ML20128N140
| ML20128N140 | |
| Person / Time | |
|---|---|
| Issue date: | 06/14/1984 |
| From: | Ryder C NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Auer P CORNELL UNIV., ITHACA, NY |
| Shared Package | |
| ML20127A894 | List:
|
| References | |
| FOIA-85-110 NUDOCS 8507130050 | |
| Download: ML20128N140 (7) | |
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L'.' L Dr. Peter Auer Laboratory of Plasma Studies Cornell Univeristy Ithaca, NY 41850
Dear Dr. Auer:
Enclosed are the following items: .
(1) ' Enclosure 1 is an overview of new NRC codes that might eventually replace the Battelle suite of codes. An index of the research projects to develop or upgrade the codes is given in Table 2.
Each of the project numbers in Table 2 has a paper describing the objectives and the funding level. If these papers are of interest to you, then I can send them at your recuest.
i (2) Enclosure 2 is a related, but somewhat more comprehensive description, of the NRC analysis program for severe accidents in LWR's.
I can be reached on 301-427-4337.
Sincerely, l Christopher P.* Ryder Accident Source Term Program Office Office of Nuclear Regulatory Research l
Enclosures:
As stated 8507130050 850415 PDR FOIA ALVAREZ85-110 PDR
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ENCLOSURE 1 INFORMATION ON COMPUTER CODES RELATED TO THE SOURCE TERM REASSESSMENT Figure 1 identifies the Battelle suite of codes that was used for the source-tem calculations in BMI-2104. In the following paragraphs we will try to answer the question: What new codes might replace the Battelle suite of codes and lead to improved results in the near future? We will a_lso indicate how our research dollars are being spent for the development and valitation 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 thermal, physical, and chemical effects associated with core-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.
Tne 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 releasc and transfer from fuel, transport and deposition in reactor-vessel and primary-system components (ptping, steam generator, etc.), and transport and deposition in the centainment.
, 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 corpenents 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 consequence modules. As such, MELCOR has a broader scope than the BMI-2104 suite of codes, but it contains much less mechanistic detail, MELCOR is designed to be fast running, which is necessary for the large number of runs 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
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 siirrned-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 that 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 alcng 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 information 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 link has been developed to study the details of. core melt progression. Melt progression and related phenomena have been shown by our source-term work tn be very important in detennining 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, TRAPMELT, CORCON, VANESA, and CONTAIN. Such a linkage would
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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 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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Fission Product Reactor
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Pathway Behavior i r_-_______________,
! I i i i g l Fission Product , Comprehensive I
Inventory in Fuel Behavior of
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g l Release Retained i Containment l i _
l l l g from Fuel i in Fuel i MARCH I
jl CORSOR I
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Detailed Temp.,
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I Transport and =E - - - - - - -
Press. & Flow in 0 1 l Primary System
- Retention e i I 3 '
I TRAPMELT j ,
MERGE r______ _______
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VANESA I g i CORCON i
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l l and Retention 1 I NAUA, SPARC, ICEDF I 1
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" Source Term" Figure 1. Battelle suite of codes as used in BMI-2104
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f-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
! ' 'RCS Thermal-Hydraulics MARCH, MERGE TRAC, RELAP-5
, COBRA Fuel Heatup, Degradation MARCH SCDAP, MIMAS, MELRPI, MELPROG Fission Procuct Release CORSOR GRASS, from Fuel (in vessel)
Y1CTORIA.
, RCS Fission Product Transport TRAPMELT TRAPMELT
' Debris-Concrete I $n eractions CORCON CORCON Fission Product Releas'e from VANESA VANESA Core-ConcreteMelt^(ex-vessel)
Containment ThermallHydraulics MARCH CONTAIN hydrogen Benavior MARCH CONTAIN (HECTR)*
Containment Fjssion Product NAUA, SPARC, NAUA, CONTAIN Transport '
ICEDF (SPARC, 6
ICEDF, MAER05)*
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o Table 2. Corresponding Research Project to Develop or Upgrade Newer Codes.
Newer Code Description Research Project Number TRAP, RELAP-5, RCS Thermal-Hyrdaulics 3, 4, 6, 7, 10, 13, COBRA 14, 16, 24, 26, 31 SCDAP, MIMAS, Fuel Heat, Degradation 1,2,3,4,5,6,7, MELPRI, MELPROG 8, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18, 19, 26, GRASS Fission Product Release 4, 5, 9, 12. 13, 14, VICTORIA from Fuel (in vessel) 16, 18, 19, d , 26 TRAP-MELT RCS Fission Product 5, 13, 14, 15, 20, 21 Transport 22, 24, 26, 30, 31, 35 CORCON Debris - Concrete 7, 26, 27, 28 Interactions VANESSA Fission Product Release 5, 7, 20, 26, 27, 28 from~ Core-Concrete Melt (ex-vessel)
CONTAIN(HECTR) Hydrogen Behavior (Larkins), 4, 6, 7,13, 14, 16, 22, 24, 26, 29, 34 NAUA, CONTAIN Containment Fission 16, 20, 22, 24, 25, 26, (SPARC,ICEDF, Product / Aerosols Transport 29, 31, 33, 34 MAEROS,QUICKM)
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