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JUN 211984 Dr. Mary Shoaf Plasma Physics Lab.
P. O. Box 451 Princeton University Princeton, NJ 08544 Dear Dr.
oaf:
l,L4L44 Enclosed per your recent request to Bob Bernero are copies of 189's for FY 84 source-tem related research.
I have also enclosed a draft supplementary package which indexes the research projects to the major BMI-2104 codes, as well as related improved codes now under develoment.
This package was prepared following the Berkeley meeting in anticipation of the needs of the Study Group.
The supplementary package includes a brief sunnary of the objectives of each project and the most recent financial data for FY 84, 85, and 86.
The financial information for many of the 189's are out of date.
Some of the projects in the supplementary package do not have a 189 because they'are either new starts in FY 8A or international projects. Duplicate copies of all of this material will be available Thursday for the APS attenoees.
If you have any cuestions, please call me on FTS 427-4737 I look famed to seeing you on Thursday.
Sincerely,
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(b-fc. M 4 M. Silberberg, Assistant Dire.or for Research and Technical Support Accident-Source Tem Program Office Office of Nuclear Regulatory Research cc:
R. Bernerc 8507130167 050415
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SUMMARY
OF PLANS FOR SOURCE TERM RELATED RESEARCH
-ENCLOSURE 1 - Information on Computer Codes Related to the Source Tem Reassessment ENCLOSURE 2 - The NRC Analysis Program for Severe Accidents in LWRs ENCLOSURE 3 - Brief Description of Related Research Projects ENCLOSURE 4 - Project and Budget Proposal for NRC Source Tem Related Research Projects (Forms 189)
U.S. Nuclear Regulatory Comission Washington, D.C.
June 28, 1984
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ENCLOSURE 1 4
INFORMATION ON COMPUTER CODES RELATED TO THE SOURCE TERM REASSESSMENT Figure 1 identifies the Battelle suite of codes that was used for the source-tenn calculations in BMI-2'104. In the following paragraphs we will try to answer the question: What new codes might replace the Battelle%uite of codes and lead to improved results in the near future?
Wewill.alsofndicatehowour research dollars are being spent for the development and validation of these codes. Two approaches will be described below.
One approach is the development of the MELCOR risk-assessment code, which is less detailed, but faster running than the Battelle codes. Thus, the intent of the MELCOR code is not to provide detailed models of all thennal, physical, and chemical effects associated wf th care-melt accidents, but rather to provide an approximate method to estimate the timing and extent of fuel degradation, fission product and aerosol release, and containment failure for an overall risk assessment.
The other approach is the development or modification of a collection of state-of-the-art codes on separate effects.
These codes to the extent possible mechanistically treat the details of the thermal, physical, and chemical effects associated with fission product release and transfer from fuel, transport and deposition in reactor-vessel and primary-system components (piping, steam generator, etc.), and transport and deposition in the containment.
Some of these codes are too complex to be used for repeated calculations in a suite of codes; others, with some modification, could replace present components of the Battelle suite.
MELCOR is a risk-assessment code that includes thermal-hydraulics modules, fission product behavior modules, ex-plant consequence modules, and economic i
consequence modules. As such, MELCOR has a broader scope than the BMI-2104 suite of code', but it contains much less mechanistic detaiF. MELCOR is designed to ce fast running, which is necessary for the large numoer of runs i
required in a risk assessment study.
It is intended to replace the WASH-1400-generation codes, MARCH, CORRAL / MATADOR, and CRAC. While those codes had been
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" Source Tenn" Figure 1.
Battelle suite of codes as used in BMI-2104 9
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improved since the original WASH-1400 analyses, MELCOR will have improved, consistent treatments of severe accident phenomena and enhanced capabilities for sensitivity studies.
HELCOR will, in fact, employ slimed-down or pseudo versions of some of the state-of-the-art codes to be mentioned later, and it will be extensively bench-marked to those more mechanistic codes. The initial version of MELCOR is scheduled for completion in September 1984, and QA testing will be done in FY 85. A running version of MELCOR with a full complement of models is scheduled for September 1985. Benchmarking will start in FY 85 and continue through FY 86.
Table 1 lists state-of-the-art codes on separate effects th.t could replace components of the Battelle suite of codes. A number of codes (e.g., TRAPMELT, CORCON, VANESA) appear in Table 1 as both BMI-2104 codes and newer codes.
These codes are still under development or validation and improved versions can be expected.
Table 2 references the corresponding research projects for each of these newer codes. The research projects include code development and validation as well as experimental programs to provide a data base for the phenomenon in question.
A brief description of each of these projects along with funding levels for FY 84, FY 85 and FY 86 is enclosed. Also enclosed is a draft paper entitled, "The NRC Analysis Program for Severe Accidents in LWRs," that gives additional infonnation on our severe accident code strategy.
Linkages between some of the newer codes have been (or are being) developed.
For example, a TRAC /MIMAS/MELPROG 11nk has been developed to study the details of core melt progression. Helt progression and related phenomena have been shown by our source-tenn work to be very important in determining containment loads as well as fission product availability. A link between RELAP-5 and SCDAP, on the other hand, addresses core behavior up to the point of loss of fuel-rod geometry and should be more valuable for studying terminated accidents and mitigation features.
One could certainly conceive of a linkage between TRAC, MIMAS, MELPROG, VICTORIA, TRAPNELT, CORCON, VANESA, and CONTAIN. Such a linkage would e
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Table 1.
Newer codes that have the potential for upgrading components of the Battelle suite of codes used in BMI-2104, Description BMI-2104 Newer Code a
RCS Thermal-Hydraulics MARCH, MERGE TRAC, RELAP-5 Fuel Heatup and Degradation MARCH SCDAP, MIMAS,
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- MELRPI, MELPROG Fission Product Release CORSOR
- GRASS, from Fuel (in vessel)
VICTORIA RCS Fission Product Transport TRAPMELT TRAPMELT Molten Fuel Interaction MARCH WISCI with Coolant Debris-Concrete Interactions CORCON CORCON Fission Product Release from VANESA VANESA Core-Concrete Melt (ex-vessel)
Containment Thermal-Hydraulics MARCH CONTAIN Hydrogen Behavior MARCH CONTAIN (HECTR)*
Containment Fission Product NAUA, SPARC, NAUA, CONTAIN Transport ICEDF (SPARC,
- ICEDF, MAEROS)*
Subroutines in CONTAIN
- Table 2.
Corresponding Research Project to Develop or Upgrade Newer Codes.
Newer Code Description Research Project Number TRAC, RELAP-5 RCS Thermal-Hyrdaulics 3, 4, 7. 10, 1
14, 26, 31 SCDAP, MIMAS, Fuel Heatup and 1,2,3,4,5,6,8, MELRPI, MELPROG Degradation 9, 10, 12, 13, 14, 15, 16, 17, 18, 19, 26, 28 GRASS Fission Product Release 4, 5, 9, 10, 11, 12, i
VICTORIA from Fuel (in vessel) 13, 14, 15, 16, 18, 19, 20, 22, 26 TRAP-MELT RCS Fission Product 5, 13, 14, 15, 18, 20 Transport 21, 22, 26, 30, 31, 35 CORCON Debris - Concrete 7, 25, 27, 28 Interactions
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VANESA Fission Product Release 5, 20, 26, 27, 28 from Core-Concrete Melt (ex-vessel)
CONTAIN(HECTR)
Hydrogen Behavior 4, 16, 22, 26, 29, 34 NAUA, CONTAIN Containment Fission 16, 20, 22, 25, 26, (SPARC,ICEDF, Product / Aerosols Transport 29, 31, 33, 34 MAER05,QUICXM)
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constitute a complete substitute for the Battelle suite of codes and could be used for future detailed source term analysis. At this time, however, such a
. linkage is not planned. Instead, the newer state-of-the-art codes will be used individually or in small groups to understand in some detail the important
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physical phenomena affecting severe accidents. These codes will in turn be used to benchmark the fast-running MELCOR code, which will be used for the large number of code runs required for risk assessment.
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4
ENCLOSURE 2 THE NRC ANALYSIS PROGRAM FOR SEVERE ACCIDENTS IN LWRs a
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l THE NRC ANALYSIS PROGRAM FOR SEVERE ACCIDENTS IN LWR'S i
George P. Marino I.
Introduction The analysis of severe accident consequences for light water reactors has been a project of major importance in the NRC prior to and, even more vigorously. -
after the event at Three-Mile Island Unit 2.
In the development of a 1
methodology for ascertaining the consequences of such events for use in risk l
analysis and source term studies, one must appreciate beforehand that the nature and complexity of the phenomena involved limit the extent to which an exact analysis capability is possible. An accurate analysis capability would require a set of experimentally validated models capable of treating in great detail all the phenomena occurring in the multi-phase, multi-component, nuclear steam supply system (NSSS) over temperature ranges from 300C to 2800C, pressure ranges from 15 psi to 2350 psi, and timei periods over days, weeks, and possibly months. As a practical matter, the capabilities of an analysis technique of this type must be limited since:
1.
Fully integral tests to validate all the models under all possible 1
conditions for all types of plants would be prohibitively expensive, take decades to accomplish, and be difficult to evaluate.
y 2.
Risk analyses and the necessary sensitivity and uncertainty analyses
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which are part of them require fast-running analysis codes so 1
i that many sequences for many plants can be made in a reasonable time frame. Therefore, fast running computer codes require extensive model simplification to achieve this goal.
3.
An exact quantification of the technology of all the processes expected to occur is not possible since uncertainties will always exist in experimental data for material properties, physical and chemical l
parameters, the models themselves, and the sequences predicted to occur.
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06/26/84 1
SEVERE ACCIDENTS i
Therefore, in order to make the goal of quantifying the consequences of severe accidents tractable, one must approach the problem in a more realistic manner.
That is, recognizing that (1) uncertainties will always exist, (2) complete experimental validation is not feasible, and (3) analysis codes for risk studies must be fairly fast (i.e., less than a few hours on a CRAY for a given secuence and plant). a methodology must be developed that will result in a fast-running analyst.s code that evaluates the entire nuclear steam supply syri,em and.is composed of phenomenological models whose uncertainty can be quantified. Such a' goal requires the accomplishment of two major prerequisite tasks; namely:
(1) The establishment of a data bas'o to the extent practically possible in the temperature ~and pressure ranges of interest and,
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(2) The development of specialized detailed rnalysis codes for specific components of the NSSS such as the primary system T/H, the core, and the containment. These codes should be validated using<iten(1)above.
g Given the above two elements, the final goal of a fast-running overall NSSS code containing ouantified uncertainties in its output can be achieved by benchmarking its simplified mLdels against the corresponding models in the more mechanistic analysis codee.- The uncertainty in the latter codes will be quantified -- to the extent,)ossible -- by the established data base.
It is to be expected that the unceitainty of a given model in the overall code will be creater than or equal to thatAa the more mechanistic code because of the need fa'r fast-running times. However, whatever the magnitude of the uncertainty,_it will be quantified and the results can therefore be used in deedsfon making.
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06/26/84 2
SEVERE ACCIDENTS i
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II. Current Procram of the NRC In order to achieve its goal of developing a fast-running NSSS analysis code with quantifiable uncertainties for analyzing the consequences of severe accidents in LWR's, the Office of Research established the Severe Accident Research Program (SARP) as part of its long-range plan shortly after the TMI-2 event. This program consists of approximately seventy research contracts to develop data bases for core behavior; primary system behavior; containment behavior; fission prcduct release, transport, deposition, and re-evolution; hydrogen behavior; and detailed analysis codes for all of the above including severe accident sequence analysis (SASA); detailed risk analysis; and risk reduction studies. The current program utilizes approximately one-third of the annual Research budget, and is scheduled for completion in the 1986/87 time frame. The risk analysis code mentioned above (f.e., the fast-running NSSS computer code with quantified uncertainties) is under developr,
and has been given the name MELCOR. The initial, unvalidated version, is seneduled for completion by the end of FY 1984 The development of the supporting mechanistic analysis codes for benchmarking and quantifying MELCOR has been undenvay since 1980. This set of codes consists of a few newly developed codes, previously developed codes (such as TRAC and RELAPS), and linked packages of new and previously-developed codes.
An extensive effort has been made to utilize previously developed analysis packages to avoid "re-inventing the wheel" and, of course, to minimize expenditures.
In order to clearly show how all these codes (there are 25 of them) fit into the general scheme outlined in the Introduction, it will be necessary to graphically illustrate their function in relation to the NSSS.
In order to do' this we must first separate those codes intended for detailed analysis for benchmarking HELCOR from the current " risk" codes that MELCOR is intended to replace.
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SEVERE ACCIDENTS
4 A.
Current NRC Risk and Source-Term Analysis Codes This group of codes is currently being used in the current NRC evaluation of the source term for selected plants and sequences. Earlier versions of some of the.models were used in the WASH-1400 study.
It must be remembered that these codes are not " detailed" in the sense mentioned above; i.e., they are fast-running codes which necessarily requires
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considerable use of simplifying assumptions, empirically-based correlations, 1
and user input options. They also have not had the benefit of an adequate cata base from which to validate and assess their simplified models.
However, they represent our current best integrated anal'ysis capability for i
source term analyses until the_ SARP program is completed and MELCOR has been fully validated and quantified. Table I gives a list of the codes and their-application to the NSSS. Note that this family of codes is commonly referred to today as the "Battelle Suite of Codes" since the Battelle-Columbus laboratory has been prime contractor for their application.
It should be noted that most of the models in this code series ~will be used directly in MELCOR with only slight changes. These models are ORIGEN, TRAP / MELT, VANESA, SPARC, CORCON, and ICEDF. Major improvements are expected for the MARCH, MERGE, CORSOR,. and containment aerosol applications. MELCOR will, however, contain additional models for containment temperature and pressure response as well as ex-plant consequence models.
I B. " Mechanistic Specialized Codes for Benchmarking MELCOR and Special Applications This group of codes represent best-effort modeling with little emphasis on speed, b'ut great emphasis on model accuracy. In other words, these codes are intended to represent the best state of technology for specialized phenomena and for applications to specific areas of interest. These codes, when completed, will be maintained as a best-estirate base of modeling expertise to be used when the necessarily simplified models in MELCOR are judged inadequate for highly specialized applications. The general philosophy being applied here is that the NRC staff must maintain " state of knowledge" expertise to be able to do in-depth studies of important phenomena whenever the need arises. However, as stated above, the primary application of this code group is to benchmark and 06/26/84
- 4 SEVERE ACCIDENTS
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quantify the modeling uncertainties that will inherently be present in MELCOR, and, incidentally, to the Battelle Suite of Codes. Table II presents a summary of this group and where they are applied in the NSSS. With regard to the availabilit;y dates given in Table II, it should be understood that these codes will be continually updated and improved beyond that date as new experimental data for validation become available.
C.
Linkages of Mechanistic Codes for'Special Applications Where " Feedback" is Important There are many situations where the input period from one specialized code is affected by the output of the receiving code over a given tiqle period.
For such cases, more accurate modeling is accomplished by " linking" the codes together so that input / output information can be interchanged during' the run. An example of this is a TRAP / MELT calculation which relocates a significant part of the decay heat source into the upper plenum. This re-located heat source will affect the MERGE input on flows and structures in the upper plenum and other parts of the primary system and, therefore, future TRAP / MELT computations. Therefore, it is essential to develop hard links between the mechanistic fuel behavior modules (SCDAP, MELPROG) and system T/H codes (RELAP, TRAC) because of the intimate coupling between fuel' degradation and the T/H behavior of the reactor coolant system. Note that MELCOR will be a fully integrated code package with inter-model feedback included as an inherent part of the programing.
For the detailed code series a few major code links are planned and are currently being implemented. They are shown in Table III. Completion dates for these linkages have not been fimly established, but they are expected to be available by mid 1985. The major reason for the SCDAP/RELAP5 link in addition to the TRAC /MIMAS/MELPROG link is that detailed modeling of the fuel pins is necessary for attentuated accidents but may not be necessary for " core on the floor" accidents. Therefore, using the less detailed MIMAS code in place of SCDAP will increase computational speed dramatically for non-attenuated accidents scenarios. Moreover, the SCDAP/RELAPS link will give early capability to analyze both PWR and BWR systems for these less drastic events.
It should be noted that the above 06/26/84 5
SEVERE ACCIDENTS
procedure is very cost-effective since the system codes have already been developed and validated as part of the ECCS research program, and the linkages can be accomplished in a very short time to give essentially universal appli.cability. That is, capability for PWR's and BWR's for
" core on the floor events" and for attenuated events such as the accident in TMI-2. Finally, the linked codes can be used to benchmark a wide range of MELCOR's integrated package and quantify the effects of feedback in severe accident analyses.
III. Summary The code development plan outlined above will provide the NRC with the analysis capability required for decision-making on severe accidents in LWR's. An integrated risk code is provided - MELCOR - for large-scale PRA and source term studies as well as special-application, more mechanistic codes for less broad, more specific decisien-making processes. The plan utilizes to the broadest extent possible the codes developed for other purposes as well as the extensive new data base being developed under SARP for NSSS behavior under severe accident conditions.
Table IV summarizes how all these codes are used in NSSS analyses by classifying them by NSSS component application and by phenomenological categories. Finally, Figure 1 summarizes the codes graphically to s)ow how they will " fit" together to accomplish their intended purpose. Note'that some of the current "Battelle Suite" of codes will be used directly in the mechanistic set and will not be "re-invented". Note also that the CONTAIN code consists of many models from current codes used in a subroutine capacity such as CORCON, HECTR, etc.
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SEVERE ACCIDENTS i
TABLE I BATTELLE SUITE OF CODES FOR RISK AND SOURCE-TERM STUDIES NAME OF CODE APPLICATION IN THE NUCLEAR STEAM SUPPLY SYSTEM ORIGEN MODELSFISSION{RODUCTINVENTORYINTHECOREPRIOR TO SCRAM
-i MARCH 2.0 PRIMARY SYSTEM T/H CORE PHENOMENA (CONTAINMENT T&P)
MERGE IN-VESSEL GAS FLOW AND HEAT TRANSFER TO STRUCTURES.
USED AS AN INTERFACE BETWEEN MARCH & TRAP / MELT TRAP / MELT MODELS FISSION PRODUCT AND AEROSOL TRANSPORT AND DEPOSITION WITHIN THE REACTOR COOLANT SYSTEM CORSOR MODELS FISSION PRODUCT AND AEROSOL RELEASE FROM THE CORE. AN EMPIRICAL CODE BASED UPON EX-PILE, FISSION-PRODUCT RELEASE EXPTS CORCON MODELS EX-VESSEL MOLTEN CORE INTERACTION WITH REACTOR CAVITY BASEMAT MATERIAL VANESA MODELS FISSION PRODUCT AND AEROSOL RELEASE DURING MOLTEN CORE /BASEMAT INTERACTION NAUA-4 MODELS AEROSOL BEHAVIGR IN THE CONTAINMENT SPARC MODELS AEROSOL RETENTION IN SUPPRESSION POOLS (BWRONLY)
ICEDF MODELS AEROSOL RETENTION IN PWR ICE-CONDENSER CONTAINMENT SYSTEMS 06/26/84 7
SEVERE ACCIDENTS
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_ TABLE II DETAILED NRC SEVERE ACCIDENT COMPUTER CODES TO BENCHMARK AND OUANTIFY MELCOR NAME OF CODE APPLICATION IN THE NUCLEAR STEAM SUPPLY SYSTEM AVAILABILITY DATE SCDAP DETAILED CORE BEHAVIOR TO LOSS OF R00 GEOMETRY, OCTOBER 1984 I.E., TO ABOUT 2400K MELPROG DETAILED CORE BEHAVIOR THRbOGM MELTDOWN AND EXIT DECEMBER 1984 THE REACTOR PRESSURE VESSEL
),
MIMAS SIMILAR TO SCDAP-BUT MUCH LESS DETAILED--USED COMPLETE AS INPUT TO MELPROG FOR UNATTENUATED " CORE 2-0 VERSION IN ON THE FLOOR" EVENTS DECEMBER 1984 FASTGRASS DETAILED FISSION PRODUCT RELEASE FROM INTACT OCTOBER 1984 FUEL USED IN SCDAP & MELPROG TRAP / MELT SEE TABLE I - THESE MODELS WILL BE MODIFIED &
COMPLETED USED IN SCDAP, MIMAS, AND MELPROG VICTORIA DETAILED FISSION PRODUCT AND AEROSOL RELEASEOCTOBER 1985 FRCM MOLTEN FUEL FOR USE IN MELPROG CONTAIN INTEGRATED DETAILED CONTAINMENT MODEL:
LE.IT JUNE 1984 USES SUBM00ELS FOR T/H, AEROSOL AND FISSION PRODUCTS (MAER05), CAVITY MODELS (CORCON, MEDICI, VANESA), HYOROGEN SURNING (HECTR) &
ESF MODELS HECTR SEE CONTAIN CODE ABOVE. HECTR MODELS HYDROGEN DECEPSER 1984 BEHAVIOR IN CONTAINMENT.
TREATS OEFLAGRATIONS, SPRAYS, HEAT TRANSFER, IGNITERS, SUPPRESSION POOLS, SUMPS, AND FANS e
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TABLE III DETAILED NRC CODE LINKS FOR SEVERE ACCIDENT ANALYSES
- CODES LINKED APPLICATION TO NSSS SCDAP/RELAPS INTEGRATED PRIMARY SYSTEM / PARTIALLY DEGRADED CORE BEHAVIOR FOR ATTENUATED ACCIDENTS SUCH AS TMI-2 TRAC / MIMA' /MELPROG S
INTEGRATED PRIMARY SYSTEM /MOLTE" CORE BEHAVIOR FOR " CORE ON THE FLOOF~ EVENTS.
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TABLE IV
SUMMARY
OF NRC SEVERE ACCIDENTS COMPUTER CODES (MELCOR MODELS ALL)
NSSS COMPONENT RISK CODES MECHANISTIC CODES PRIMARY SYSTEM MARCH, MERGE TRAC, RELAPS, TRAP / MELT TRAP / MELT CORE MARCH, CORSOR SCDAP, MELPROG, MIMAS, FASTGRASS, VICTORIA, TRAP / MELT REACTOR SUMP CORCON, VANESA CORCON, VANESA, MEDICI CONTAINMENT NAUA-4, SPARC, CONTAIN (INCLUDES SUMP ICEDF, MARCH CODES)
BY PHENOMEN0 LOGICAL CATEGORY T/H CODES:
MARCH, TRAC, MERGE, RELAP5 FISSION PRODUCT RELEASE CODES: CORSOR, FASTGRASS, VICTORIA, VANESA FISSION PRODUCT TRANSPORT AND DEPOSITION CODES: TRAP / MELT AEROSOL BEHAVIOR CODES: NAUA-4, MAEROS (IN CONTAIN)
REACTOR CAVITY MODELS: CORCON, MEDICI HYDROGEN BEHAVIOR CODES: HECTR (IN CONTAIN)
ESF CODES: SPARC, ICEDF (BOTH IN CONTAIN) 06/26/84 8
SEVERE ACCIDENTS
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MELCDR RIPt ACf 5 RINCHMAREfD_ RT I
l l Current Suite of Source Term Codes Detailed Code Series l
MARCH TRAC /MIMAS REL AP 5
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ENCLOSURE 3 BRIEF DESCRIPTION OF RELATED RESEARCH PROJECTS e
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Project 1 (FIN NO. B2455)
TITLE:
SEVERE CORE DAMAGE MATERIALS PROPERTY TEST OBJECTIVES: The primary purpose of this program is to generate data for use in modeling the behavior of fuel exposed to high tem-perature transients. A qualitative assessment of the so-called runaway Zr-0, reaction will also be made. Such an assessment could be important to understanding the severity and consequences of fuel damage where the maximum heat input is expected from reaction with steam.
Program tests are intended to provide the following data:
(a) Weight gain due to oxygen uptake as a function of temp.
(b) Rate constants for Zr/H 0 reactions as a function of temp.
(c) H /H O content as a function of time for a given temp 2
(d) H /H 0 content as a function of time during transient 2
oxidation (e) Uncontrolled temperature rise as a function of initial heating rate (f) Effect of H,/H O on K 2
p (g) Effect of U0 on.Zr/H O reaction rates 2
2 (h) Viscosity of Zr-UO as a function of temp (1) Ha production from Zr-UO up to 1700*C (j) He production from Zr-UO from 1700'C - 2000*C BUDGET ($K):
FY 84 FY 85 FY 86 240 300 3G0
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Project 2 (FIN NO. B2277)
TITLE:
NRU COOLANT BOILAWAY AND DAMAGE PROGRESSION (CBDP) TESTS OBJECTIVES: To provide well characterized full-length in-reactor test data on the severe fuel damage behavior of prototypic fuel rod clusters during and after coolant boilaway. In-reactor severe fuel damage tests on full-length prototypic fuel rod clusters are needed to validate oxidation and other models based on data from separate effects or short-length in-reactor tests.
The NRU CBOP tests will:
s (a) evaluate length effects on damage progression (b) determine the relationship of other SFD programs to NRU prototypic experimental test conditions measured from full length tests, including temperatures, temperature gradients, steaming rates, hydrogen generation, damage progression (c) evaluate instrumentation performance under prototypic environments and temperature gradients.
The three tests in the current NRU SFD program are as follows:
NT-6, FY 84, 21-rod clad ballooning and low-temperature oxidation test; FLHT-1, FY 85,12-rod SFD rapid oxidation test to 2150; FLHT-2, FY 86,12-rod SFD rapid oxidation test to 2500K.
BUDGET (SK):
FY 84 FY 85 FY 86 2466 1450 2250 (NRC funds only) 4
Project 3 (FIN NO. A7303)
TITLE:
TRAC /MELPROG INTEGRATION OBJECTIVES: To develop two-dimensional thermal-hydraulic degraded-core models and a subcode to be used as a module in MELPROG under development at SNL. This project is an extension of the MIMAS code (developed at LANL for DOE) in which one-dimension thermal-hydraulics was used for degraded-core modeling. As part of the project, TRAC-PWR is being used to calculate flow recirculation in a PWR vessel; ifnking TRAC-PWR with MELPROG will be explored in the future to provide analysis capabilities for the reactor coolant system.
BUDGET (SK): FY 84 FY 85 FY 86 328 330 350 h
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Project 4 (FIN NO. A1389)
TITLE:
ACRR SOURCE TERM EXPERIMENTS - PLANNING OBJECTIVES: Plan for a series of experiments in ACRR on fission-product release from reactor fuel, the chemical form, and aerosol formation under in-core severe-accident conditions at temper-atures up to fuel melting. The purpose of these experiments is to provide a data base for the development of fission-product release-r ^ models for the conditions of in-vessel severe fuel damage and care-melt progression. This planning is to include analysis of experiment requirements and diagnostics capabilities and engineering analysis of the pro-posed experiments through the preliminary design phase.
Explicit attention will be given to obtain results for the following phenomena: release from solid fuel, release during fuel liquefaction, and release following fuel slumping. A report on this planning is to be submitted to NRC to form a basis for an NRC decision on whether to carry out these experiments as a major part of the alternative program to the PBF Phase II tests. Forseen is a possible program of ten experiments over a period of three years that includes development time.
BUDGET ($K):
FY 84 FY 85 FY 86 300 920 2000 i
e
% de e
3 4
i m,
project 5 (FIN NO. A1227)
TITLE:
HIGH TEMP FISSION PRODUCT CHEMISTRY OBJECTIVES: To investigate the high temperature chemistry of fission products in the vapor phase 'following' release from the fuel. To determine the chemical interaction of fission products with RCS surface 4
materials. To provide data for development of fission product transport models.
This project focuses on the kinetics of the chemical retention processes for fission products in the primary system during severe accidents. (Physical retention processes are being studied elsewhere). Research in FY 85 and 86 will include the study on the effects of radiation and pressure on the reaction kinetics.
BUDGET ($K):
FY 84*
FY 85 FY 86 980 400 400
- The FY 84 funding includes 600K for the QUEST-I study.
5 5
Pro. ject 6 (FIN NO. A1335)
TITLE:
LWR DEBRIS FORMATION & 4 LOCATION, MELT PROGRESSION, AND FISSION PRODUCT RELEASE OBJECTIVES: To develop a data base and verified analytical models on the formation and relocation of LWR severely damaged fuel (core debris) and fission-product release applicable over the range of the risk-significant severe accident sequences. Separat'e-effects experiments up to full fuel-melt temperatures (3100K) are to be perfonned,in the ACRR test reactor to give continuous-time data on the development of severe fuel damage, on debris fonnation by quenching at different times, on hydrogen generation, on debris relocation processes, and on fission-product release.
Phenomenological models of the governing processes involved in the development of core-debris relocation, core-melt progression and in-vessel fission-product release are to be developed from the results of these experiments, the severe fuel damage tests in PBF, other in-pile tests, and laboratory experiments in both U.S. and foreign countries.
These experiments all use cinematography for time-continuous measurements of surface temperatures and of the damage-progression processes. The nine-experiment program includes four experiments on debris formation and relocation under core-uncovery conditions, two experiments on debris formation under reflood-quench conditions, and three experiments with pre-irradiated fuel on melt progression and fission-product release under in-vessel melt-progression conditions. Two experiments will be performed in FY 84, four in FY 85, and the final three experiments with pre-irradiated fuel in FY 86. This program is part of the integrated severe fuel damage research program of the Fuel Systems Research Branch.
BUDGET ($K):
FY 84 FY 85 FY 86 2000 1500 2000 i
6
.. -, _ _ - - -, - -. ~. -
m Project 7 (FIN NO. A1340)
TITLE:
LWR CORE DEBRIS COOLABILITY
~
OBJECTIVES: To develop a data base on the dry-out coolability limits by reflooding and the post-dry-out behavior of LWR core debris by performing a series of experiments in the ACRR test reactor.
To develop and verify relevant phenomenological debris-coolability models and codes for use in safety assessment. Emphasis is to be on verification of the existing LMFBR debris-bed coolability models for LWR specific conditions, in particular, high pressure, deep beds, coarse LWR debris, and inlet coolant flow as well as for the different coolant properties. All three experiments in this program utilize fission heating (simulating decay heating) of the artificial debris beds in ACRR, relatively deep (50cm) debris beds, and cover the full pressure range from 1 to 170 bars (2500 psi). The first two experiments have relatively fine and relatively coarse LWR debris operating, at dry out, in the laminar and the turbulent vapor-flow regiems, and will be performed in FY 84 The third experiment with a vertically stratified bed and variable inlet flow will be performed in FY 85. This program is part of the in'tegrated severe fuel damage research program of the Fuel Systems Research Branch..
BUDGET ($K):
FY 84 FY 85 FY 86 500 0
0 t
8 7
d Profect 8 (FIN NO. A1342)
TITLE:
MELT PROGRESSION ANALYSIS (MELPROG)
OBJECTIVES: To develop analytical models and a computer code (MELPROG) to simulate LWR in-vessel melt progression from the onset 9
severe core damage to core debris interaction with the structures in the vessel lower plenum and to reactor vessel
-5 failure. As part of the project, a subcode (VICTORIA) will' be developed to model the fission product release from care debris; VICTORIA will be linked with TRAP-MELT as a module to be used in MELPROG to handle fission product release and transport in the vessel. MELPROG will use the two-dimensional thermal-hydraulic models under development at LANL and will also use some subroutines adopted from SCDAP for degraded core behavior. MELPROG will calculate the timing and release rates of the fission products and core debris from a failed vessel to the containment.
BUDGET ($K):
FY 84 FY 85 FY 86 590 650 800 s
8
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Project 9 (FIN NO. A2016)
TITLE:
TRANSIENT FUEL RESEARCH & FP RELEASE OBJECTIVES: To develop physically realistic models which describe fission product release from LWR fuels during thermal transients.
This project centers on the development of a mechanistic fission product release code (and a fast running version) for the early phase of a severe accident. Work plans include the development of models to describe the effect of enhanced fission product release by fuel oxidation and the incorporation of the semi-volatile fission products such as Te, Ba, Sr, etc.. into the code.
BUDGET.(SK):
FY 84 FY.85 FY 86 314 300 250 9
m
--_y
Project 10 (FIN NO. A2220)
TITLE:
TMI FUEL EXAMINATION OBJECTIVES: Conduct macro, microstructural, and chemical analyses of selected specimens removed from the core of the TMI-2 reactor pressure vessel to determine thermal and chemical changes that occurred.
With respect to the utility of TMI-2 examinations, there are a number of phenomena and for which TMI-2 data could potentially be of significant benefit. This includes the unknown degree of aerosolization of control rod material, and the chemical behavior of tellurium in the presence of metallic zircaloy.
The principal TMI-2 core examinations at ANL will be carried out under DOE auspices at several facilities. The larger portions of the core will be first sent to EG&G hot-cells for dissemination of smaller units to satellite facilities for in-depth examinations.
Selected TMI-2 core debris specimens will be analyzed using optical and electron microscopy, X-ray, and electron microprobe ch'emical analyses, auger and scanning electron microscopy, to characterize the changes that occurred in the TMI-2 accident.
Following the beginning of TMI-2 defueling operations ANL will continue analyses an' components including fuel and control material pieces, fuel assembly structural pieces, and core debris from primary system locations.
BUDGET ($K):
FY 84 FY 85 FY 86 190 270 500 9
10 O
Project 11 (FIN NO. A2232)
TITLE:
POSTTEST FUEL EXAMINATION OBJECTIVES: To conduct post-test examinations (PIE) of the concercially irradiated fuel rods specimens subjected to high temperature fission product release testing at ORNL. To provide information on fission product release mechanisms for use in developing.
detailed models of fission product release from fuel under severe core damage and care melt accidents.
To accomplish the objective of determining the fission product release mechanisms from the ORNL tests, the investigator will look for diffusion along grain boundaries, interlinkage of voids, grain boundary sweeping, microcracking, etc., and compare to pretest fuel morphology. Post test fuel examination will also include the determination of the spatial distribution of the oxygen concentrations within the fuel and the locations of fission products remaining within the fuel pellet structure.
BUDGET ($K):
FY 84 FY 85 FY 86 150 150 150 11
Project 12 (FIN NO. A6050)
TITLE:
FUEL BEHAVIOR MODEL DEVELOPMENT
.0BJECTIVES: To improve analytical models used in the FRAP-T6 code and in th4 MATPRO code. FRAP-T6 is used to simulate fuel rod behavior including clad ballooning, rupture, and fission product release during reactor transients and design-basis accidents. MATPRO is used to provide material properties input to FRAP-T6 and also to SCDAP which is for severe accident simulations. As part of the project, INEL will maintain and update both FRAP-T6 and MATPRO as new data become available.
BUDGET ($K):
FY 84 FY 85 FY 86 380 200 0
l l
12
Project 13 (FIN No. A6044 and A6057)
TITLE:
PBF ENGINEERING & PLANT OPERATIONS OBJECTIVES: Provide operating crew for and engineering support of the Powr Burst Facility.
BUDGET ($K):
FY 84 FY 85 FY 86 8000 5100 0
13 O
i l
l i
l Pro.fect 14 (FIN NO. A6305) l TITLE:
PBF SEVERE FUEL DAMAGE EXPERIMENTAL PROGRAM l
OBJECTIVES: To provide a data base and analytical models on fuel behavior under the core-uncovery and reflood quench conditions of severe LWR accidents by performing integral multi-rod fuel-bundle tests in the P8F test reactor, by analysis of the test results, and by model development. Fission-product release and transport, hydrogen generation, and temperature distributions during the test transients are to be measured, and post-test character-iration of the test fuel debris, including fuel relocation is to be made by neutron radiography and tomography and by post-l irradiation examination (PIE). This program is part of the l
integrated seve,re fuel damage research program of the Fuel Systems Research Branch, c
l The program includes four Severe-Fuel-Damage (SFD) tests in the PBF test reactor, analysis of results, and Post-Irradiation Examination (PIE) of the damaged 32-rod fuel
{
bundles. Test SFD 1-1, performed in 1983, used preconditiened (at power) fresh fuel as did the scoping test. Test SFD 1-3, to be performed in 1984, will use preconditioned pre-irradiated fuel as will the similar early 1985 test SFD 1-4 that will include As-In-Cd control rods. Irradiated-fuel tests 1-3 and 1-4 will have substantially improved fission-product diagnostics.
I BUDGET ($K):
FY 84 FY 85 FY 86*
i
)
4200 2600 2000 i
{
l 14
_u._._
Project 15 (FIN NO. A6321)
TITLE:
IN-PILE FISSION PRODUCT STUDIES AT PBF OBJECTIVES: To measu a fission product behavior during the PBF Severe Fuel Damage tests (see Project 14). To design systems to obtain fission product release data. To collect, reduce, analyze, and report fission product data.
BUDGET ($K):
FY 84 FY 85 FY 86 3200 2500 1000 9
e i
e 15 ww -
Project 16 (FIN NO. A6352)
TITLE: RESIDENT SCIENTIST AT KFK KARLSRUHE, FRG OBJECTIVES: To facilitate the exchange of nuclear safety-related infor-mation between the U.S. and Germany.
BUDGET ($K);
FY 84 FY 85 FY 86 175 140 200 4
l 4
i 1
I j'
d I
l i
1 16
Project 17 (FIN NO. A6360)
TITLE:
SEVERE FUEL DAMAGE MODEL DEVELOPMENT (SCDAP)
OJ8ECTIVES: To develop analytical models and a computer code (SCDAP) to simulate severe fuel damage and degraded core accidents in LWR's. SCDAP models fuel / clad melting and relocation, blockage formation, fission product release, and hydrogen generation.
It has been used to simulate the TMI-2 accident and P8F severe fuel damage experiments. As part of the project, SCDAP will be linked with TRAP-MELT and RELAP5/M002 to provide system-wide modeling capabilities to calculate the timing and release rates of the fission products and hydrogen from a degraded core through a pathway to the containment during severe accidents in which the reactor vessel. remains intact. Close cooperation will be mein-tained between SCDAP developers and MELPROG developers.
BUDGET ($K):
FY 84 FY 85 FY 86 712 700 800 i
9 f
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0 17
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.s Proftet 18 (FIN NO. A6829)
TITLE:
MODIFICATIONS TO FISSION PRODUCT MEASUREMENT SYSTEMS FOR SFD SERIES-I TESTS OBJECTIVES: Upgrade fission prcductmeasurement systems to obtain fission o
. product release data d'uring planned SFD Series 1-3 and 1-4 f
tests with pre i radiatea fuel. Obtain chemical fann and
/'
aerosol data for fission products released in these tests.
Collect, analyze, and report these fission product release and transport data.
Determine '1'sotopic and chemical composition of aerosol samples collectedo[th'efiltersofthegassamplebombs. Collect and anal [ze fission product data obtained from Ge and Nal gamma-ray spectrometers located on the steam line near the IPT. Determine chemical forms of-fission products by analyzing samples with laser Raman spectrometer and fluorescence spectrometer. Improve analyt1 cal methods for identification of fissien products for:
(a) chemical forms in solution, (b) Beta-particle emitting radio-nuclides, and (c) hydrogen iodide and organic iodines. Incorporate aerosol measurement system on steam line near in-pile-tube and measure the time-dependent aerosol particle' concentration and the particle size fractions. Estimate mean aerodynamic diameter size of aerosol particles.
BUDGET ($K):
FY 84 '
FY 85 FY 86 1340 345 0
I f
18 I
r
Project 19 (FIN NO. 80127)
TITLE: FP RELEASE FROM LWR FUEL OBJECTIVES: Investigate experimentally the magnitude and physiochemical for1n of fission products and aerosols released from connercial irradiated LWR fuel under the elevated temperature and environ-mental conditions characteristic of severe fuel damage and core melt accidents.
l This project consists of two tasks. Task A is a separate effect study which measures the quantitative release of fission products from a commerically irradiated fuel rod segment due to parametric variation. Task B is an integral study of the silver and borop control rod behavior in the presence of adjacent fuel rods. The simulant fuel rod bundle used here simulates a portion of the reactor core where a vertical as well as a radial temperature gradient exists.
Task A will focus en a more detailed study of the effects of time, temperature, steam flow, fuel burnup and heating rate on release. Task 8 will focus on the aerosolization of Ag and Cd after the control rod burst and the decomposition of l
CsI by the borate compounds.
BUDGET ($K):
FY 84 FY 85 FY 86 1055 1460 1200 t
l 19
Project 20 (FIN NO. 80453)
TITLE:
POST ACCIDENT FISSION PRODUCT CHEMISTRY OBJECTIVES: Detennine the chemical species of various fission products present in aqueous reactor solutions and their liquid / vapor distribution under representative reactor accident conditions.
The main objective of this project is to detennine the gaseous iodine concentration in the containment atmosphere given that most of the fodine released went into solution in the containment sump. The re-evolution of the dissolved iodine at various conditions (particularly in the presence of radiation and silver) is the main concern of this project. Investigation will focus on the modelling of the fodine partition coefficient and the study on the organic iodine formation.
BUDGET ($K):
FY 84 FY 85 FY 86 300 300 300 l
I O
e B
e 20
Project 21 (FIN NO. B0488) 9 TITLE:
TRAP-MELT VERIFICATION PROGRAM OBJECTIVES: To conduct small scale tests on fission product and aerosol transport to develop a data base for the early assessment of the TRAP-MELT code. Small scale experiments concentrating on 4
the validation of phenomena of fission product and aerosol transport included into primary system transport and deposition code (TRAP-MELT). This includes an examination of aerosol transport through vertical pipe, using plasma torch to generate aerosols, for a spectrum of conditions, an examination of resuspension phenomena during fission product and aerosols transport, and an examination of upper plenum simulation on the effect of transport and deposition.
To participate in the review, planning, and implementation, and analyzing of the results of the severe fuel damage (SFD) series I tests; to participate in the review, planning, implementation, and analyzing of the results of the large scale fission product and aerosol transport tests in the MARVIKEN facility.
- BUDGET ($K):
FY 84 FY 85 FY 86 407 325 500 21 i
.-.~.-.
Project 22 (FIN NO. B0815)
TITLE:
FP DEPOSITION OF AEROSOLS OBJECTIVES: To examine fission product movement and interactions via aerosols by determination of the deposition capacit'y of aerosols, the rate of deposition, and the deposition of key fission product vapors onto aerosols relative to other structures.
Provide experimental data on the sorptive capacity of prototypic aerosol materials for selected important fission products and to determine the sorption rates and controlling deposition mechanisms (condensation chemisorption, etc.).
BUDGET (SK):
FY 84 FY 55 FY 86 285 325 300 1
22
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Project 23 (FIN NO. 80827)
TITLE:
IN-PILE FISSION PRODUCT AND AEROSOLS TESTS: TECHNICAL SUPPORT OBJECTIVES: Participate in the review process of all technical reports and other technical documents in relation at PBF Series 1-3, and 1-4 tests and provide comments to support evaluation of and planning for PBF Phase I experiments.
Continue the review process of the results obtained during PBF scoping test and 1-1 test. Participate, in the collaboration with EG&G, in report (s) preparation.
Conduct, if necessary and warranted and as directed by NRC, independent analyses of selected samples obtained during the scoping and 1-1 tests to independently ccnfinn findings and/or interpretation provided by others.
Participate in the review process of all technical reports and other technical documents in relation to the planned separate effect fission product and aerosols test at ACRR (SNL) and to the planned large integral tests at NRU (PNL/AECL) to support the evaluation of the technical approach and experiments planning.
Integrate the on-going ORNL fission product release from fuel experiments (HI series) and small scale care-melt experiments
.into planned ACRR/NRU in-pile test program and test matrix.
SUDGET ($K):
FY 84 FY 85 FY 86 143 150 150 23 O
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Project 24 (FIN NO. 80831) i TITLE: MARVIKEN FISSION PRODUCT AND AEROSOL TESTS: TECHNICAL SUPPORT OBJECTIVES: To provide technical support for the fission product and aerosol large tests of the MARVIKEN facility (Sweden) especially in the area of fission product and aerosols generation (fissium and corium) including efficiency determination, measurements and selected design aspects (upper plenum simulator design, PWRIBWR corium recipe, etc.).
Participate in the review process of all applicable MARVIKEN (as specified by NRC) technical documents and provide comments as a part of on-going review process to support and evaluate the experimental results.
As directed by NRC provide analyses and conduct small-scale experiments which might be needed to support the evaluation of the MARVIKEN experiments.
directed by NRC participate and provide a technical support in the inter-laboratory review process (SNL, EG&G, 8CL) of the MARVIKEN experiments.
As directed by NRC provide support for the U.S. scientist parti-cipation in the MARVIKEN experiments.
BUDGET ($K):
FY 84 FY 85 FY 86 166 175 0
24
. - - - -. - - -,, - ~. -. - - - - -
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Project 25 (FIN NO. 80121)
TITLE: AEROSOL RELEASE & TRANSPORT OBJECTIVES: Perfonn experiments to investigate phenomena affecting generation
~
and subsequenct behavior of aerosols over a range of severe accident conditions. Develop a data base for validating pre-diction models on generic source terms for molten fuel within containment.
Two series of experiments are planned at ORNL to obtain quanti-tative information on the phenomena which affect either the characteristics or behavior of LWR aerosols.
The first series of experiments will be conducted in the Nuclear Safety Pilot Plant (volume of 38.3 m' vessel). The independent variables of interest include: relative humidity and all possibla combinations (single and multi-component) of U 0, Fe 0, and 38 23 limestone concrete aerosols.
The second series of experiments and will be conducted in the Aerosol-Moisture' Interaction Vessel (volume of 0.56 m').
The independent variables of interest include: relative humidity.
aerosol components (U 0, Fe2 3, ifmestone concrete, Ag + Cd),
0 38 aerosol mass ratios, and aerosol concentrations.
The questions of interest include: at what level of relative humidity do the various aerosols or combinations of those aerosols change shape factors to look more ifke spheres and nt like chain agglomerates, what are the functional relationship among aerosol fallout behavior and the set of independent variables of interest, and what quantitative change in measured responses (mass concen-tration, aerodynamic size distribution, fallout as a function of time, microscopy) can be shcwn to be due to a given cuantitative change in initial conditions (i.e., independent variables)?
BUDGET ($K):
FY 84 FY 85 FY 86 1300-1300 1000 25
Project 26 (FIN NO. A1339)
TITLE: MELCOR OBJECTIVES: Provide the replacement risk code (MELCOR) for the MARCH, MATADOR, and CRAC codes which: has a structure readily amenable to incorporation of new models based on the ongoing experiment research program; and pennits the quantitative analysis of both "best estimate" severe accident consequences and the associated uncertainties. -
BUDGET ($K):
FY 84 FY 85 FY 86 1670 1050 500 e
M*
26 O
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Project 27 (FIN NO. A1019)
TITLE: MOLTEN FUEL-CONCRETE INTERACTIONS OBJECTIVES: To characterize the chemical and physical phenomena associated with the interactions between molten LWR core materials and concrete likely to be encountered during hypothetical fuel melt accidents and to develop models for predicting these phenomena.
BUDGET ($K):
FY 84~
FY 85 FY 86 280 320 350 1
I r
s
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I L
27
s project 28 (FIN NO. A1030) '
TITLE:
MOLTEN CORE-COOLANT INTERACTIONS OBJECTIVES: To develop an understanding of the nature of core. melt-coolant interactions during light water reactor hypothetical accidents; To develop a data base for assessing the threat of core melt-coolant interactions to the integrity of the reactor vessel' and containment; To provide analysis and models for tha reactor meltdown progression and the phenomena associated with core melt-coolant interactions; and To perform experiments to investigate phenomena affecting core melt-coolant interactions.
These series of experiments are planned to obtain quantitative information on the phenomena which affect one or more of the three phases of molten-core coolant interactions:
coarse mixing, triggering, or propagation.
In order to obtain the most information per dollar spent, each series will be based on a statistically designed experiment, e.g., a fractional factorial design.
For all three series of experiments the cuestions of interest include:
Is there a limit to the amount of melt mass which can participate in coarse mixing (series 2)? What is the functional relationships of melt mass and conversion ratio (thermal to kinetic) and what quantiative change in measured response (e.g.,
pressure, hydrogen production, debris particle size) can be shown to be due to a given quantitative change in initial conditions
~
(i.e., independent variables)?
BUDGET (SK):
FY 84 FY 85 FY 86 700 1115 1500 28
t i
Project 29 (FIN NO. A1198) 4 TITLE:
CONTAINMENT ANALYSIS OBJECTIVES: To develop a computer program CONTAIN that models a power-reactor containment system and predicts the response of the system to accident-imposed conditions.
BUDGET ($K):
FY 84 FY 85 FY 86 1100 950 1100 e
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Project 30 (FIN NO. B6747)
TITLE:
RADIONUCLIDE RELEASE UNDER SPECIFIC LWR ACCIDENT CONDITIONS (BMI-2104)
OBJECTIVES: Provide. analytical models and computer codes for application to the analysis of release and transport of fission products in LWR plants under severe accident conditions. Improvement to the primary system transport and deposition code (TRAP-MELT).
4 BUDGET ($K):
FY 84 FY 85 FY 86 1036 0
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Project 31 (FIN NO. 82444)
TITLE:
ESF SYSTEMS EFFECTIVENESS EVALUATION -
OBJECTIVES: Obtain and develop information that will aid in providing best estimates of the chemical and physical properties of environments associated with severe accident phenomena to predict performance of selected ESF systems (in these envi.-
ronments) emphasi$ing fission product depletion mechanisms.
This includes filtration, sprays, suppression pools and ice condensers.
To develop mechanistic code for suppression pool effective-ness in removal fission product and aerosols (SPARC).
To develop mechanistic code for the ice condenser effective-ness in removing fission product and aerosols (ICEDF).
Validate SPARC using EPRI suppression pool experiments.
Validate ICEDF by performing large integral tests or small separate effects experiments (planned).
BUDGET ($K):
FY 84 FY 85 FY 86 675 875 800 31
Pro.fect 32 (FIN No. A1383)
@EST I (described elsewhere)
@EST II (Planned)
(To be provided later)
O e
4 4
4 32
Project 33*
TITLE:
SUPPRESSIONPOOLSCRUB8INGEXPERIMENTS(EPRI)
(Described elsewhere)
BUDGET:
Not Known Not an NRC experimental program, but it is expected to provide a valuable data base to validate the SPARC (NRC) and SUPRA (EPRI) suppression pool scrubbing codes.
6 6
4 4
33 8
k - - - -
l Project 34**
TITLE:
LWR AEROSOLS CONTAINMENT EXPERIMENTS (EPRI LACE PROGRAM) l OBJECTIVES: To provide a data base for validating containment aerosol and thermal hydraulic computer codes, and to investigate (in large scale) inherent radioactive aerosol retention behavior for postulated high consequence accident situations. This will include containment bypass sequences, early containment leakage or failure to isolate, and delayed containment failure (if aerosol resuspension proves to be significant, or core damage was similarly delayed)
BUDGET:
Not known; foreign participation is being negotiated.
- Not an NRC experimental program, but it is expected to provide valuable data base to validate some of the containment aerosols transport ecdes, including the investigation of containment bypass release (V Sequences).
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project 35 TITLE:
AEROSOL TRANSPORT TESTS (MARVIXEN PROGRAM) 08JECTIVES: To provide information on the deposition rates in primary system components of selected fission product, control rod, and structural materials at temperatures and scale approaching those found in reactor systems under accident conditions.
The MARVIKEN experiments provide for a large, integral, high temperature system with a complex mixture of fissium and corium being injected into the reactor vessel and then transported through various pipes and components.
The results are intehded to show the effects of multiple components along the flow path, fissium/corium interactions, fissium interactions with surfaces and flow irregularities in large vessels. The effects of scale and multiple flow components are not being studied elsewhere and, hence, the MARVIKEN experiments serve as a unique source for this type of infomation.
BUDGET ($X):
FY 84 FY 85 FY 86 460 400 400 (NRC share 12.47. of the total budget.)
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9 ENCLOSURE 4 PROJECT AND BUDGET PROPOSAL FOR NRC SOURCE TERM RELATED RESEARCHPROJECTS(FORMS 189) e D
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i ENCLOSURE 4 PROJECT AND BUDGET PROPOSAL FOR NRC SOURCE TERM RELATED RESEARCH PROJECTS (FORMS 189) i i
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March 1, 1984 FY 1984 PROGRAM BRIEF PROGRAM: AE TITLE: MODIFICATIONS TO FISSION. PRODUCT MEASUREMENT FIN NO: A682,9 SYSTEMS FOR SFD SERIES 1 TESTS CONTRACTOR: EG&G SITE: 'INEL i STATE: IDAHO NRC PROGRAM MANAGER:
P. REED EG&G PROGRAM MANAGERS:
D. B. VAN LEUVEN, P. MACDONALD PRINCIPAL INVESTIGATORS:
P. MACD0NALD, A. APPELHANS, R. HOBBINS, D. OSETEX OBJECTIVES:
l (1) TO UPGRADE FISSION PRODUCT MEASUREMENT SYSTEMS TO OBTAIN FISSION P RELEASE DATA DURING PLANNED SFD SERIES 1-3 AND 1 4 TESTS.
(2) TO OBTAIN CHEMICAL FORM AND AEROSOL DATA FOR FISSION PRODUCTS RELEA IN SERIES 1 SEVERE FUEL DAMAGE TESTS.
(3) TO COLLECT, ANALYZE, AND REPO,RT FISSION PRODUCT RELEASE AND TRANSPORT OATA OURING SERIES 1 SEVERE FUEL DAMAGE TESTS.
l j
(4) TO INVESTIGATE FISSION PRODUCT TRANSPORT AND BEHAVIOR UNDER SEVERE ACCIDENT CONDITIONS.
BUDGET ACTIVITY: 60190201 FY 1984 OBLIGATION: $1010K O
n.
2 TASK A: MODIFICATIONS FOR SFD SERIES 1 TESTS.
PERFORM THE FOLLOWING UPGRADES FOR THE SFD SERIES TESTS:
1.
INCORPORATE FILTERS ON SIX.AODITIONAL GAS SAMPLE BOM8S TO COLLEC PAR 1ICULATES PRESENT IN THE GAS PHASE OURING THE TESTS.
FY 84 COSTS: 5170K DETERMINE THE ISOTOPIC AND CHEMICAL COMPOSITION OF THE AEROSOLS a.
AND THE CHEMICAL FORM OF THE FISSION PRODUCTS ASSOCIATED WITH T AEROSOLS.
2.
A00 Ge AND Na! GAM 4A-RAY SPECTROMETERS TO THE STEAM LINE NEAR THE IP FY 84 COSTS: 5345K 515K IS AUTHORIZED TO PREPARE TECHNICAL ASSESSMENT AND SCHEDULE a.
ESTIMATES FOR INSTALLATION OF THE GAMMA-RAY DETECTORS TO VIEW THE STEAM LINE JUST A80VE THE IN-PILE TU8E HEAD. THE ASSESSMENT SHOULO INCLUDE M00!FICATIONS TO PORTIONS OF THE WORKING PLATFORM, INSULATING JACKET, COLLIMATOR AND STEAM LINE SHIELOING AND DESIGNS FOR AN INCLINED COLLIMATOR TUBE, SHIELO 80X, STRUCTURAL SUPPORT SYSTEM, AND PLANT SUPPORT SYSTEMS (NITROGEN, AIR, ELECTRICAL, CONTROL, ETC.)
PREPARE A REPORT ON THE DESIGNS AND MODIFICATIONS AND PROV10E A
. FIRM SCHEDULE TO NRC.
b.
IF THE ASSESSMENT AND SCHEDULE ESTIMATE CONCLUDE THAT THE DETECTORS CANNOT BE PLACED ABOVE THE IN-PILE TUBE, THE DETECTORS SHOULO BE PLACED AS CLOSE AS POSSIBLE TO THE IN-PILE TUBE HEAD ANO UPSTREAM OF THE MAIN FLOOR STEAM SAMPLE BOMBS.
3.
IMPROVE FISSION Pk0b'UCT COMPOUND IDENTIFICATION BY RAMAN AND SPECTROSCOPY AND DETERMINE LIBRARY (REFERENCE) SPECTRA STANDARDS OF EXPECTED FISSION PRODUCT COMPOUNDS FOR THE LASER RAMAN SPECTROMETER AND FLUORESCENCE SPECTROMETER FY 84 COSTS: $205X
3 M00!FY THE EXISTING MOLECULAR OPTICAL LASER EXAMINER TO INCLUDE A a.
QUARTI FLUORESCENCE SPECTROMETER.
b.
DEVELOP A SET OF FISSION PRODUCT COMPOUNO STANDARD REFERENCE SPECTRA FOR COMPARISON TO SERIES 1 TESTS SAMPLES.
c.
DEVELOP CATALOG OF RFr
.CTRA.
d.
DEVELOP METH005 FO..
.fSES OF TEST SAMPLES.
PREPARE REPORT ) RESENTING CATALOG OF SPECTRA AND PROCEDURES e.
I DEVELOPED FOR TEST SAMPLE ANALYSES.
4 IMPROVE ANALYTICAL METH005 FOR IDENTIFICATION OF FISSION PRODUCTS a)CHEMICALFORMSINSdLUTION,b)8ETAEMITTINGRADIONUCLIOES, FOR:
AND c) HYOIROE COMPOUNDS, NYOR0 GEN IODIDE AND ORGANIC 100!NES.
i FY 84 COSTS: 5220K 5.
ADO A LIQUID COLLECTION SYSTEM TO THE LIQUID SAMPLE LINE AND DETERMINE FISSION PRODUCT CONCENTRATIONS IN THE GAS AND LIQUID PHASES OVER A PERICO 0F TIME FOLLOWING THE TESTS 70 SIMULATE POST-ACCIDENT BEHAVIOR OF FISSION PRODUCTS.
FY 84 COSTS: 570K a.
THE WATER AND GAS SAMPLES SHOULO 8E ANALYZED FOR FISSION PRODUCTS BEFORE AND AFTER THE ADDITION OF SOCIUM 80 RATE TO THE WATER SAMPLES.
. ~
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1
May 14, 1984 REVISED FY 1984 PROGRAM BRIEF PROGRAM: AE TITLE: TRAC /MELPROG INTEGRATION FIN NO: A7303 i
CONTRACTOR: LANL SITE: LOS ALAMOS STATE: NEW MEXICO NRC TECHNICAL MONITOR:
J. T. HAM PRINCIPAL INVESTIGATOR:
R. HENNINGER OBJECTIVE: TO PERFORM DETAILED THERMAL-HYORAULIC CALCULATIONS IN THE UPPER PLENUM REGION TO STUDY FISSION PRODUCT TRANSPORT AND DEPOSITION.
800GET ACTIVITY: 60190201 FY 1984 OGLIGATION: $208K l
THIS ORDER: $120K
]
suiAL past FY 1984 SCOPE REVISED (10/01/83 - 09/30/84):
1.
ISSUE A REPORT FOR A TRAC NATURAL-CONVECTION CALCULATION IN THE UPPER PLENUM REGION OF THE SURRY PWR VE5SEL.
2.
DEVELOP TWO-0!MENSIONAL DEGRADED-CORE THERMAL-HYORAULIC CAPA8!LITIES FOR TRAC /MELPROG.
3.
ISSUE A TRAC /MELPROG USER MANUAL INCUDING THE DESCRIPTION 0F THE COUPLING 8ETWEEN THE MELPROG AND TRAC-PF1 CODES.
4 COOPERATE WITH $NL IN MELPROG.0EVELOPMENT.
5.
PROVIDE ON-CALL TECHNICAL ASSISTANCE TO NRC.
=mme%.
4 4
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4
GENERAL RESEARCH HISTORY BEGINNING TO 1974 o
ATOMIC ENERGY COMMISSION o
DEVELOP REACTOR TECHNOLOGY o
DATA BASE FOR SAFETY ISSUES ECCS/ THERMAL HYDRAULIC METALLURGY FUELS 1975 TO PRESENT o
NUCLEAR REGULATORY COMMISSION o
SAFETY RESEARCH o
CONFIRMATORY o
EXPLORATORY O
m.
e HISTORY OF SEVERE ACCIDENTS L
I 1957 WASH-740 o
A CONSEQUENCE ANALYSIS OF CORE MELT WITH DIFFERENT CONTAINMENT PERFORMANCE 1963 TID-14844 o
A TECHNICAL RATIONALE FOR REGULATORY TREATMENT OF SEVERE ACCIDENTS t
1975 WASH-1400 0
A RISK ASSESSMENT OF LARGE POWER REACTORS o
A PRIMER FOR ACCIDENT CAUSES j
o A MECHANISTIC ANALYSIS OF CORE MELT BEHAVIOR--PRODUCING A SOURCE TERM 1
1978 IMPROVED REACTOR SAFETY PROGRAM i
0 RISK BASED IMPROVEMENTS f
CORE CATCHERS FILTERED VENT CONTAINMENT ADD ON DHR SYSTEMS UNDERGROUNDING RELIED ON WASH-1400 MECHANISTIC ANALYSIS OF COREf o
MELT f-b o
s mm...
E HISTORY OF SEVERE ACCIDENTS (CONT.)
1980 REVIVAL OF INTEREST IN CORE MELT o
RECOGNITION THAT CORE MELT ACCIDENTS DOMINATE RISK o
TMI ACTION PLAN--DEGRADED CORE RULEMAKING 1
o LETTER FROM STRATTON, MALINAUSKAS, AND CAMPBELL t
o NRC INTEROFFICE COMMITTEE ON DEGRADED CORE RULEMAKING o
PAPER BY LEVENSON AND RAHN 1981 INITIAL EVALUATION o
WASH-1400 SOURCE TERMS APPEAR TO BE PESSIMISTIC o
SARP PLAN BASED ON PHASE ! WITH WASH-1400 SOURCE TERMS, PHASE !! TO CONFIRM OR REDUCE 1982 DEMAND FOR NEW SOURCE TERM o
SOURCE TERM PLAN o
SOURCE TERM PROGRAM 0FFICE o
DELAY OF SARP PHASE I f
RES FTE MANAGEhENT PLOT - 6/13/84 NSTANTAEOUS FTE USAGE RATE 3
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ORGANIZATION AND FUNCTION 1.
WHO IS DOING THE RESEARCH?
o NRC OFFICE OF NUCLEAR REGULATORY RESEARCH o
SANDIA NATIONAL LABORATORY (SNL)
IDAHbNATIONALENGINEERINGLABORATORY(INEL) o o
0AK RIDGE NATIONAL LABORATORY (ORNL) o PACIFIC NORTHWEST LABORATORY (PNL) o LOS ALAMOS NATIONAL LABORATORY (LANL) l l
0 BATTELLE COLUMBUS LABORATORY (BCL) o ARGONNE NATIONAL LABORATORY (ANL) o BROOKHAVEN NATIONAL LABORATORY (BNL) l l
0 SELECTED CONTRACTORS l
2.
HOW ARE THEY SELECTED?
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Office of the Director
{
Director:
R. Minoaue Dep. Director: D. Ross i
Associate Director for Policy. Planning and Control Scientific Programs:
Staff C. Kelber Director: F. Gillespie
- I F
Olvision of Risk Analysis
{ Sivision of Engineering Division of Accident and Operations Technology Evaluation Director: R. ternero Ofrector: K. Goller Director: 0. Bassett Dep. Dir.: M. Ernst Dep. Dir.: P. Comella Director: G. Arlotto Dep. Dir.: W. Morrison Dep. Dir: L. Shao Material Engineerint Containment Systems Reactor Risk Branch Waste Management Branct Research Branch Branch Chief:
G. Burdick Chief:
E. Coatt Chief R. Curtis Chief C. Serpen Earth Sciences granch Mechanical /Structurel Regulatory Analysis g Engineering Srsach Reactor Systems Materials Risk Branch Research Branch g
Chief: Vacant Chief:
J. Malere Chief L. Shotkin Health Effects Orsach Electrical. Engineering Fuel Systems Research manen Facters &
Chief:
W. Mills l
Instrumentation &
Safeguards trench Branch Control Aranch Chief:
W. Morris Chief:
M. Silberbert Chief: J. Norberg 7
Occupational Radletion Protection Branch Chanical Engineering Aranch Chief:
R. Alesander
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WHAT INTERNAL AND EXTERNAL ADVISORY AND TECHNICAL REVIEW COMMITTEES ARE IN PLACE?
A.
HOW ARE MEMBERS CHOSEN?
B.
WH0' WRITES THEIR CHARGES?
o RESEARCH REVIEW GROUPS o
ACRS:
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS T
ANNUAL REVIEW 0F RESEARCH BUDGET AND PLANNING SPECIAL SUBCOMMITTEE REVIEWS iZ.
o SELECTED SUBJECTS PEER REVIEW INDIVIDUAL
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STANDING COMMITTEE
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WHO ARE THE PROGRAM MANAGERS AND WHAT PROTOCOLS GOVERH THEIR PARTICIPATION IN THE DESIGN OF THE RESEARCH
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5.
HOW DO YOU MANAGE QA/QC?
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6.
HOW ARE THE RESEARCH RESULTS DISSEMINATED BOTH INTERNALLY AND EXTERNALLY?
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SOURCE TERM RESULTS A.
SOURCE TERM TECHNOLOGY
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o BMI-2104 REFERENCE PLANTS REPRESENT SPECTRUM OF DESIGN T
fi _
o CONTAINMENT GROUPS' NEW PERSPECTIVES o
GENERAL ESTIMATES OF LWR RISK AND REGULATORY SIGNIFICANCE B.
SARRP REPORTS
_ - ~ ~ - ~
o BASED ON SOURCE TERM AND CONTAINMENT GROUPS o
SPECIFIC ESTIMATES OF REFERENCE PLANT RISKS o
COST-BENEFIT ANALYSES FOR REFERENCE PLANTS
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o EXTRAPOLATIONS Nhee 6s'u'N e--w.
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a NUCLEAR REGul.ATORY RESEARCH APPLICATION OF RESEARCH PR0'JECTS TO REGULATORY PROCESS 1
i 1
SEPTEMBER 15, 1982 1
I
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ilTLE DESCRIPTION RELEVANCE APPLICATION TO REGULATORY PROCESS PR(,
Sealscale Evaluate PWit systems behavior 10CFR50, App. A
. Provided data on:
under simulated accident Criterion 34,
- Core uncovery and two-phase natural circulation.
sequences leadlag to degraded Appendix K, and
- Pumps-on versus pumps-off effects.
core conditions.
proposed rulemaking
- Small break tests with natural circulation effects.
. Will investigate UHI plant safety features.
NRC Action Plan Pumps on/off resolution, tuo phase natural circulation including (NUREG-0660and-0737)
Inert gas effects, small, intermediate, and large break data for non-UHI and t#fl. loss of power transients, transients involving steam generator tube breaks, secondary side initiated transients.
-and general NRR support.
8WR FIST Evaluate BWR systems behavtor' 10CFR50. App. A Will provide data from simulated BWR small and intermediate break under a verf aty of simulated Criterion 34 & 50.
LOCA. ATWS and other transients to:
accident sequences Appendix K, and planned
- leprove understanding of 8WR transients rulemaking
- assess TRAC-BWR calculational capability
- Imorove response of operators to BWR transients
- evaluate system design leprovements.
Separate Effects Measure BWR & PWR bundle 10CFR50 App. A
. Provided data on:
Exp. and Model thermal-hydraulle phenomena Criterion 35 & 50.
- Bundle heat transfer under degraded core cooling conditions.
Development and develop predictive models. Appendix K, and planned
- 5.G. performance under accident conditions.
rulemaking
- Pressurized thermal shock to vessels.
Unresolved safety issue support
- PWR ECCS systems performance.
Evaluate 8WR ECCS response
- Ifundle blockage effects under severely damaged core conditions.
under accident conditions.
- 8WR separate effects data to support IRAC-8WR assessment
- Natural circulation long tens cooling.
Advanced instrumentation to meet research needs and reactor safety needs.
2D/30 Program Conduct large-scale test of 10CFR50, App. A Provide reflood data concerning steam binding assumptions in Appendix'K.
emergency core cooling during Criterton 35, & App. K.
ref t11-reflood portion of large-break loss-of-coolant accident, Provide small break natural circulation data.
and natural circulation during a small-break loss-of-coolant Provide blocked bundle data for reflood assumptions in Appenglx,K and degraded core rulemaking.
accident.
Provide large-scale ECC bypass infonnation for design base accident.
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I't f DESCRIPTION RELEVANCE APPLICATION TO REGULATORY PROCESS q
Code Improvement and Develop computer codes to 10CFR40 App. A Codes used in audit of PWR and SWR plants.
Malatenance analyze reactor operational Criterion 35, 38 & 54, transients, small breaks, and App. K, and planned Codes used to calculate plant response to LOCAs and operational transients design base accidents, rulemaking Codes used to evaluate operator guidelines.
Code Assessment and Assess the a'ccuracy of computer 10CFR50 App. A Evaluate vendor margins in design base accident analysis.
I I
codes to analyze reactor opera-Criterion 35, 38 4 50 tional transient small breaks App. K, and planned Evaluate ECCS response under operational transients and LOCAs.
and design basis accidents.
rulemaking.
These will be used to evaluate plant response and licensee provide analysis base for 11 censing and safety issues, such as j
conserva tism.
pressurized thermal shock.
Fuel Behavior Under Detenefne fuel and clad behavior 10CFR50 App. A Previous data have been used to assess and confirm these areas in l
Operational Transients under anticipated operational Criterton 35, 50 & 60, 10CFR50. Current work to confirm clad ballooning and blockage for transients and design basis and App. K accidents.
08A-LOCA will complate this effort.
Loss-of-Fluid-Test in a scaled PWR, test the 10CFR50 App. A Helps evaluate the degree of conservatism in 10CFR50 App. K.
(LOFT) hydraulics and fuel thermal General Design Crit.
transients during large break LOCAs and operational transients.GDC 13 Instru. &
Tests confirmed that the emergency core cooling system required by 10CFR50 prevents major core damage during design
- basis loss-of-and anticipated transients without Control scram (ATWS).
coolar4 accidents.
.GDC 15 Raactor Coolant During these accidents, assess
System Design
Large-scale data on ECC bypass used to' leprove codes.
the response of plant lastrumen-tation and new operator diagnos *.60C 35 Emergency Core Small-break pumps off/on tests provided data needed to determine the licensing position on issue of tripping reactor coolant pumps during tic apd display techniques in Cooling System small break.
helping the operator to identify and respond appropriately to App. K-ECCS Evaluation Operational transients and multi-fallure accidents provided data to the accident.
Model e
confins corrective operating procedures and ef fectiveness of verlous t
cooldown procedures including natural circulation.
Assess the effectiveness of Task Action Plan prescribed plant recovery INUREG-0660)
Intermediate sized break used to assess codes and models.
procedures.
I.D.5 Control Room Large break LOCA test L2-5 with pressurfred fuel will address the
/
Design II.E.2 ECCS LOFT -
margins to clad ballooning and rupture in FV82.
specifically referenced. The generic issue of plant response to an anticipated transient without scram will be addressed by two ATWS tests in FY82.
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, ' p5 TI LE DESCRIpTI0ll SELEVANCE APPLICATION TO REGULAT0lrt PROCESS i
man 1-(
Loss-of-Fluid Test
.(cont'd)
I.A.2 Training & Qual The generic issue of plant response to an anticipated transient (ffcation of Operators without scram will be addressed by two ATWS tests in FV82.
~
I.A.4 Simulator Use &
Five anticipated transients will stub the boron dilution accident.
Development the control rod withdrawal accident, and the steam generator tube break which recently occurred in the Ginna reactor.
1.C Operating precedures Large break fuel damage test in FY83 wifI demonstrate core coolability 11.8 Degraded Cores when in a degraded cooling condition, and confirm criteria for clad ballooning and burst in a large fuel bundle in actual nuclear core accident conditions.
Accident Evaluatten Research on coolability of planned rulemaktsgs Support for accident management procedures for the prevention, early and Mitigation severely damaged cores and termination and/or mitigation of severe reactor accidents. Specific 4
severe accident hydrogen and radiological source terms.
appilcation to the NRC development of hydrogen control rules, reevaluation Severe accident sequence of the technical basis for radiological source term for use.in the l
development of policy on siting and emergency preparedness and development analysis and mitigation of of policy on equipment quellfication requirements. Direct inputs effects of combustible gases to the improvement of the phenomenological base for reliability and and retention of molten core risk assessments.
materials.
Fast treeder An integrated research program 10CFR50 10CFR100 Support for NRC licensing actions on CASR. Development of Generic to provide NRC with data and methods of analysis to regulate Design Criteria. Siting Criteria and Regulatory Guides and Standards for IJIFOR.
breeder reactors when required.
,a Advanced Converter Development of safety and 10CFR60 10CFR100 Support for IIRC licensing actions and for preappilcations and for licensing related. data and 4
evaluation methods for 945-preapplication review for large demonstration plant.
cooled, graphite moderated nuclear power reactor plants
[r 1
,n (HTGRs). Work currently focuses on preparations to 4
enable licensing of the new.
l more advanced design HTGRs including an early examinatten of siting characteristics.
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TITLE DESCRIPTIOI8 RELEVANCE APPLICATION 10 REGULATORT PROCESS Mechanical and Develop methodologies and pro-10CFR50. App. A Assess and recoemend improvements to the Standard Review Plan Sections Structural vide infonnation to characterize GDC l-5 2.4, 2.5, 3.5, 3D, 3.8 and 3.9.
Engineering the behavior of structures, com-ponents, systems and equipment 10CFR50, App. 8 Provide research support for unresolved safety issues pertaining to under operational. environmental and asyneetric blowdown, seismic design, snubbers, tornado missiles and postulated accident conditions. 10CFR50, 50.55a safety relief valve dynamic loads.
Verify mechanical / structural computer codes used to perfons 10CFR100. App. A Support rulemaking on degraded core cooling by determining safety safety analyses. Detenmine the margins and modes and probabilities of failure for structural behavior of structural and components, mechanical systems under earth-quake and other accident loads Support Systematic Evaluation Program for operating plants.
In a probabilistic format to risk. Assess hou loads are Support development of Regulatory Guides applicable to mechanical /
combined. Evaluate qualification structural components and systems.
criteria of mechanical components and equipment.
Support rulemaking on qualification of mechanical equipment.
including research to oewelop regulatory guides on equipment.
l quellfication in support of MRR action plan.
Fracture Development of elastic-plastic 10CFR50, App. A. GDC Assess RPV and pipe toughness in accidents Mechanics fracture mechanics analysis, 14, 15, 30, 31; App, test method and data base.
A. G.
Develop thermal shock 10CFR50, App. A.
Modify piping design basis accident; assist resolutions of analysis and test asymmetric load problem.
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10CFR50, App. A.
Upgrade, rewrite piping design rules. Assist S.E.P. evaluation Integrity of cracked.
ASME Roller and of Past pipe fracture consequences, degraded piping.
Pressure Vessel Probability and structural consequences of pipe breaks.
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- Pt(L ftE DESCRIPTION RELEVANCE APPLICATION TO REGULATORY PROCES$
i Operating Effects Escrittlemen't of RPV Steel.
10CFR50. App. G. H.
Medley or confire vessel toughness requirements.
1 Fatigue crack growth 10CFR50. App. A.
Assess future safety of component with growing flaw per GDC 30. 31.
ASME-XI.
Stress corrosion cracking 10CFR50. App. A.
Assess "flaes" prop. by vendors; modify water chemistry of 8WR. PWR piping.
GOC 14. 15, 31 procedures.
Integrity of steam 10CFR50. App. A.
Assess safety of cracked tubes; assess proposed design generator tubing.
GDC 14, 15, 31.
and operating.
ASME toller and Pressure Vessel Code liondestructive Reliability of flaw 10CFR50. Sec. 50 Tighter inspection of ASME-XI will find smaller flaws.
Examination detection and evaluation 55a. App. A.
of current code methods.
GOC 1. 32.
Improved flaw detection and App. 8. Criterien leproved techniques and equipment to find smaller flaws.
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accuracy of characterization III.
i during 151.
1 Continuous monitoring of ASME Boller and Early warning of cracking, and guide for better I5I during reactors for on-line Pressure Vessel shutdowns.
evaluation of structural Code i
integrity.
4 Fuel Cycle Factitty Develop esperimental data 10CFR40.32 Provide data on possible design basis accidents and Safety and validated analysis 10CFR50.35 methods for assessing safety in fuel cycle facilities.
methods to assess radiological 10CFR50.50 source tems for accidents of 10CFR50.57 major consequences in fuel 10CFR70.23 cycle facilities.
10CFR70.31 10CFR72 l
Develop experimental data and 10CFR70.22 Provide NRC staff with improved capability to verify I
validated analysis methods to 10CFR70.24 criticality safety for configurations and conditions assess criticality safety for 10CFR72 not previously assessed.
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configurations applicable to 10CFR71.33 fuel fabrication, shipping 10CFR50 aM surap.
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1 FuelCycleFactitty)
Collect and analyze experi-10CFR72 Provide data to establish and verify eatsting criteria for Safety (continued mental data to assess the
. storing spent fuel.
effects of storing spent List i
fuel in a dry environment.
Decommissloaing Detemfrelevels of radio-10CFR20 Provide technical data to support the development of activity present in nuclear 10CFR30 decomelssioning standards and guides and to evaluate facilities, assess tech-10CFR40 Ilconsee implementation.
niques, safety and costs 10CFR50.82 for reducing residual levels lie'970 to specified levels and 10CfR72 assess methods to verify r
these residual levels.
Effluent Control Provide experimental data on 30CFR20.106 Provide a' technical basis for ifconsing evaluation of waste the radionuclide concentrations 10CFR20. App. 5 stream characteristics and the perfomance of waste treatment In waste streams in LWRs and 10CFR50.34 systems to assure affluents are ALARA.
8 the performance of methods to 10CFR50.34a reduce releases to the 10CFR50. App. I environment.
l Electrical Equipment Evaluate qualification test 10CFR50. App. B.
Provide a technical basis for licensing evaluation of proposed Qualification methodologies including test Criterion III.
qualification test programs.
sequencing and aging effects.
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Fire Protection Evaluate and validate 10CFR50. App. A Provide technical bases for revision to App. R and for new positions in present and Criterion 3 and regulations establishing fire protection and requirements proposed fire protection.
App. A for new plants.
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Human Engr / Man-Machine Evaluate ways to enhance human performance for safe quengiggcetton or Re ulations - 10CFR34, staffing requirements.
human performance for T IR54. conditfor.s of licensee - fitness for duty.
nuclear gf'Aso use in probabilistic 10CIR50, Appendia A General Design Criterion 19 -
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risk analysis Control Room Design; and 10CFR55, 'Jperator Training Requirements. 10CFR50.54 Control Room Staffing.
Improved reactor Results providing input to Control Room Review safety through human Guidelines, NUREG-0700 and standard review plans engineered equipment relevant to human factors.
and facilities.
Providing input to Reg. Guides:
- 1. 8 Operator Training f
1.47 Status Monitoring 1.97 Accident Instrumentation 1.97 Accident Instrumentation
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1.149 Training Simulators Hf-608-4 15FSI Certification and Training ANSI 3.1 Selection, Standards:
Qualification and AN53.5 Sieulators Training of Personnel ANS N-660 Operator Action Time at Nuclear Power Plants IEEE Colors and Symbols IEEE Subcommittee 7 on Human Factors Human Engineering Design Requirements IEEE 566 and 567, control Room Design 15Ad567.14 Certift.
Guidelines and criteria for computerized diagnostic cation and Training and display systems, of Instrument and Control Technicians Emergency Preparedness Improve the capability of 10CFR50.33, 50.47 Provide input and upgrade Reg. Guides 1.101, Federal, state and local 50.54 and App. E 2.6 and 3.42 and emergency preparedness government and licenses to regulations, licensees to mitigate the consequences of an 10CFR30, 40, 70 Establish emergency preparedness requirements accident.
for fuel cycles and saterial licensees.
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iLE DESCRIPTIGIl RELEVAllCE APPLICATION TO REGULATORV PROCESS Instrumentation and Evaluation of Safety Implication 10CFR50. App. A Principal effort in Resolution of USIA-47 Control of Control Systems and Associated Criteria 2.13,17 Electric Power Systems 18 and 26.
Evaluation of design, manu-10CFR50. App. A.
Technical Rasts for Guidelines on'the design, manufacture, facture. Installation operation. Criteria 2 and 13 lasta11ation, operation testing and maintenance of electric periodic testing and maintenance equipment important to safety, practices of l&C and electric equipment important to safety.
i Assessment of the state-of-the 10CFR50. App. A.
Data
- Base needed for regulatory decisions on Implementation of R.E.1.g7 art and current practices for Criterion 13 measuring important parameters during the cause of and following an accident.
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Evaluate regulatory lunact of use 10CFR50. App. A.
Regulatory Guides on Programunable Olgital Computers and Isolation devices.
of prograasnable digital computers with associated l
Isolation devices and other -
j advanced concepts in safety, control, alarm and information 1
systems.
Diagnostic Instrumentation 10CFR App. A.
TechnicalBasisforLicensingCriteriaforDiagnosticInstrumentation.
Research to Monitor Reactor Criteria 2 and 13 Safety i
Occupational Assess the sources of high 10CFR20.1 Provide a comprehensive emperimental basis for. assess 14g Ilcensee Protection radiation esposures la LWRs 10CFR20.101 plans to assure that collective (manress) occupational exposures i
and evaluate safety.
10CFR20.105 are ALARA a technical basis for modification of approaches to effectiveness and cost of 10CFR50.34 ensure ALARA.
41ternative decontamination EPA Occupational methods and operational approaches Esposure Guidelines to reduce these exposures.
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Communicated Threat Develop capability to perform 10CFR73 Provide advice and recommendations to the NRC. FBI. etc.. on Assessment comprehensive assessments of credibility of nuclear threats.
nuclear threats.
Reactor Vulnerability To identify power reactor 10CFR73 Studies systems which are vulnerable Provide NRC the technical basis for assessing the adequacy of current to insider sabotage and safe-regulations with regard to insider sabotage.
guard measures to offset this vulnerabi11ty.
Safeguards Aspects To identify human factors 10CFR73 considerations wRich could Provide the NRC with a technical basis for assessing the adequacy of f"
impact negatively on safe-current regulations with regard to human factors in safeguards.
guards systems and to identify a way to mitigate oi eliminate such impacts.
Power Reactor Safety / To identify and resolve safety / 10CFR50 and 73 Safeguards Interface safeguards interface problems Provide the NRC with technical basis for assessing current regulations at nuclear power reactors.
with regards to safety / safeguards interface problems.
Integration of hital Develop. demonstrate, and 10CFR50 and 73 Permit timely retrieval of site-specific information to support NRC's Area Products transfer a method of organizing analysis of and response to safeguards events at power reactors.
and rapidly retrieving existing site-specific systems information in response to reactor safe-quards events.
ISFSI Design Study To identify probable designs 10CTR72 and 73 Provide informat8on which can be used for assess'ing adequacy of which will be used for storing spent fuel.
safeguards at ISiSIs.
Standards and To improve measurement methods 10CFR70 Measurement Control standards. for calibration.
To provide the technical basis for regulatory decisions associated with' for NDA calibration methods, and measurement technology.
procedures for the quality control of measurement systems.
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Persmeter Alare Testing Evaluate under harsh winter 10CFA13 Provide the NRC with a technical basis for assessing adequacy of conditions specific perimeter current regulations la regards to perimeter alares.
alarms systems used at power reac tors.
Statistical Methods for Provfde standardized statts-10CFR70 Will serve as the basis for revision of several statistical regulatory nuclear Material ttcal procedures to be used guides.
Accountability by the nuclear industry.
Recossendations for Analyze past and current 10CFR75 Will identify areas of 10CFR75 for revision or update.
Strenghtening IAEA initiatives for strengthen-Safeguards IAEA safeguards in order to
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identify areas for further NRC efforts.
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Passive NDA Reference Documents current uses of 10CFR70 To be used as a guide for Ifcensing process.
Manual Passive NDA as used by the Nuclear Industry.
13 Statistical.
To develop methods and 10CFR70 To hel'p standardize statistical terminology and to address deficiencies Defletencies apsiroaches that will in the use of statistics in NRC regulations.
address the deficiencies noted in SECY 80-514.
Estimation Methods for Describe experimental programs 10CFR70 Provide's guidance for Ilcensee implementation of proposed prompt Material Holdup to better estimate the nuclear accountability requirements, material held up la process equipment.
Reactor Risk Analysis Develop methods for analyalog 10CFR50 App. A Review susceptibility of reactors to haan error and equipment failure.
Methods Dev4lopment reactor risk
& Data Evaluation Develop methods to apply safety goal and decision theory to safety s.
regulation.
o Predict operator and 10CFR50 App. A.
Operator training, procedures and control s'oom design.
maintenance errors.
Assess equipment failure rates 10CFR50, 34 Review industry analysis of' reliability data and recommend acceptable la reactor plants values for use in risk assessments.
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Reactor Risk Analysis Identify risks of nuclear power 10CFR50,100 Discover strengths and weaknesses in regulatory program.
plants.
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' Study sensitivity of risk 10CFR50. App. A prediction models to explore 10CFR50. App. E Support reactor safety standards development in rulemakings, safety issues '
10CFR100 resolution of generic issues. and regulatory reform.
I Improve methods for predicting 10CFR50. App. E.
consequences of large releases 10CFA100 Siting. emergency planning. rulemaking and policy studies.
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of radioactivity, j
Train selected NRC regulatory li/A l
staff mesMrs in risk assessment Risk Assessment is becoming a licensing technique.
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Transportation and Develop methods for analyzing 10CFR30. 40, 70. 72 Materials Risk risk for fuel cycle facillties Evaluate safety over-regulation, and weak spots in standards for fuel cycle, material facilities, and transportation.
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Assess the ablif ty of products 10CFR30.33 8
Provide experimental demonstration of the adequacy of the current containing Ilconsed nuclear 10CFR40.32 Ilcensing approach or a techalcal basis for modification to ensure materials to safely respond to - 10CFR70.32 safety.
use. accident. and disposal environments.
i Define and/or prowIde capa-10CFR71. App. A.4 Clar1fy ambiguous performance test spectfications; provide state-of-the-
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1 bility to evaluate radioactive 10CFR71.37 materfal transportation package 10CFR71.36 art analytical tools to estabilsh compliance. Provide technical basis I,
safety under normal and abnormal 10CFR11. App. R.1 for decision on revision of 10CFR71 to adopt model standards.
I transport conditions. Establish 10CFR71. App. A.2 l
a set of radioactive material 10CFR71. Subpart C. App. B package performance tests based
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conditions.
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High Level Weste (E W) Test methods to predict long-10CFR60 para. 60.111 Provide technical data and test methods that will be used to guide the package and Geochemical term performance of proposed Interaction matrin materials for high development and to assess the performance of waste forms and packages level waste forms and containers with respect to containment and controlled release of radionucildes.
i and assess release rates of radionuclides i
E W Repository Develop leformation to identify 10CFA60, 20 Provide technical bases for formulating regulatory guide rules and' Siting j
critical technical requirements ipa Standard 40CFRigt I
for E W repositories.
regulatory decisions, for ensuring that the host rock will safely accommodate repository structure and thet site and environs provide adequate long-tem natural barriers to radionuclides.
4 EW Repository Design Assess methods te evaluate 10CFR28. 60 l
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Ensure that regulations will provide containment capability of
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Operational Safety and reliability and safety of *
(EpAandStateress.)
repositories, and permit compliance with health and safety requirements Performance design, structures and Rellahility operational plans and the and with environmental quality requirements,
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/ reifability of sealants, backftll t
materials and monitoring l
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systems Analysis and Develop and demonstrate the 10CFR60 and epa 80ethods developed will be used to enable letC to use appilcant's plans Risk Assessment of EW appitcabillty of methods for Standard 40CFRitt i
i risk predictions of HLW and data to 1 t;: ' _.tly predict reliability of proposed repository repositories in various media performance and help determine the adequacy of characterization information.
Low Level Weste (LLW) Test methods and data to 10CFR61 Form and Container confirm the properties of law Provide technical bases for technical directives, and regulatory level waste forms and containers guides with respect to performance of waste forms and containers.
j to assess their performance in
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shallow land burial fatillties.
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Shallow Land Surial Evaluate perfomance of 10CFR30, 51. 61. 150 L
i of LL tedioactive existing shallow land burial Provide technical bases for assesslag and leproving regulatory actions l
hastes and Alternative sites; test validity of site for siting. designine, operating and monitoring shallow land burial and t
Disposal Methods characterization methods.
alternative disposal methods. Works cooperatively with states l
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facility designs and monttering (llentucky and New York) in developlag site closure methods.
I and site closure methods.
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.ti t DESCRIPTION RELEVANCE APPLICATION TO REGutATORY pa0CE55 Uranium Recovery Evaluate operational methods 10CFR20, 40, 51 Weste Characterization, and identify alternative pro-Provide technical bases for urantun afil taillags regulatory crf teria Operaticas; Management cedures to reduce risk and and tested methods for assuring that unacceptable anosnts of radon.
airborne particslates and liquid wastes are not discharged lato the Decommissioning and evaluate decommissioning environnent.
Survelliance and surveillance techniques to prevent radtoactive and toxic substances from escaping from urantum mill taillags Uras.tum Recovery Assess methods for selecting 10CFR20, 40, 51 Provide technical basis for regulatory standards and confirmatory 5tting. Pathways sites for urantim mill tallings and for predicitag the risks data to support Ilcensing actions to ensure the long-tern effectiveness lapacts
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to surface water and ground-of alli tallings and solution mining management procedures and assure water contamination by protection to the health and safety of the public is adequate, radioactive and tonic substances from uranium alli tatilngs sites and solutions mining.
Site safety setssology M 1ogy Studies of earthquakes and 10CFR100. App. A Provide basis for future rulemaking on Appendfx A 10CFR100..
other geologic hazards to (SeismicandGeologic develop realistic assessment Siting Criteria for Clarify ambiguities in the Regulation, of potential for ground shaking Nuclear Power Plants) ground rupture, soll foundation failure, and volcanic effects lagrove capability to ottifze advances in the state of knowledge in earthquake sciences, at nuclear fac111ttes sites.
Assessment of feastbtitty of seismic regionalization, Eliminate redundancies in ifcense appilcations concerning regional earthquake potential.
Meteorology.
Acquisition and analysis of 10CfR100. App. E Vertfy models and calculations used to satisfy requfrements for site 1(ydrology high quality field test data on meteorological dispers1on evaluation and emergency response.
of accidentially released radionucifdes out to 50 miles update methods and bases for revision of Draf t Regulatory Guide 1.145.
and over a variety of terraf as typical of nuclear power plant j
sites.
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- l Characteriae severe steen hazards 10CFA50, App. A Verify levels of conservatise of design bases for natural hazards.
(tornades, lightAlag and fleeds) i in difforent raglens i
i Studies te develop criteria.
Generic Site Develop siting criteria and specify of tigation practices to prevent criteria for entsting contaAinatten of public mater supplies and damage to the environment
& future sites through this segment of the liquid pathuey.
considering effects of j
Class g accidents Agustic & Alrhorne
,Radienuc11de transport in 10CFR20 t
Technical bases to support Ilconsing judgments and for environmental
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Effluents -
' surface innters 8 effects of 10CFt50 App. 1-1spect statements te leplement regulattens.
f Environmental lopects effluents on bieta 10CFR51 Alrhorne effluent pathways
- models and codes Enviremental lapact; populatlen dose assessments.
i llealth Effects Assess the ispect of 10CFA20 radiation exposure from To develop tegulatory guidance, set empesure standards and provide facilities and activities technical dets for risk / benefit analyses.
regulated by IIRC on heseen j
health and safety.
Etting Assess potential sites from 10CFR100 Siting rulemaking Early $lte Review, Environmental fapact. Assessment, the perspective of populatten 10CFR$1 Evaluation of Alternative $ltes.
density and land use, and 10CFR50. App. 4 develop methods for i
evaluating alternative sites.
l Assess social and scenamic Itational Environmental Enviromental Ispact Assessment Cast-Genefit Analysis. Evaluation l
lapact of nuclear reacters polic Act (1g63) of Alternative Sit &s.
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ORAFT REVISED BMI-2104 ASSESSMENT OF NUREG 5
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UNCERTAINTY RESEARCH 7
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Battelle suite of codes as used in BMI-2104.
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MAJOR UNCERTAINTY AREAS IN BMI-210li SOURCE-TERM ANALYSIS DESCRIPTION OF PilEN0ENA BMI-210ll CODES AREAS OF MAJOR UNCERTAINTY PRIMARY SYSTEM TilERMAL-ilYDRAULICS MARCll, K RGE UPPER PLENUM RECIRCULATION l
FUEL llEATUP AND DEGRADATION MARCil CORE ELT PROGRESSION FISSION PRODUCT RELEASE CORSOR CONTROL R00 VAPORIZATION FROM FUEL (IN VESSEL)
TE RETENTION PRIMARY SYSTEM FISSION PRODUCT TRARELT FISSION PRODUCT DEPOSITION TRANSPORT AND RETENTION DEBRIS-CONCRETE INTERACTIONS CORCON FISSION PRODUCT RELEASE FROM VANESA TE RETENTION CORE-CONCRETE E LT CONTAINMENT TilERMAL-ilYDRAULICS MARCil llYDR06EN BEllAV10R MARCil.
CONTAINMENT FlSSION PRODUCT NAUA, SPARC, TRANSPORT ICEDF
Area of Uncertainty:
Upper Plenum Recirculation
==
Description:==
In the MERGE code, it is assumed that the flow of gases and aerosols in the upper plenum is one-dimensional.
In reality, it would be expected that circular flow patterns would be established in this region because of steep temperature gradients. Accounting for this natural recirculation would significantly increase the calculated effective area for aerosol deposition and vapor condensation causing a substantial increase in predicted fission product retention in the priv, y system.
Resolution:
There is underway a small experimental program sponsored by EPRI at Westinghouse to investigate the natural con-vection pattern in simulated upper plenum designs. These results and the utilization of more detailed codes like TRAC and RELAP can be used to investigate the consequences of one-dimensional simplistic modeling. Present work at SNL with a TRAC /MIMAS code linkage will quantify this deficiency. Modifications to MELCOR in this area will be made if indicated by the TRAC /MIMAS studies.
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Area of Uncertainty:
Core Melt Progression
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Description:==
The modeling of core melting, slumping, and vessel melt-through' is uncertain because of the absence of large-scale tests,' and uncertainties in related hydrogen generation and melt composition (and quantity) have been identified as very important because of their effects on containment loads and ex-vessel releases respectively.
Resolution:
Major experimental programs in the ACRR, PBF, and NRU test reactors are underway to study core melt progression.
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Improved out-of-pile meltdown tests are also planned.at ORNL (Parker). Revised out-of-pile tests to study fission product and aerosol release at ORNL (Lorenz) and KfX (SASCHA) are also expected to provide some melt-progression infor1 nation.
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These experimental programs are strongly coupled to SCDAP (before melting) and MELPROG (after melting) modeling work at INEL and SNL.
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Area of Uncertainty:
Control Rod Vaporization 0%.cription:
I Earlier KfK tests produced a lot of silver vaporization while ORNL tests produced only a little. Using this cannon data base, NRC's contractors assume significant Ag aerosol generation in their model while IDCOR's contractors assume none. The timing of any control rod vaporization is also very important because Ag aerosols can be carriers for fission products and affect deposition only if they are present when the fission products are being released.
Resolution:
Most of the core-melt-progression experiments include tests with silver-alloy control rods, and the new out-of-pile tes 3 at ORNL (Parker) and KfK (SASCHA) will specifically address this issue. THI-2 control rod leadscrew examinations are also contributing data on Ag aerosol deposits. Explanation to reconcile the earlier KfK and ORNL test results have bee put forward,' and the expected new data should lead to model improvements. The new modeling will be incorporated in the VICTORIA subroutine of MELPROG.
==
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Area of i
Uncertainty:
Te Retention
==
Description:==
Tellurium is a very reactive element that readily forms tellurides with other metals. Evidence exists that Te is retained in unoxidized Zr cladding, and rudimentary modeling of this effect is included in BMI-2104. However, it is not known exactly how Te will be released from the Zr as Zr oxidation progresses in the vessel or how Te will be released from the core melt during its reaction with concrete in the containment. Furthermore, it is not known if Te will react with stainless steel surfaces in the vessel or in the con-tainment and be retained there after release from the core.
Resolution:
The fission-graduct-release programs at ORNL (out-of-pile) and PBF are all specifically investigating the behavior of Te. The TMI-2 core examinations will also attempt to detennine the disposition of Te. Recently initiated mass-spectrometry studies of the reactions of Te with Zr will provide basic vaporization and reaction data. The high-temperature chemistry program at SNL should also provide supporting information. Modeling of these results will be done in VICTORIA and TRAPMELT (in-vessel) and VANESA (ex-vessel).
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'M Area of Uncertainty:
Fission Product Deposition and Retention
==
Description:==
Deposition and retention of fission products during transport through the upper plenum and primary system piping are calculated by the TRAPMELT code. The predicted retention on surfaces in many cases is substantial, but the uncertainty regarding the amount of deposited fission products is large because several potentially important phenomena are neglected (chemical reactions between deposits, and surfaces.and between vapors and deposits, and local addition of decay heat where fission products are deposited).
Thus local temperature increases due to the deposits them-selves may cause revaporization of some of the fission products.
Resolution:
A high temperature chemistry program is underway at SNL to determine. chemical interactions with primary system surface materials of fission products in the vapor phase.
A study at ORNL of fission products and aerosols will provide experimental data for the controlling deposition mechanisms. Large aerosol transport tests (MARVIkEN, LACE) will also ccatribute to improve current modeling.
Presently we have no experiments to examine the revapor-ization aspect, but EPRI is conducting small-scale experiments at ANL to investigate aspects of revaporization from metal surfaces. We also have under investigation the possibility of utilizing mass spectrometry techniques to setup single experiments addressing revaporization of selected fission products from surfaces. Other proposals are also being evaluated. Modeling improvements reflecting new results wf11 be made to TRAPMELT.
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Table 1.
Newer codes that have the potential for upgrading comconents of the Battelle suite of codes used in BMI-2104
. Description BMI-2104 Newer Code RCS Thermal-Hydraulics MARCH, MERGE TRAC, RELAP-5 Fuel Heatup and Degradation MARCH
~i SCDAP, MIMAS.-
- MELRPI,
~
MELPROG Fission Product Release CORS0lr
- GRASS, from Fuel (in vessel)
VICTORIA RCS Fission Product Transport TRAPMELT TRAPMELT Molten Fuel Interaction MARCH WISCI with Coolant Debris-Concrete Interactions CORCON CORCON Fission Product Release frem VANESA VANESA Core-Concrete Melt (ex-vessel)
Containment Therwal-Hydraulics MARCH CONTAIN i
Hydrogen Behavior MARCH CONTAIN (HECTR)"
Containment Fission Product NAUA SPARC, NAUA, CONTAIN Transport ICEDF (SPARC, ICEDF,.
MAER05)*
Subroutines in CONTAIN e
. Table 2.
Corresponding Research Project to Develop or Upgrade Newer Codes.
Newer Code Description Research Project Number TRAC, RELAP-5 RCS Thermal-Hyrdaulics 3, 4, 7, 10, 14, 26, 31 SCDAP, MIMAS, Fuel Heatup and 1,2,3,4,5,6,8, MELRPI, MELPROG Oegradation 9, 10, 12, 13, 14, 15, 16, 17, 18, 19, 26, 28 GRASS Fission Product Release 4, 5, 9, 10, 11, 12, VICTORIA from Fuel (in vessel) 13, 14, 15, 16, 18, 19, 20, 22, 26 TRAP-MELT RCS Fission Product 5, 13, 14, 15, 18, 20, Transport 21, 22, 26, 30, 31, 35 CORCON Debris - Concrete 7, 26, 27, 28 Interactions VANESA Fission Product Release 5, 20, 26, 27, 28 from Core-Concrete Melt (ex-vessel).
CONTAIN (HECTR)
Hydrogen Behavior 4, 16, 22, 26, 29, 34 NAUA,-CONTAIN Centainment Fission 16, 20, 22, 25, 25, (SPARC,ICEDF, Product / Aerosols Transport 29, 31, 33, 34 MAEROS,QUICXM) l
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I ORGANIZATIONAL CNART March 1984 Includes the NRC reorganization.
Office of Research 4
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Office'of the Director l
Intf$c (oNcy,PlanningandControl o ans elvision of Alsk Analysis ivision of Radiattoa Divistoa of Engineering Division of Accident and Operations Pr grams and tarth Sciences Technology Evaluattoa p
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Meterial Eaglaserin0 Containment Systems Reactor Risk treach Waste k nagement Brance i
Branch gesearth Branch 4
Mechaatcal/Stnactural Regulatory Analysis &
Earth Sciencas Branch Aeactor Systems i
Eagleeering Branch 3
Research branch Materials Risk Branch Health Effacts trnach Electrical.Englaeorlag lastrumentation &
Fuel Systems Research thaman factors &
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Safeguards Branch
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Control Branch Brarzh i
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Chemical Enhecering Protection Branch aragh
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2 DIVISION OF ENGINEERING TECHNOLOGY Plans, develops, and directs comprehensive research and standards programs for 9 nuclear safety and nuclear materials safety in the design, qualification, construction, inspection, testing, operation, and decomissioning of nuclear power plants, nuclear reactors, and fuel cycle facilities with emphasis on the mechanical, structural, materials, electrical, instrumentation and control engineering, and chemical e~ngineering aspects of these facilities and materials; establishes or recomends policy, planning, and procedures for the research and standards programs as required to carry out the functions of the Division; coordinates these research and standards programs with other NRC offices to ensure that the programs are responsive to their needs; provides technical, assistance within NRC regarding resolution of generic issues and the development and application of research and standards to the solution of
' specific safety problems; provides funding guidance to NRC contractors, DOE laboratories, and other government agencies within the Division budget and consistent with NRC policy; maintains liaison with and provides technical input to other Federal Agencies, ANSI, professional societies, international agencies, and other organizations in assigned areas.
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3 Mechanical / Structural Engineering Branch Develops, recommends, plans, evaluates, and manages research programs and develops standards for the design, qualification, construction, inspection, testing, and operation of nuclear power plants, nuclear reactors, and fuel cycle facilities with emphasis on the mechanical engineering and structural engineering aspects of structures and components. Specifically, this branch has the responsibility for these mechanical and structural engineering aspects, including qualification of mechanical components and related personnel, effects of natural phenomena (e.g., seismic analysis, tornado missiles), classification of, mechanical components, load ccmbinations and associated design limits, inservice testing for functional adequacy of mechanical components, vibration, soil structure as a support material, soil / structure interaction, spent fuel shipping casks, waste disposal casks, and cranes.
This branch has the lead responsibility for the coordination and interfacing activities associated with Sections III and XI of the ASME Code.
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4 Materials Engineering Branch Develops, recommends, plans, evaluates, and manages research programs and develops standards for the design, qualification, construction, inspection, testing, and operation of nuclear power plants, nuclear reactors, and fuel cycle facilities and transportation of radioactive materials with emphasis on-the materials engineering aspects of structures and components. Specifically, this branch has the responsibility for these materials engineering aspects, including inservice inspection for structural integrity, corrosion, fracture mechanics, thermal shock, effects of environment on materials, and overall nondestructive examination programs 'and provides general support to other branches within and outside the Division of Engineering Technology for their materials-related needs.
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5 Electrical Engineering, Instrumentation and Control Branch Develops, recommends, plans, evaluates, and manages research programs and develops standards for the design, qualification, construction, inspection, testing, and operation of nuclear powe. plants, nuclear reactors, and fuel cycle facilities with emphasis on the electrical and instrumentation and control engineering aspects of these facilities. Responsibilities for the electrical engineering aspects include qualification of electrical components and related personnel, classification of electrical systems and equipment, effects of natural phenomena on electrical equipment, research on protection against electricity-related fires,' electric power generating distribution equipment, and the effects of plant electrical transient and anomalies on electrical equipment. Responsibilities for the instrumentation and control engineering aspects include ensuring that the methods of control and the instrumentation used in nuclear facilities are designed, applied, and utilized to minimize the probability of abnormal operation or accidents and to ameliorate the consequences of an accident if one does happen. Also included is the consideration of unusual operating conditions, hostile environments, equipment and material failures, system interactions, and man / equipment interactions.
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w-6 Chemical Engineering Branch Develops, recmanends, plans, evaluates, and manages research programs and develops standards for the design, qualification, construction, inspection, testing, and operation, and decommissioning of nuclear power plants, and nuclear reactors, and fuel. cycle facilities and for nuclear materials with emphasis on chemical engineering aspects of these facilities and materials.
Specifically, this branch has the responsibility for these chemical engineering aspects, including water and corrosion chemistry, criticality, decontamination, chemical cleaning, decommissioning, hydrogen control, fission product control, ventilation, and fuel handling, onsite waste treatment and storage, onsite and independent fuel storage, and fuel cycle unit processes.
This branch has the lead responsibility for development and coordination of the agency's decommissioning program.
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7 DIVISION OF ACCIDENT EVALUATION Plans, develops, and directs comprehensive research and standards programs for predicting nuclear power plant behavior under normal and abnormal conditions.
These activities include developing systems analysis capabilities related to all plant systems with emphasis on the primary and secondary coolant systems, the containment system, and the full system and considering their interactions with each other and the balance of the plant. Also included is the administration of the s'upport facilities needed to carry out test programs.
Establishes or recommends policy, pJanning, and procedures for the research and standards programs as required to carry out the functions of the Division.
Coordinates these research and standards programs with other NRC offices to ensure that the programs are responsive to their needs.
Provides technical i
assistance within NRC regarding resolution of generic issues and the development and application of research and standards to the solution of specific safety problems.
Provides funding guidance to NRC contractors, DOE laboratories, and other government agencies within the Division budget and consistent with NRC policy. Maintains liaison with and provides technical input to other Federal Agencies, ANSI, professional societies, international agencies, and other organizations in assigned areas.
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Reactor Systems Research Branch Plans, recommends, evaluates, and manages analytical and experimental research programs on the behavior of the primary coolant system of nuclear power plants, including interactions with the balance of plant under normal, abnormal, and severe accident conditions.
These programs include the development of verified codes and models of coolant behavior under accident conditions, the development of an understanding of special thermohydraulic phenomena encountered in abnormal or severe accident conditions and their implications for reactor safety and improved reactor safety system design, and the development of an understanding of the implications of selected precursor events and dominant accident sequences for primary system safety. This branch is responsible for a program of code and model development and verification; 1
the conduct of appropriate test programs to provide the necessary empirical I
data, including administration of the facilities associated with these programs that are dedicated to NRC; the detailed analysis of selected precursor events; and the development of a plant analyzer and other aids to reactor plant analysis.
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9 Containment Systems Research Branch Plans, recommends, evaluates, and manages analytical and experimental research programs on the behavior of the containment systems of nuclear power plants, including interactions with the balance of plant, under normal, abnormal, and severe accident conditions. These programs include the development of verified codes and models of the content and behavior of containment atmospheres and the transport of aerosol and fission products under accident conditions; the development of an understanding of how abnormal and severe accidents may be managed so as to reduce the frequency of challenges to containment integrity or mitigate their consequences; and the development of an understanding of the implications of selected precursor events and dominant accident sequences for containment system safety. This branch is responsible for a program of co'de and model development and verification; the detailed analysis of selected precursor events; the conduct of appropriate experimental programs to provide empirical data necessary to support the assigned analysis and regulatory support responsibilities, including administration of any associated facilities that are dedicated to NRC; and the coordination of LMFBR safety research within the Office of Research.
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10 Fuel Systems Research Branch Plans, recommends, evaluates, and manages analytical and experimental research programs on the behavior of fuel systems of-nuclear power plants under normal, abnormal, and severe accident conditions, including interactions with the primary system boundary. These programs include the development of verified codes and models of fuel behavior in the primary system under the above 3
conditions, ' including the release and transport of fission products and hydrogen from the fuel into ~~the containment; the devlopment of an under-standing of the coolability limits of the fuel system; the determination of how safety and control system interactions affect the fuel system under dominant accident sequences; and the possible implications of selected precursor events.
The branch is responsible for a program of code and model development and the' conduct of fuel tests and the administration of the test facilities that are dedicated to NRC, verification and analysis of fuel system test data under severe accident conditions, and the coordination of research on the radiological source term and on HTGR safety within the Office of Research.
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11 DIVISION OF RISK ANALYSIS AND OPERATIONS i
. Plans, develops, and directs comprehensive research and standards programs for '
independent risk analysis of all. elements of NRC-regulated nuclear activity.
The activities include systematic evaluation of reactor issues; development of analysis techniques to determine reactor reliability and overall risk, 4
including risk assessments and reliability analysis associated with the nuclear fuel cycle; human factors; and safeguards. Develops and improves risk 1
methodology and translates these efforts into effective tools to aid in making j
licensing and other regulatory decisions; establishes or recommends policy, i
planning, and procedures for the research and standards prograns as required i
to carry out the functions of the Division; coordinates these research and standards programs with other NRC offices to ensure that the programs are responsive to their needs; provides technical assistance within NRC regarding i
resolution of generic issues and the development and application of research and standards to the solution of specific safety problems; provides funding guidance to NRC contractors, DOE laboratories, and other government agencies i
within the Division budget and consistent with NRC policy; maintains liaison with and provides technical input to other Federal Agencies, professional societies, international agencies, and o'ther organizations in' assigned areas.
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12 Reactor Risk Branch Develops and uses systematic analysis techniques for assessing reactor plants to determine system or overall reliability; overall risk, including radioactive releases and consequences. P{epares reports, standards, and regulations related to its areas of resp.qnsibility. The Reactor Reliability Section develops new techniques for systems analysis, systems reliability engineering, and accident sequence prediction and has lead responsibility for applications of risk assess' merit to reactors in such programs as the Reactor Safety Study Methodology Applicaticns Program and the Interim Reliability Evaluation Program.
The Reactor Ri'sk Section develops probabilistic techniques for analyzing the response of reactors in serious accident sequences once core damage has begun. This includes the processes of core melt, energy release, and containment response and the analysis of fission product release and transport within plant systems and in the environment.
The Reactor Risk Section develops techniques to use comparative probabilistic analyses as a means to identify and evaluate useful alternatives in reactor siting, emergency planning, and plant design.
Develops probabilistic analys.
methods, assists in collecting reliability data, and prepares reports, standards, and regulations that cover the use of specific probabilistic analysis methods and reliability data in the regulatory process. Work includi; research data to obtain component failure rates, human error rates, and the frequency of occurrence of multiple failures of common causes and resear'iih ti develop probabilistic analytical methods for determining component, subsyster and system reliability. These methods cover challenges and failures from bo i external and internal causes and include, to the extent possible, means to identify and quantify uncertainties. This research supports development of quantitative criteria for acceptable reliability and acceptable risk and a........
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13 Human Factors and Safeguards Branch Develops and manages research and standards programs on human factors for the safe design, construction, and operation of nuclear facilities. These activities deal with safety-related aspects of the man-machine interface; plant procedures and tests; qualification, training, and licensing of persons in certain functions; and the organization and management of the plant pperating staff and the licensee corporate staff as a whole. The pursuit of these activities requires close coordination with the NRR Division of Human Factors Safety, the NRR Quality Assurance Branch, and the Office of Inspection and Enforcement. Develops research ahd standards programs on safeguards and the protection of certain nuclear materials and facilities. These activities support the NRC's objective of ensuring protection of the public health and safety and the national defense and respond to the unauthorized possession, theft, diversion, or use of special nuclear material and the sabotage of nuclear facilities. The pursuit of these activities requires close coordination with the NMSS Division of Safeguards.
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j Regulatory Analysis & Materials Risk Branch i
Carries out a systematic evaluation (which includes the use of PRA) of relevant safety issues and the infonnation needed to address those issues.
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Uses the results of NRC and other research programs to identify regulation 3
l changes needed to correct deficiencies in safety-significant areas or l
eliminate unnecessary regulatory constraints. Proposes or initiates i
i rulemaking, as appropriate, and manages complex rulemakings that span the 1
technical or organizational responsibilities of several RES branches or that involve novel or complex questions of regulatory policy. Plans, organizes, i
l and manages a research program dire ted toward improving the effectiveness of i
the NRC regulatory process (i.e., the process used in reaching and
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implementing decisions on regulatory issues) through the use of value/ impact i
analyses and PRA. Monitors and analyses administrative, judicial, and I
legislative developments that could affect NRC regulatory and research programs. Develops, documents, and implements policies and procedures for 1
i developing regulations, including preparation of a regulatory analysis on the i
impact of a proposed regulatory activity handling of petitions for rulemaking, and RES interactions with the CRGR. Prepares standards, guides and i
regulations for the pbduction and use of industrial, commercial and consumer products containing regulated quantities of radioactive materials. Develops and implements an agency-wide technology transfer program to train a cadre of l
PRA practitione.rs capable of evaluating PRA submittals and applying PRA j
techniques to regulatory problems, apprise NRC management and selected staff t
l of PRA results having an impact on the regulatory process, and expand the'use a
of PRA technology to support the NRC safety goal.
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DIVISION OF RADIATION PROGRAMS AND EARTH SCIENCES i
f Plans and directs research and standards programs in the earth sciences related to the performance of nuclear facilities in the managenent of radioactive wastes, natural phenomena, siting factors affecting public health j
and safety, and health effects and occupational protection from ionizing t
radiation. Establishes or recommends policy, planning, and procedures for the research and stamfards programs as required to carry out the functions of the Division and coordinates these research and standards programs with other NRC offices to ensure that the programs are responsive to their needs. Provides technical assistance within NRC on the resolution of generic issues and i
specific safety problems; provides funding guidance to NRC contractors, DOE i
Iaboratories, and other government agencies within the Division budget and I
consistent with NRC Policy; and maintains liaison with and provides technical input to other Federal Agencies, international and domestic agencies, and other organizations in assigned areas.
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16 Waste Manacement Branch Develops and manages research 6ad standards programs to provide regulations, criteria and guides and manages contractual work relative to the overall performance of nuclear facilities and the management of radioactive waste.
This involves research to improve understanding of the factors and phenomena resulting from routine licensed operations that significantly affect the public health, accidental releases of radioactive material, and waste facility system performance. These factors and phenosena include external events that affect facility safety; institutional and physical factors that affect the consequences of routine operations and accidents; and the operating, engineering, and system performance factors that affect waste isolation and containment. The branch develops standards and guides and recommends regulations in the areas of its responsibility. Carrying out these functions requires extensive coordination with other NRC Divisions. 00E, EPA, and other Federal and State agencies.
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