guide and a comprehensive regulatory 20.1208. 'lhis guide might be revised to incorporate the aformation presented in the addendum.'the methods analysis to be published as a NUR EG report, was drafted developed under this project are also useful in calculating and is expected to be published for comment in December doses in cases of accidental retcases of radioactive materi.
1993. this rulemaking action addresses the requests als.
of two petitions for rulemaking: PRM-20-20 from Dr. Carol S. Marcus and PRM-35-10 from the American Criticality and fuel Cycle Safety. A draft regulatory guide College of Nuclear Medicine.The petitioners requested for criticality safety was published for comment on Janu.
that the Comrmssion adopt a dose limit of 0.5 rem for individuals exposed to patients who have been adminis*
ary 25,1993 (58 FR 6022).This draft guide was developed to provide guidance to licensees on an appropriate nu.
tered radioactive material. It is expected that the final clear criticality safety training program for the use of rulemaking, as well as the regulatory guide and NUREG special n uclea r ma terial, especially the preven tion of criti.
report, will be completed in FY 1994, cality accidents, improvement of IIcalth Effects Models. "llealth Effects
'the les Alamos National Laboratory continued its ex.
Models for Nuclear Power Plant Accident Consequences amination and revision of TID-7016, " Nuclear Safety Analysis" (Revision I to NUREG/CR-4214) contains Guide," for simplification of use, evaluation against new 7-3 N U REG-1266
. ~
- 7. Nuclear Materials experimental data, and use of current computational description of the critical experiments modeled calcula-codes. 'Ihe document is a standard guide and reference tional results, quantification of trends in calculated k.
used tyj i he industry and the NMSS staff for initial critical-effective's for different types of experiments, and recom-ity safety evaluations.
mended calculational uncertainties to be applied.
The Oak Ridge National Laboratory (ORNL) continued The availability of a draft regulatory guide for the fire i
its methods validation of the criticality analytical se-protection of fuel facilities waspublished on April 22,1993 quences in SCALE-4 using ENDF/Il-V cross-section (58 FR 21606). This regulatory guide was developed to data. The validation effort will qualify the applicability of provide guidance to applicants and licensees with respect SCALE-4 to criticality safety problems covering the to the information needed for the preparation of the fire range of interest within the Fuel Cycle Safety liranch of protection sections (or chapters) of an application for a NRC/NMSS. The SCALE code system was developed at new license or for renewal of or amendments to an exist-ORNL for criticality, shielding, and thermal analysis of ing license for a fuel cycle facility.The guide also presents nuclear facility and package designs. The system is cur-a standard format for submitting this information.
rently used at ORNLin support of several tasks fundedby NMSS. In particular, SCALE-4 is used by ORNL and the 7.1.3.4 Uranium Enrichment NRC staff for criticality safety analyses relevant to licens-ing issues. Vaiid criticality safety analyses require vahda-The Commission is considering issuing a proposed rule-tion of both methods applied and the user who applies
- c. ding to amend 10 CFR Part 76," Regulation Governing them.The goal of this project is to validate the Criticality the Operation of Gaseous Diffusion Facilitics."This rule-Safety Analyses Sequences within the SCALE-4 system making is required by the Energy Policy Act of 1992 and by analyzing a large number of benchmark critical experi.
will establish both the procedural and technical require-ments whose parameters (enrichment, geometry, fissile ments for certification of the operation of the gaseous fuel / moderator ratio, etc.) cover the range of interest diffusion facilities by U.S. Enrichment Corporation. It is within the NMSS Fuel Cycle Safety Ilranch.The work will expected that a proposed rulemaking will be published be documented in a NUREG report that will include a carly in FY 1994, i
NUREG-1266 7-4
- 8. LOW-LEVEL WASTE DISPOSAL NRC research in support of licensing activitics for low-Disposal criteria for LLW have evolved as experience, level radioactive waste (LLW) disporal facilities centers knowledge, public awareness, and political controversy on (1) the safety and performance of engineered enhance-have grown. In particular, through the Low-Level Radio-ments and alternatives to conventional shallow land buri-active Waste Policy Amendments Act of 1985, the Con-al for LDV disposal and (2) evaluation of the overall gress has required the NRC to provide guidance for regu-performance of disposal systems. The NRC LLW re-latory decisionmaking regarding engineered LLW search program is described in NUREG-1380, published disposal methods.This change has broadened the scope of in 1989. NUR EG-1380 identifies issues, regulatory needs, NRC LLW research.
a strategy, and a schedule for resolving them.
NRC-funded LLW research is useful not only to the NRC licensing staff but also to the States regulating. LW dis.
8.2 Program Strategy L
posal. In order to make their research results available to the States, NRC research contractors, besides publishing NRC research in support of licensing activitics for LDV j
their work, gave presentations at meetings well attended disposal facilities is examining enhancemen ts and alterna-
)
tives to shallow land burial, LDV waste forms, infiltration by State representatives-such as " Waste Management 93" and the Annual DOE LUV Management Confer-of water, radionuclide migration in the soil, hydrology and ence. In addition, the NRC and the U.S. Geological Sur-contaminant transport, performance assessment, and
)
vey conducted a 3-day meeting at Reston, Virginia, with LLW source term modeling. 'Ihe NRC's LUV research State participation, on hydrology and geochemistry re-staff also prepares rulemakings that affect LLW disposal.
search addressing LDV concerns.
'Ihe diverse LLW regulatory user community makes the coordination and definition of LLW rescarch and the dis-8.1 Statement of Problem semination of associated products a much more compli-cated undertaking than similar activitics for the high-level waste (H LW) program. B ecause many Stat es are licensers Disposal of LDV involves issues concerning waste form of LLW disposal and are looking to the NRC for technical and waste package integrity, transport of radionuclides support in their LLW licensing and regulatory programs, through the disposal facility environment, and evaluation NRC's LDV research has to be more prescriptive and oflong-term doses from releases of radionuclides beyond developmental than the HLW research program.
the disposal facility environment. Research is required to establish regulatory criteria and license application assessment information to permit sound evaluation of 8.3 Researcli Accomplishments m, FY proposals for disposal facilities and to ensure that all 1993 regulatory requirements, particularly those on radionu-clide release limits, will be met. Perforrnin;, the needed research in a timely manner is made more urgent and 8.3.1 Materials and Engineering complex by two factors. First, the Low-Level Radioactive 8.3.1.1 Engineered Enhancements and Alternatives to Waste Policy Amendments Act of 1985 (P.L 99-240) sets a Shallow Land Hurial very tight schedule for establishing facilitics within indi-vidual States or compacts of States.
Many States and State compacts are considering alterna-tives to shallow land burial for the disposal of LLW. 5ever-Second, the States and compacts of States have chosen to al concepts have been proposed as alternatives, and, of contiir alternative disposal methods to shallow land these, the niost popular is the use of concrete as the WM Certain of these alternatives must be critically principal construction material for engineered barriers to contain LLW. The National Institute of Standards and owiM 'w tightly focused research to determine their acceptabiiny and to give guidance to the States and com' Technology (NIST) has continued investigating, on behalf of the NRC, the durability of concrete as an engineered pacts.
alterna tive to shallow land bu rial, while t he Idaho Nation-al Engiacering Laboratory (INEL) completed an evalua.
The direction of the LLW research program has re-tion of concrete barriers in limiting radionuclide transport sponded to legislative action, the changing policy of States (NUREG/CR4070). The NIST studies include concrete now responsib!c for disposal. and the lessons learned sulfate resistance rcsearch, the determination of diffusion from the history of shallow land burial of wastes at a coefficients for sulfate and chloride ions, the modeling of number of sites for several decades. Vague and differing stresses caused by sulfate attack in concrete, investigation criteria as to site suitability, waste package design, etc.,
of cracking in concrete, and the durability of superplasti-have been employed and may characterize future efforts.
cizers that may be used in concrete to reduce its transport 8-1 NUREG-1266
- 8. Low-Level Waste n, asal properties and improve its strength. 'Ihree reports are determine scaling factors for assessing hard to-measure being prepared as NUREG documents that will be avail-radionuclides in LLW, and to obtain activated metals from able in FY 1994; they describe (1)a new method to deter-operating reactors for teaching and field lysimeter re-mine chloride diffusion coefficients, (2) the determina-search studies. Studies were started at PNL to determine tion of sulfate diffusion coefficients, and (3) the modeling the presence of radionuclide-chelating complexes in lea-of stresses caused by sulfate attack in concrete. A new chates and behavior in soils.
effort was started at NIST to develop computer models on 8.3.1.4 Infiltration of Water the degradation of concrete. It is expected that a peer review panel report assessing these models and their
'the University of California at llerkeley,in cooperation application to LLW performance assessment will be is-wit h thc University of Maryland, is continuing to field test sued early in FY 1994, a variety of covers for LLW disposal units at the Maryland Agricultural Experiment Station in llettsville, Maryland.
(Results are reported in NUREG/CR-4918, Volume 6.)
8.3.1.2 Anion Retention in Soil Two designs are provmg to be particularly effective. One, called bioengineering water management, not only re-The University of California at Berkeley investigated the duced water infiltration to a negligible amount but also use of natural soil materials to retard migration of anions at radioactive waste disposal facilities. Most soils are dewatered two experimental cells. Since this is a surface cover, it lends itself to use as a remedial action cover for much more effective in retarding the migration of cations sites susceptible to subsidence.The New York State Ener-than anions. Certain long-lived radionuclides such as 1-129 andle-99 may be in an anionic form at LLW dispos, gy Research and Development Adm, istration finished m
construction m 1993 of a bioengineering water manage-al facilities.'the anticipated application of this work is to ment cover over such a trench at the West Valley LLW identify materials that could be used to condition the near disposal facility, and the monitoring of performance has field either in or around LLW disposal facilities to retard migration and attenuate activities of radionuclides in just begun. A second cover consists of a resistant layer anionic form. A literature review (NUREG/CR-5464) barrier (compacted clay) over a conductive layer barrier.
This second system has functioned perfectly since its com-indicated that a group of soils called andisols, which are derived from the weathering from volcanic parent materi.
pletion in January 1990.1Iowever, its long-term perform-ance needs to be assessed.
al, have significant potential for retardation of anionic forms of I-129 and 'It-99. Field work is under way in the PNL has developed an " infiltration evaluation methodol-Western United States to determme if exploitable depos-ggy.. (NUREG/CR-5523) that is being tested on the Las I
its of andisols with anion exchange capacity are available.
Cruces Trench field data set (NUREG/CR-5998) in the Preliminary results were published (NUREG/CR-5974)
INTRAVAL project. The incorporation of various one-and work is contmumg.
and two-dimensional analyses has also been applied and tested for conducting infiltration analyses for perform-8.3.1.3 LLW Waste Forms ance assessment and engineering design analysis (NUREG/CR-6114).
The stability of decontamination waste obtained from nu-clear reactors using commercial decontamination pro-8.3.2 Hydrology and Geochemistry cesses and solidified in cement is being studied. Decon-8.3.2.1 Radionuclide Migrat. ion m Soil taminated LLW (collected from the Peach Hottom nuclear power station)is being tested at INEL The stu-A significant area of uncertainty in predicting site per-dies are aimed at ensuring that radionuclide and chelating formance is the degree to which soils can retard radionu-agent leaching characteristics, as well as the compressive clide migration. Tb reduce this uncertainty, the NRC is strength of the cement solidified waste, are consistent investigating mechanisms controlling radionuclide move-with NRC technical positions and the requirements of ment through soils. The Sandia National Laboratories 10 CFR l' art 61 for waste form stability. Field lysimeter (SNL) are working on characterizing retardation mecha-studies containing radioactive ion exchange resins solidi-nisms. The University of California at Davis is investigat-fied in cement and vinyl ester-styrene are being con-ing the mechanisms and rates of dissolution for a variety ducted at the Oak Ridge and Argonne National Laborato-of silicate minerals.This will be useful for understanding ries to determine radionuclide rclease rates under certain and modeling processes occurring on mineral surfaces environmental conditions. Studies continued at INEL to that affect both sorption and leaching. PNL is cxamining 4
i i
l investigate biodegradation of LLW by micro-organisms to the role in radionuclide transport played by microparticu-ensure stability requirements, as required by 10 CFR lates and naturally produced organic complexants.
Part 61. Studies continued at the Pacific Northwest Labo-8.3.2.2 Hydrology and Contaminant'Ransport ratories (PNL) to investigate activated metal and radioac-tive waste streams for radionuclides not included in the PNL has evaluated and developed a data set from j
listing of long-lived radionuclides in 10 CFR Part 61, to an earlier field study involving subsurface injection of 1
j NUREG-1266 8-2 1
- 8. Low-Level Waste Disposal j
radioactive tracers in heterogeneous unsaturated porous facilitics for LLW are applicable to aboveground disposal media at the llanford site. 'lhe data sets reported in (i.e., built on the ground without an carthen cover) was NUREG/CR-5996 cover a period of 10 years and will published in June 1993, allow confirmatory analyses of existing flow and transport models that are to be used in LLW performance assess.
A petition for rulemaking (PRM-61-2) from the New ment. Work is being completed by the Massachusetts In-England Coalition on Nuclear Pollution was published in stitute of"Icchnology (Mrr) and Princeton University on the Rderal Register on July 23,1992. The petitioner re-the application of stochastic methods for simulating flow quests that the Commission amend its regulations regard-and transport in heterogencous soils. In particular, Prin.
ing waste classification of low-level radioactive waste to ceton University has comp!cted and is now testing restrict the number and types of waste streams that can be ground water ventilation models for simulating vapor disposed of in near-surface disposal facilities. Recom men-phase t ransport associated with LIAV facilitics, and Mrfis dations on the need for rulemaking will be determined by applying their stochastic approach to field data sets.
December 1993.
833 Compliance, Assessment, and Modeling 83.5 Environmental Policy and Decommis-8.3.3.1 Performance Assessment 8.3.5.1 Enhanced Participatory Rulemaking Research is continuing on a performance assessment methodology. Emphasis is being given to engineering en-An Enhanced Participatory Rulemaking on radiological hancements to shallow land burial, specifically the incor-criteria for the decommissioning process was initiated to poration of concrete degradation models and the per.
actively solicit early input from a wide spectrum ofinter-formance of cover materials. SNL has evaluated the ests. Seven public workshops were held across the United current status of models used in performance assessment States (Chicago, San Francisco,lloston, Dallas, Philadel-and published the results (N UREG/CR-5927, Volu.me l),
phia, Atlanta, and Washington, D.C.), and these were SNL is currently making improvements to the perform.
followed by cight generic environmental impact state-ance assessment methodology and assessing the valida.
ment (GEIS) scoping meetings in foor citics(Washington, tion approaches for the performance assessment models.
D.C., San Francisco, Oklahoma City, and Cleveland).The Mrf has been investigating the use of stochastic methods Environmental Protection Agency participated in these for dealing with large-scale non-uniformity of site hydro.
workshops and meetings and is a cooperating agency in logic characteristics The University of Arizona and New the development of the GEIS.The staff began evaluating Mexico State University are working cooperatively with the input received and worked on preparing a staff draft of Mff by providing a field test at Las Cruces, New Mexico, decommissioning criteria.The staff plans to complete the of Mrl"s theoretical work.
draft early in FY 1994 and continue seek.ing early input by releasing it to Agreement States and other interested 8.3.3.2 LIAV Source Erm Modeling parties in January 1994. The proposed rule is scheduled for publication in mid-1994.
During FY 1993, the existing LIAV source term code,Ill;r (breach, leach, and transport), was benchmarked against The staff will continue to build experience with the use of field data and verified. Extensions to incorporate addi-the methodology described in the Jmft " Manual for Con-tional geochemistry and gaseous reicase are currently ducting Radiological Surveys in Support of LicenseTermi-being investigated and planned for inclusion in the code nation"(NUREG/CR-5849)in the context of the present during the next fiscal year, and proposed criteria for unrestricted release.
8.3.3.3 Modeling of"Iritium Migration at Arid Sites The NRC issued " Residual Radioactive Contamination The Um.versity of California at Ilcrkeley, work, g coop-from Decommissioning: Technical llasis For Translating m
cratively with CSIRO, developed a three-dimensional de' Contamination Levels to Annual'lbtal Effective Dose 1
terministic model based on soil physics to predict tritium Equivalent" (NUREG/CR-5512, Volume 1) in October migration at arid disposal sites (published as NUREG/
1992. The complete report will consist of three volumes CR-5980). The model awaits confirmation through the and one supplement. This first volume is designed to use of a controlled release from a known source at an arid provide screening models, ma'hematical formulations for site. Planning for such an experiment is under way.
the screening models, and referenced parameter valucs for estimating doses, above natural background, to indi-viduals from residual radioactivity associated with lands 83.4 Low Level Waste Regulatory Standards and structures after decommissioning licensed facilitics.
The modeling structure permits the use of either generic A final rule to amend 10 CFR Part 61 to clarify that or site-specific parameters to be used as screening esti-requirements related to the performance of land disposal mates of radiation doses from rnultiple environmental 8-3 N UR EG-12ff>
- 8. Low-Level Waste Disposal pathways. It is expected that the software to implement rule of June 8,1993, was prepared for Commission consid-the models, D&DSCREEN, will be released for testing cration and is expected to be published early in FY 1994.
late in FY 1994 and will be accompanied by the user This rule would address the timeliness of completion of manual, NUREG/CR-5512. Volume 2.
radon barriers on uranium mill tailings and verification of the effectiveness of those barriers.
8.3.5.2 R.gulation Development Responding to the Energy Policy Act of 1992, the Com-A proposed rulemaking (10 CFR Parts 30,40,70, and 72) mission's Below Regulatory Concern Policy Statement on timeliness in decommissioning a materials facility was was withdrawn. The withdrawal notice was published on published on January 13,1993 (58 FR 4099).The proposed August 24,1993 (58 FR 44620).
rule would amend the Commission's regulations to estab-lish timeliness criteria for decommissioning nuclear sites The petitioner, University of Utah, requested that its or separate buildings or areas following permanent cessa-petition, PRM-20-14 (disposal of biometical waste con-tion of licensed activitics. The principal effect of these taining small amounts of radioactivity) be withdrawn.The amendments is to formalize and codify the NRC's re-withdrawal notice was published on July 22,1993 (58 FR quirements for timeliness in the decommissioning of ma-39173).
terials facilities. Seventeen comment letters were re-ceived, and the final rulemaking should be completed in 8.3.5.3 Other Decommissioning-Related Actions FY 1994.
In September 1993, the NRC awarded a Small Business innovative Research Phase 11 contract to Shonka Re-A final rule (10 C FR Parts 30,40,70, and 72) was published search Associates for a " Contamination Monitor Using on July 26,1993 (58 FR 39628) to amend the NRC's de' Visual Identification." The monitor, using a position-sen-commissioning regulations to require holders of a specific sitive proportional detector to detect contamination, is license for possession of byproduct material, source mate-mounted on wheels and ccm be pushed over the floor like a rial, special nuclear material, and independent storage of lawn mower. The output from the detector goes to a spent nuclear fuel and high-level waste to prepare and computer, which creates a visual display in goggles worn maintam additional documentation identifymg areas by the operator. As the operator pushes the monitor over where licensed materials and equipment were stored and the floor, a " virtual reality" display of contamination can used.The Commission's intent is to provide both the NRC be seen on the floor.The monitor is fasterand less expen-and the hcensee with the necessary mformation to ensure sive than conventional contamination detectors.
complete decommissiomng of licensed facilities. This ac-tion is consistent with similar requests made at the Synar Brookhaven National Laboratory continued its determi-Committee hearing on decommissioning and an earlier nation of technical and safety criteria that should remain GAO report.
as part of decommissioning regulations under 10 CFR Part 50 when a licensee initiates action to permanently A final rule (10 CFR Part 20) on disposal of waste oil by shut down the nuclear reactor in preparation for decom-incineration at nuclear power plants was published on missioning activities. This project will develop a compari-December 7,1992 (57 FR 57649).The rulemaking action, son of the safety requirements for a shutdown versus an responding to a petition for rulemaking originally filed by operating nuclear power reactor after the reactor has Edison Electric Institute and the Utility Nuclear Waste permanently shut down. It will also perform financial Management Group (PRM-20-15), allows reactor licens-assurance analysis for offsite liability requirements for ces to pursue the option of incineration of waste oils shutdown reactors. It will examine the environmentalim-contaminated with small amounts of radioactivity without pact of the potential increase in the spent fuel transport the need for specific authorization.
and radiological exposure to the public in the event the licensees prefer to ship and store their spent fuel. It is An advance notice of proposed rulemaking (10 CFR expected that a draft report for public comment will be Part 40) concerning updating requirements for licensing issued early in FY 1994.
of source material was published on October 28,1992 (57 FR 48749). The contemplated rulemaking would improve PNL continued providing support for nuclear facility de-the control of source material through more specific regu-commissioning issues. An update of waste burial costs, lation and update applicable requirements to conform
" Report on Waste Burial Charges," was published in May with the revised standards for protection against radi-1993 as Revision 3 to NUREG-1307." Revised Analysesof ation. A broad range of issues are being considered in the Decommissioning for the Reference Pressurized Water categories of exemptions, general licenses, specific li-Reactor Power Station" (NUREG/CR-5884) was pub-censes, and mills and mill tailings. One specific issue iden-lished for public comment in September 1993.This report tified in the advar we is being handled in a separate re-evaluated decommissioning costs for an earlier decom-rulemaking. A g
-J rule conforming NRC's regula-missioning study of a reference PWR reactor. " Estimating tions governir num mill tailings to a proposed EPA Pressurized Water Reactor Decommissioning Costs" NUREG-12u, 8-4
- 8. Low-Level Waste Disposal (NUREG/CR-6054) was issued in October 1993. This tered at LWRs at the time of decommissioning, using report describes the computer program used for estimat.
actual field sampling and theoretical analysis. PNL is also ing reactor decommissioning costs in the NUREG/
updating and extending an information base on the tech.
CR-5884 reevaluation study.
nology, safety, and costs for decommissioning fuel cycle PNL continued developing an information base on the and non-fuel cycle nuclear facilities, using actual decom-actual radioactive contamination expected to be encoun-missi ning data and analysis.
1 l
1 8-S NUREG-1266
~
PART 4--ASSESSING THE SAFETY OF IIIGII-LEVEL WASTE DISPOSAL I
- 9. IIIGII-LEVEL WASTE RESEARCII The NRC maintains active rescarch programs in rock program objective is to provide the technical capability mechanics and engineering, hydrology, geology, waste necessary to evaluate D.OE's site characterization activi-package performance, materials science, geochemistry, ties as required by the NWPAA and to assess DOE's and several other disciplines related to the management license application when it is submitted.
of high-level waste (iILW).The research combines thco-retical study with laboratory and field experiments to im-provc understanding of the physical processes that control 9.2 1*rogram Strategy and determine repository performance in the unsaturated volcanic tuff at the Yucca Mountain (Nevada) site cur-The research program has been guided by the need to rently under consideration by the Department of Encry provide the technical foundation for NRC development (IXm) as directed by the Congress in December 1987.
of a set of regulations and a licensing process for the The ultimate goal of the NRC's HLW management re-review and licensing of the HIM repository.This frame-scarch is to provide the technical bases for the hcensmg work for NRC review will allow the formal licensing activi-staff to make mdependent judgments as to the appropri-ties and the supporting research to be focused on the significant technical issues.
ateness and adequacy of DOE's demonstration of com-pliance for the HLW repository with NRC requirements and witn the Environmental Protection Agency's HLW At present, the NRC has active research programs in standard. Key technical issues being addressed include hydrology, geology, materials science, geochemistry, and unsaturated flow and transport mechanisms, assessment several other disciplines related to HLW management, of the potential for volcanic and scismic events, geochem-The research combines theoretical study with laboratory scal processes, and the long-term performance of engine-and field experiments to identify and quantify the physical cred waste isolation systems.
processes and phenomena important to waste isolation so that the NRC can assess repository performance and quantify the uncertaintics associated with characteriza-g Statement or i3roblem tion and measurement of these processes. All this work is
)..i intcy,,tco into on inacpcnaco, now pc,fo,mancc The 1ILW disposal policy for the United States is defined assessment methodology. Effort is also required to vali-by the Atomic Energy Act, the Energy Reorganization d te manyof the models that underlie the methodology.
The ultimate goalof tt e NRC's HLW research program is Act, the Nuclear Waste Policy Act, and the Nuclear Waste to provide the techm} cal basis to support the licensing Policy Amendments Act (NWPAA). The last, signed into law in 1987, provides for the development of a geologic staff s mdependent review of the appropriateness and repository for the permanent disposal of high level radio-adequacy of DOE s demonstration of comphance with active waste in the State of Nevada at Yucca Mountain 10 CFR Part 60and the EPA's HLW standard. In addition, and assigns responsibility for repository development to NRC's waste management research secks to provide tech-the DOE. According to the Federal Government's Reor-nical support to the licensing staff m their mteractions ganization Plan No. 3 of 1970, HLW environmental stan-with DOE, the State of Nevada, and other participants dards dcvelopment is the responsibility of the Environ-and interested parties and to develop regulatory stan-mental Protection Agency (EPA), and the Ener d rds to support the liensing of the disposal and manage-Reorganization Act assigns the regulation of HLW dispo'gy ment of high-level radioactive wastes, s-al to protect public health and safety and the environment to the NRC.
9.3 Research Accomplishments in FY An HLW repository poses problems involving regulatory considerations and uncertainties related to waste em-placement, monitoring, and per formance assessment that 9.3.1 E.ngineered Systems Research are unique in the history of the NRC. Much of this 9.3.1.1 Stability of Underground Opening uniqueness stems from fl>e type of facility, first-of-its kind geologic disposal installation, its very long performance When specifying suitabic site conditions for an HLW re-time (specified as 10,000 years by the EPA), and the fact pository,10 CFR Part 60 specifically requires consider-that it willlm placed in low permeability / low flow geologic ation of natural phenomena and site conditions that could systems that he not been investigated previously be-adversely affect achievement of the prescribed perform-cause of their low economic value. The NRC must have an ance objectives. An important phenomenon that could independent capability to evaluate the DOE safety analy-affect both the short-and long-term performance of a ses and decide whether long-term releases predicted by repository is grou nd motion resulting from seismic activity DOE will be within established limits.The NRC research or motion caused by underground nuclear explosions at 9-1 NUREG-1266 j
)
- 9. liigh-Level Waste Research l
the Nevada 'li:st Site. Ground motion from either source to permit the NRC to judge whether test data and models could cause rock displacement and pressure changes in offer reasonable assurance of compliance with regulatory
)
ground-water levels that could violate repository per-requirements During 1993, the CNWRA conducted re-t formance objectives.
search on Walized corrosion rates for candidate IILW package maierials in tuff ground waters and continued to
'Ib investigate the effects of seismicity on the under.
evalua'e potential regulatory problems arising from stress ground openings for an 1ILW repository, the NRC is spon-corrosion cracking and metallurgical instability. Research soring research at the Center for Nuclear Waste Regula-was also initiated on the effect of water refluxing (due to tory Analyses (CNWRA). The research includes the t hermal gradients) on degradation of waste package mate-laboratory characterization of jointed fractured rock ex-rials.
pected at the llLW repository horizon, the assessment of 1
computer codes to calculate rock response to carth-The CNWRA continues investigations of contaminant quakes. and field studies at the Lucky Friday Mine, Idaho, transport and material corrosen on ancient Minoan cop-to measure rock displacements and ground-water re-per, bronze, and lead artifacts that were buried under sponse to mine seismic events. Results from the study silicic tuff 3,600 years ago. New metallie artifacts were indicate that underground openings at high states of stress uncovered, and the tuff immediately adjacent to the arti-are more sensitive to seismic loads than previously fact was sampled to determine the extent of elemental thought and that repetitive carthquake loading causes the transport in the unsaturated tuff. Ilydrologic tests were displacement of rock joints, making the rock mass less conducted at the site in an attempt to model the hydrolog-stable. Seismic events of even small magnitudes cause ic conditions to which the metals have been exposed.This changes in ground-water pressures as a result of volume work is providing data related to ongoing modeling re-changes in the rock. Rock / ground-water response to scis-scarch in copper alloy corrosion on unsaturated environ-mie events greater than those at the Lucky Friday Mine
- ments, will be studied at the Garner Valley site in Cahfornia.The California site can be subject to seismicity up to magni-93.2 Geologic Systems Research tude 6.5.
9.3.2.1 Ilydrogeology 9.3.1.2 Thermohydrological-Mechanical Coupled In-Since ground water is considered to be the primary agent teractions of radionuclide transport from an 11LW facility to the Onc important component of the safety analyses for liLW accessible environment, the NRC is actively studying disposal is the coupling of the interactions between the ground-water infiltration, recharge, flow, and transport rock mass, the ground water, and the thermal stresses processes in partially saturated fractured rock. An exper-induced by the high-temperature wastes. Coupling of the imental site in unsaturated fractured tuff, similar to Yucca processes implies that one process affects the initiation Mountain, called the Apache Leap 'Ibff site, has been and progress of the other, and independent consideration instrumented and characterized for testing instrumenta-of each process is bound to be flawed. The NRC is a tion, methods, and analyses similar to those being used or participant in an international multidisciplinary and proposed by DOE. In particular, field data sets are being cooperative research effort to study the coupled collected and interpreted for testing mathematical mod-thermohydrological-mechanical (Film) processes under cls of flow and transport in partially saturated fractured the acronym DECOVALEX (Development of Coupled rock. This work has been incorporated into the interna-1 Models and their YALidation against Experiments).'Ihc tional project called INTRAVAL for model validation of objectives of the study are to increase the basic under-ground water transport models using field experiments.
standing of TiiM coupled processes, support the applica-Key technical uncertainties dealing with determination tion of codes for TIIM modeling for jointed hard rocks, and confirmation of ground-water travel times, presence and design validation experiments by means of TIIM and influence of perched-water systems, and preferential model studies. In FY 1993, twobenchmark problems sim-flow duc to persistent discontinuitics are being addressed.
ulating near-field conditions in an IILW repository were modeled, analyzed, and compared to results obtained by Scientists at the CNWRA in San Antonio, Texas, are researchers from various countries using different com.
testing approaches to large-scale unsaturated flow in het-puter codes. There was good agreement in the results. It is erogeneous, stratified, and fractured geologic media.The expected that the DECOVALEX study will be published BIGFLOW code has been developed, tested, and docu-in the International Society of Rock Mechanics Journal.
mented for use in their stochastic methodology (NUREG/CR-6028) for simulating flow in variably satu-rated, heterogeneous geologic media. 'Ib quantitatively j
9.3.1.3 Matenals Science account for spatial heterogeneity, CNWRA has applied An understanding of the materials science aspects of the the real space renormalization group method for parame-engineered barriers in IILW disposal systems is necessary ter estimation. Work has begun on an analysis of the NUREG-1266 9-2
- 9. liigh Level Waste Research regional hydrogeologic processes in the vicinity of the cesses. Assessments of thermodynamic data for key min.
Yucca Mountain focusing on the appropriate integration crats (/colites) that strongly affect the ground-water of hydrogeologic, geophysical, and geochemical informa-chemistry of Yucca Mountain were completed by the tion and methods for testing alternative mathematical CNWRA. Precise thermodynamic data and modeling models of the regional flow and transport system.
showed that dissolution of the mineral analcime controls Yucca Mountain water chemistry. Research on the ex-
'lhe validity of conceptual and numerical models used to change of ions in solution with accessible sites in zeolite describe ground-water flow and radionuclide transport minerals continued to show the success of an ion-ex-for various hydrogeologic settings is being evaluated in change model applied by the CNWRA to model interac-the INTRAVAL project. The NRC staff and research tion between zeolites and a number of dissolved fission contractors from the CNWRA, University of Arizona, product radionuclides. CNWRA rcscarch showed that the Sandia National Laboratories, Massachusetts Institute of behavior of dissolved actinide elements could not be char-l
- Ibchnology, Princeton University, and Ibcific Northwest acterized well by simple linear models, and research con-Laboratories are participating in this international effort tinucs on improving assessment methods for these impor-involving 13 scientific parties from 10 countrics.
tant radionuclides. An extensive evaluation of more complex adsorption models was completed, and experi-Cooperative experiments and data analyses being done ments were started to measure key sorption data.
under a cooperative agreement between NAGRA (Swit-zerland) and the NRC, negotiated during FY 1987, con.
From 1988 to 1993, the United States, represented by tinue to augment the ficid-testing program cited above.
NRC, was one of five countries participating in the Inter-national Alligator Rivers Analog Project (ARAP). "Ihis Research was completed on a grant at Johns llopkins project investigated the Koongarra uranium ore body in University to investigate numerical coupled thermohy.
Australia. This ore txxly is a research site relevant to drogeochemical modeling of large-scale transport pro.
nuclear waste disposal because it has been subjected to cesses that led to the formation of unconformity-type dissolution and transport over the past one million years uranium ore lxidies such as those found in the Alligator by oxidizing ground water flowing through rock fracturcs.
Rivers region of Australia and the Athabaska Ilasin in A zone of dispersed uranium and other radionuclides has Canada. The model was able to simulate the detailed formed along the ground-water flow path. Processes con-thermal, hydrological, hydrogeochemical, and mineralog.
trolling the long-term release and transport of uranium ical conditions and processes and the detailed final shape, and other radionuclidescan beintensivelyinvestigated. A structure, and mineralogy of the Cigar Lake, Canada, 2-day final project presentation was given in October 1993 uranium ore lxx!y. 'this research model has the potential to the OECD Natural Analogues Working Group in'Ible-for providing a basis for major improvements in radionu, do, Spain. Mineralogical and hydrogeochemical analysis, clide transport modeling by performance assessment combined with numerical geochemical equilibrium and models.
kinetic modeling, showed that many of the same minerals and geochemical processes as those found at Koongarra 9.3.2.2 Geochemistry will likely control radionuclide release and transport at Yucca Mountain. The project showed that integration of Geochemistry is involved in the assessment of all the multidisciplinary geophysical, geological, geochemical, IILW performance objectives and many of the siting and and hydrological data is necessary to describe a complex design criteria. llecause the chemistry of the proposed site site in which ground water flow occurs by both matrix and 1
is dominated by the chemical effects of an enormous tnass fracture flow. Numerical geochemical modeling power-of rock compared to a relatively small mass of waste and fully described and predicted long-term evolution of the engineered components, knowledge of geochemical ef-Koongarra site, but the need for beller geochemical ther-fccts on the behavior of the repository system and its modynamic data for uranium silicate minerals (which will components is essen tial. Over t he long term, the per form-also tend to form at Yucca Mountain) was identified. Key ance of engineered components will depend on their com-uncertainties in making long-term predictions at Koon-patibility with the geochemistry of the site. Moreover, garra involve characterizing long-term climatic and hy-there are geochemical clues to the past history of the site, drogeologic conditions. Performance assessment models including former site temperatures, the histories and ages were the subject of considerable investigation and discus-of ground water samples, and thc ages of prehistorie geo-sion. Some investigators concluded that avmlable per-logic events such as carthquakes and volcanic eruptions.
formance assessment models were more sensitive than The NRC has a geochemistry research program actively the Koongarra site to hydrological processes and less sen.
investigating key technical uncertainties related to 1ILW sitive to geochemical processes, it is cicar that these mod-disposal safety.
cls would benefit from further research and development.
In 1993 geochemistry research focused on improving the The NRC is sponsoring work by the CNWRA to investi-tools used for the analysis of IILW geochemical pro-gate contaminant transport in an unsaturated tuff at the 9-3 NUREG-1266
- 9. Ifigh Level Waste Research Nopal I site in the Pena liianca uranium district in Chi-dynamics and comparison with the data collected from huahua, Mexico.'lhe site is a tuff-hosted brecciated ura.
active sites.
nium ore body, which is analogous in many respects to the proposed repository at Yucca Mountain. This site is being Research on contemporaneous deformation rates in the studied to better understand the nature of contaminant Death Valley region, using global positioning satellite in-transport in a fractured, unsaturated tuff (i.e., the relative terferometry, was started in FY 1993. This investigation roles and interaction of the matrix and fractures in trans-will provide the NRC with an understandingof the region-port and the alteration of uraninite) and in an oxidizing al geologic forces that will affect volcanism, scismicity, environment. Detailed geologic, fmeture, and gamma and faulting in the Yucca Mountain area and will provide spectroscopy maps have been completed on the cleared, a basis for informed NRC assessment of the DOE's site exposed surface of the ore body. 'Iransport of uranium characterization data concerning geologic stability. At tends to be concentrated along iron-stained fract ures sup-present the possibility that measurable deformation was porting in general the findings from the ARAP project on occurring prior to and following the Little Scull Mountain the association of uranium sorption with iron hydroxide carthquake near Yucca Mountain is being investigated.
minerals. Migration is generally thought to be fracture controlled; however, samples collected both along frac-9.33 Performance Assessment Research tures and in the tuff perpendicular to the fractures indi-cate the relative mobility of uranium in the fractures ver-
'Ihe NRC will assess the claims of compliance made by the sus the matrix. Uranium series disequilibrium studies are IILW licensee, the DOE, with the NRC's quantitative being conducted to determiae the extent and nature of requirements for llLW disposal given in 10 CFR lbrt 60.
uranium mobility. The site also is being studied as an Included (by reference in 10 CFR 60.112)in thesc require.
analogue of spent fuel corrosion. Detailed mineralogic ments is the overall llLW repository performance stan-and petrologic studies of the primay uraninite indicate a dard,40 CFR Ibrt 191, set by the Environmental Protec-reasonable approximation of an oxidized spent fuel, but tion Agency. The development of a methodology to there are distinct differences in trace element concentra-quantitatively evaluate repository performance and the tion and grain size compared to unoxidized irradiated evaluation of the conceptual models used in the method-spent fuel.
ology are necessary to assess compliance.
The NRC is sponsoring research at the CNWRA to evalu-9.3.2.3 Gcology ate conceptual models used in the performance assess-ment of a repositoryin unsaturated, fractured tuff. In FY The N RC has an ongoing pnD'ect in volcanism in the Ilasin 1993, the CNWRA mvestigators completed development and Range to evaluate the potential for disruption of th of a computer program (PORFLOW, NUREG/CR-5991) repositog by igneous activity, this work at the CNWRA that is capable of simulating two-phase fluid flow condi-focuses on determining the extent and availability of vo!-
tions and radionuclide transport in geologic media. This canic, tecton e, and geophysical data from the region sur-program will be used to examine the representation of rounding Yucca Mountam.The NRC began a project on ground-water flow conditions during the thermal phase of Central liasin and Range tectonics at the CNWRA. An the repository and simplifications used to represent the extensive literature survey was completed for both volca-source term or releases from the engineered banier sys-nism and tectonics and a computenzed digital geologic tem' data base for the Yucca Mountain region has been suc-cessfully initiated. A review of age determination tech-niques for young basaltic volcanic rocks was also com-93.4 Regulation Development pleted this year.
A petition for rulemaking from the States of Washington and Oregon (PRM-60-4) was denied on March 4,1993 (58 The NRC began an additional volcanology project at the FR 12342). The petition requested the Commission to CNWRA to examine the disruptive scenarios of small-change the definition of high-level waste in its regulations volume basaltic volcanism in the Ilasin and Range.Three so that some of the radioactive waste materials being active vole:mic sites have been identified for analogue processed at the DOE IIanford site could be classified as studies of cruption dynamics, extent of disruption to the high-level waste. The petition was denied because the hydrogeology, and host rock mechanical stability based on existing NRC regulations on waste classification are well volcanic events. Ancient cindercone fields have also been established and can be applied on a case-by-case basis identified as areas for the study of magma emplacement without revising the definition of high-IcVel waste.
NUREG-1266 9-4
l l
l APPENDIX l
FY 1993 Regulatory Products from the Office of Nuclear Regulatory ReSearch Date Regulatory Product Description Part 1-NUCLEAR SAFETY RESEARCH-REACTOR LICENSING SUPPORT Reactor Aging and License Renewal Pressure Vessel Safety and Piping Integrity July 1993 Regulatory Guide Revision 29 to Regulatory Guide 1.84," Design and Fabrication Code Case Acceptability-ASME Section III, Division 1."
July 1993 Regulatory Guide Revision 29 to Regulatory Guide 1.85," Materials Code Case Acceptability-ASME Section III.
Division 1."
July 1993 Regulatory Guide Revision 10 to Regulatory Guide 1.147, " Inservice Inspection Code Case Acceptability-ASME Section XI, Division 1."
September 1993 Draft Regulatory Guide DG-1023," Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less
'Ihan 50 Ft-lb."
September 1993 Draft Regulatory Guide DG-1025, " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."
Standard Reactor Designs Regulatory Application of New Source Terms October 1992 Proposed Rule Revisions to Ibrt 100," Reactor Site Criteria."
November 1992 Proposed Regulatory Proposed Revision 2 to Regulatory Guide 4.7, Guide
" General Site Suitability Criteria for Nuclear Power Stations."
May 1993 lYoposed Rule The NRC regulation on em,ergency planning licensing requirements for independent fuel storage facilities.
May 1993 SECY-93-138
" Recommendations on Large-Release Definition."
June 1993 Proposed Rule The NRC regulation on emergency planning.
August 1993 SECY-93-232
" Staff Approach for Assessing the Effectiveness of the Present Regulations with Respect to the i
Commission's Safety Goals "
1 i
August 1993 SECY-93-226 Public comments on proposed rule on ALWR
)
severe accident performance.
l A-1 NUREG-1266
Appendix Date Regulatory Product Description Part 2-NUCLEAR SAFETY RESEARCII-REACTOR REGULATION SUPPORT Reactor Accident Analysis Severe Accident Policy Implementation October 1992 SECY-92-363
" Status of Implementation Plan for Closure of Severe Accident Issues, Status of the Individual Plant Examinations, and Status of Severe Accident Research."
April 1993 SECY-93-110
" Status of Implementation Plan for Closure of Severe Accident Issues, Status of the Individual Plant Examinations, and Status of Severc Accident Research."
May 1993 SECY-93-118
" Status of the Individual Plant Examination of External Events (IPEEE)."
Safety Issue Resolution and Regulation Improvements Earth Sciences October 1992 Proposed Rule Proposed Appendix II, "Scismic and Geologic Siting Criteria for Nuclear Power Plants," to 10 CFR Part 100 and proposed Appendix S,
" Earthquake Engineering Criteria for Nuclear Power Plants," to 10 CFR Part 50 issued for public com. ment.
November 1992 Draft Regulatory Guide DG-1015," Identification and Characterization of Seismic Sources, Deterministic Source Earthquakes, and Ground Motions."
November 1992 Draft Regulatory Guide DG-1016, Second Revision to Regulatory Guide 1.12," Nuclear Power Plant Instrumentation for Earthquakes."
l November 1992 Draft Regulatory Guide DG-1017," Pre-Earthquake Planning and immediate Nuclear Power Plant Post-Earthquake Actions."
November 1992 Draft Regulatory Guide DG-1018," Restart of a Nuclea Power Plant Shut Down by a Seismic Event."
November 1992 Standard Review Plan Section 2.5.2, Proposed Revision 3," Vibratory Ground Motion."
August 1993 Proposed Rule A proposed rule to amend 10 CFR 50.55a," Codes and Standards," to incorporate by reference the 1992 Edition with the 1992 Addenda of Subsection IWE, " Requirements for Class MC and Metallic Liners of Class CC Components of lj ht-Water g
NUREG-1266 A-2
I Appendix
)
Date Regulatory Product Description Cooled Power Plants," and Subsection IWL,
" Requirements for Class CC Concrete Components of Light-Water Cooled Power Plants," of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code.
Generic Safety issue Resolution i
October 1992-Generic Safety Issues For generic safety issues prioritized and resolved September 1992 in FY 1993, see 'Ihbles 6.1 and 6.2.
February 1993 SECY-93-004
" Resolution of GI-B56, Diesel Generator Reliability."
June 1993 Regulatory Guide Regulatory Guide 1.9, Revision 3, " Selection, Design, Qualification, and 'Ibsting of Emergency i
Diesel Generator Units Used as Class 1E Onsite Electric Power Systems at Nuclear Power Plants."
June 1993 NUREG-1364
" Regulatory Analysis for the Resolution of GI-106: Piping and the Use of Highly Combustib!c Gases in Vital Areas."
July 1993 NUREG-1463
" Regulatory Analysis for the Resolution of GI-105: Interfacing System Loss-of Coolant -
Accident in LWRs."
1 August 1993 NUREG-1461
" Regulatory Analysis for the Resolution of GI-153: Loss of Essential Service Water in LWRs."
Reactor Regulatory Standards February 1993 SECY-93-028 Staff plans for the climination of requirements marginal to safety.
April 1993 Final Rule The NRC regulations (10 CFR Ihrt 50) on training and qualification of nuclear power plant personnel. The rule amends the Commission's regulations to require each applicant and holder of a license to operate a nuclear power plant to establish, implement, and maintain programs for the training of nuclear power plant personnel that '
consider all modes of operation.
May 1993 Proposed Rule The NRC regulation,10 CFR lbrt 55, on requalification requirements for licensed operators for renewal of licenses. The proposed amendment would delete the requirement that each licensed operator pass a comprehensive requalification written examination and an operating test conducted by the NRC during the term of the operator's 6-year license as a prerequisite for license renewal.
A-3 NUREG-1266
i Appendix Date Regulatory Product Description
)
June 1993 Final Rule The NRC regulation,10 CFR 50.65, on monitoring the effectiveness of maintenance at nuclear power plants.The rule requires that the licensee conduct maintenance activities once every refueling cycle but not exceeding a period of 24 months.
June 1993 Regulatory Guide Regulatory Guide 1.160, " Monitoring the Effectiveness of Maintenance at Nuclear Power Plants."
June 1993 Regulatory Guide Regulatory Guide 8.38," Control of Access to High and Very High Radiation Areas in Nuclear Power Plants." This guide describes a framework of graded radiation protection procedures recommended to ensure that control for access to high and very high radiation areas are appropriate to the radiation hazard present in those areas.
July 1993 Regulatory Guide Revision 1 to Regulatory Guide 8.9, " Acceptable Concepts, Models, Equations, and Assumptions for a Dioassay Program." The guide describes practical methods acceptable to the NRC staff for estimating intake of radionuclides using bioassay measurement techniques.
July 1993 Regulatory Guide Regulatory Guide 8.37, "ALARA Levels for Effluents from Materials Facilities." This guide provides guidance on ALARA for materials licensees only.
July 1993 Volume 13 to NUREG-0713 A report on 1991 exposures, " Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities,1991." It provides a compilation of the statistical reports of individual exposures.
September 1993 Revision 2 to NUREG/HR-0058 A draft report on proposed regulatory analysis guidelines for public comment. The proposed guidelines represent the NRC's policy-setting document with respect to regulatory impact analyses.
September 1993 NUREG/BR-0184 A draft report," Regulatory Analysis'Ibchnical Evaluation Handbook." The purpose of the j
handbook is to provide guidance to regulatory analysts, to promote preparation of high-quality regulatory impact analyses, and to implement the policies of the guidelines.
A-4 NUREG-1266
1 Appendix Date Regulatory Product Description Part 3-NUCLEAR MATERIALS LICENSING AND REGULATION SUPPORT Nuclear Materials October 1992 Final Rule In an action related to the President's initiative to relieve the regulatory burden on nuclear licensecs, a final rule to amend the NRC regulations (10 CFR Part 35) on departures from manufacturer's instructions on recordkeeping requirements. This revision climinates recordkeeping requirements related to the justification for and a precise description of the departure and the number of departures from the Food and Drug Administration's approved manufacturer's instructions.
January 1993 Proposed Rule A proposed rule (10 CFR Parts 30,40,50,70, and
- 72) to allow self-guarantee as an additional mechanism for financial assurance. This proposed rule is in response to a petition for rulemaking (PRM-30-59) submitted by the General Electric Company and Westinghouse Electric Corporation.
The proposed rule would allow certain financially strong, non-electric utility licensees to use self-guarantee as financial assurance for decommissioning funding. It would not apply to electric utility licensees.
January 1993 Draft Regulatory Guide DG-3008, " Nuclear Criticality Safety 'Itaining "
This draft guide was developed to provide guidance to licensees on an appropriate nuclear criticality safety training program for the use of special nuclear material, especially the prevention of criticality accidents.
February 1993 Final Rule A final rule on licenses and radiation safety requirements for irradiators (58 FR 7715). The rule established a new Part 36 to specify radiation safety requirements and licensing requirements for the use of licensed radioactive materials in irradiators.
March 1993 Final Rule The rule to amend the NRC regulations (10 CFR 73.40(a) and 73.60) on physical protection requirements at fixed sites.The rule clarifies the Commission's regulatory intent that protection against both radiological sabotage and theft of special nuc! car material is not required at all facilities.
April 1993 Proposed Rule The proposed rule to amend the NRC regulation (10 CFR 72.214) adding one cask to the list of approved spent fuel storage casks. The proposed rule would increase the number of spent fuel A-5 NUREG-1266
l Appendix Date Regulatory Product Description storage casks from which holders of power reactor operating licenses can choose to store spent fuel j
under a general license.
April 1993 Draft Regulatory Guide
" Standard Format and Content for Fire Protection Sections of License Application for Fuel Cycle Facilities."
April 1993 Final Rule A rule to amend the NRC regulations (10 CFR 72.214) to add a cask to the list of approved spent fuel storage casks.The rule increases the number of spent fuel storage casks from which the holders of power reactor operating licenses can choose to store spent fuct under a generallicense.
June 1993 Final Rule The rule to amend NRC regulations (10 CFR Ibrts 26,70, and 73) on fitness for duty for Category I facilities and shipments of strategic special nuclear material (SSNM). The rule ensures that specific employees of licensees who possess, use, or transport SSNM do not have a drug or alcohol problem.
July 1993 Final Rule A rule (10 CFR lbrts 30 and 35) to extend the expiration date of the Interim Final Rule related to the preparation and therapeutic use of radiopharmaceuticals. This action allows licensees to continue to use byproduct material under the provisions of the Interim Final Rule until the NRC completes a related rulemaking to address broader issues for the medical use of byproduct material (including those issues addressed by the Interim Final Rule). It is expected that the broader rule will be issued as a final rule in FY 1994.
July 1993 Proposed Rule A proposed rule to amend NRC regulations (10 CFR lbrts 30,32, and 35) on the medical use of byproduct material.This action, taken in response to a petition for rulemaking (PRM-35-9), is intended to provide greater flexibility by allowing properly qualified nuclear pharmacists and authorized users who are physicians greater discretion to prepare radioactive drugs containing byproduct material for medical use.
August 1993 Final Rule The final rule to amend NRC regulations
( Appendix 11 to 10 CFR Ibrt 73) on day-firing qualifications for security personnel at Category I fuel cycle facilities. He amendment provides assurance that security force personnel maintain required weapon handling and marksmanship skills by annual performance testing for specific l
i I
NUREG-1266 A4 l
Appendix Date Regulatory Product Description security force personnel at facilities authorized to possess formula quantities of strategic special nuclear material.
September 1993 NUREG-1400
" Air Sampling in the Workplace." The report provides technical information on air sampling that will be useful for facilities fo!!owing the recommendations in the NRC's Regulatory Guide 8.25, llevision 1, " Air Sampling in the Workplace."
September 1993 Proposed Rule A proposed rule to amend NRC regulations (10 i
CFR Part 72) on rel,orting events at Independent Spent Fuel Storage installations and the Monitored Retrievable Storage installation.This rule would ensure that significant events such as contamination events, personal injuries, fires, and explosions at these facilities were promptly reported so that the Commission could evaluate whether the licensee has taken appropriate actions and whether prompt NRC action is necessary, i
September 1993 Proposed Rule A proposed rule to amend NRC regulations (10 CFR Ibrt 73) on physical fitness programs for security personnel at Category I fuel cycle facilities. 'Ihe amendment would require physical fitness training programs as well as annual performance testing for specific security force personnel at facilities authorized to possess formula quantities of strategic special nuclear material.
Low-Level Waste Disposal October 1992 Advance Notice of An advance notice of proposed rulemaking (10 Proposed Rulemaking CFR Part 40) concerning updating requirements for licensing of source material. "Ihe contemplated rulemaking would improve the control of source material through more specific regulation and update applicable requirements to conform with the revised standards for protection against radiation.
December 1992 Final Rule The NRC regulations (10 CFR lbrt 20) on disposal of waste oil by incineration at nuclear power plants. The rulemaking action, responding to a petition for rulemaking originally filed by Edison Electric Institute and the Utility Nuclear Waste Management Group (PRM-20-15), allows reactor licensees to pursue the option of incineration of waste oils contaminated with small amounts of radioactivity without the need for specific authorization.
A-7 NUREG-1266
Appendix Date Regulatory Product Description January 1993 Proposed Rule A proposed rulemaking (10 CFR lbrts 30,40,70, and 72) on timeliness in decommissioning a materials facility. The proposed rule would amend the Commission's regulations to establish 1
timeliness criteria for decommissioning nuclear sites or separate buildings or areas following permanent cessation of licensed activities.
June F)93 Final Rule A final rule to amend the NRC regulations (10 CFR Ibrt 61) to clarify that the requirements related to the pe'formance of land disposal facilities for low-level waste are applicable to aboveground disposal (i.e., built on the ground without an earthen cover).
July 1993 Final Rule A final rule (10 CFR Parts 30,40,70, and 72) to amend the NRC's decommissioning regulations to require holders of a specific license for possession of byproduct material, source material, special nuclear material, and independent storage of spent nuclear fuel and high-level waste to prepare and maintain additional documentation identifying areas where licensed materials and equipment were stored and used.
NUREG-1266 A-8
NRC FORM 3%
U S. NUCLEAR REGULATORY COMMISSION
- 1. REPORT NUMBER (2-89)
( Assigned by NRC, Add Vol,
t#tCM 1902 Supp., Rev., and Addendum Num-320'. 3202 BIBLIOGRAPHIC DATA SHEET b*- " *av-)
(See instrucuens on ine reversei NUREG-1266 Vol. 8
- 2. mte m) sueurLE
- 3. DATE REPORT PUBLISHLD NRC Safety Research in Support of Regulation - FY 1993 l
uenin vEAn June 1994
6, AulHORt6)
- 6. T YPE OF REPORT Regulatory L PERLOO COVERED (inclusive Dates)
FY 1993
- 8. PERFORMING ORGANIZATION - NAME AND ADORESS (if NRC, provide Division, Office or Region, U. S. Nuclear Regulatory Commission, and rnailing address; if contractor, provide name and mailirg address.)
Office of Nuclear Regulatory Research U.S. Nuc! car Regulatory CT 1 mission Washington, DC 20555-000.
- 9. 8PONSORING ORGANIZATION - NAME AND ADORESS (if NRC, type "Same as above*; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Comrnission, and mashng address.)
Same as 8, above.
- 10. SUPPLEMEN T ARY NO1 ES
- 11. ABSTRACT (200 words or 6ess)
This report, the ninth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. h summarizes the acccmplishments of the Office of Nuclear Regulatory Re-search during FY 1993. A special emphasis on occmolishrr.ents in nuclear power plant aging research reflects recog-nition that a number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects a focus on safety considerations for license renewal becomes timely.
The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with the technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission's rules, regulatory guides, or other guidance.
12 KEY WORDS/OcSCRIPTORS (List words or phrases that will assist researchers in locating the rersort.)
- 13. AVAILABILITY STAitiMENT l
Unlimited I
- 14. SECuRin cLASSNATION nuclear regulatory research I
regulatory products from research safety research Unclassified (This Report)
Unclassified
- 15. NUMBER OF PAGES
- 16. PRICE NRe FonM 335 (2-80)
l l
l l
i Printed on recycled Paper Federal Recycling Program
NUREG-1266, Vol. 8 NRC SAFETY RESEARCll IN SUPPORT OF REGULATION - FY 1993 JUNE 1994 l
UNITED STATES
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